Optimization of DU
.,CCycle Environmental and Economic Performance
by
Chad A. Bollmann
B.S., Ocean Engineering (1996)
U.S. Naval Academy
Submitted to the Department of Nuclear Engineering and the
Technology and Policy Program
in Partial Fulfillment of the Requirements for the Degrees of
Master of Science in Nuclear Engineering
and
Master of Science in Technology and Policy
at the
Massachusetts Institute of Technology
June 1998
01998 Massachusetts Institute of Technology
All rights reserved
............... .......... .....................
Department of Nuclear Engineering
Technology and Policy Program
Signature of Author .................
May 8, 1998
A
C ertified by ....................
Accepted by .......
..................
Mujid S. Kazimi
Professor of Nuclear Engineering
Thesis Supervisor
......................
... . .......
C ertified by ........ ....
..........
A
C"...
.............
..
...
...........................
Michael J. Driscoll
Professor of Nuclear Engineering Emeritus
Thesis Supervisor
Lawrence M. idsk
Chairman, Department Committee on Graduate Students
Department of Nuclear Engineering
Accepted by ........................................................
Richard de Neufville
Chairman, Technology and Policy Program
Scinci
VJt
u 4~i#
e
Optimization of DUPIC Cycle Environmental and Economic Performance
by
Chad A. Bollmann
B.S., Ocean Engineering (1996)
U.S. Naval Academy
Submitted to the Department of Nuclear Engineering and the
Technology and Policy Program on May 8, 1998
in Partial Fulfillment of the Requirements for the Degrees of
Master of Science in Nuclear Engineering
and
Master of Science in Technology anrd Policy
ABSTRACT
A study of the DUPIC (Direct Use of Spent PWR Fuel In CANDU) cycle was made to analyze
cycle performance relative to that of PWR and CANDU fuel cycles in terms of uranium
utilization and spent fuel production efficiency. The DUPIC cycle was found to be most
efficient in terms of minimizing spent fuel production as well as most efficient (within limits) in
terms of maximizing natural uranium utilization. It was found minimally productive to change
PWR fuel management practices in order to extend burnup in the CANDU portion of the cycle.
A policy analysis regarding potential implementation of the DUPIC cycle in North America,
between the U.S. and Canada, was also made.
CASMO computer models of PWR, CANDU, and CANFLEX fuel assemblies were created and
benchmarked. The PWR models were then used to develop analytical correlations that predict
PWR spent fuel isotopic compositions. Correlations that predict reactivity gain and burnup
increase in CANDU reactors due to AIROX processing of PWR spent fuel were created. An
estimate of fission product removal fractions during AIROX processing was developed. An
integrated model that predicts CANDU discharge burnup extension due to the use of spent PWR
fuel and AIROX processing was completed and used to analyze and compare the DUPIC cycle
to other fuel cycles.
The potential issues involved in implementation of a DUPIC cycle between the U.S. and Canada
were examined. Stakeholders and influential groups were identified and their values were
projected. A significant unresolved issue centers around which nation assumes custody of the
DUPIC spent fuel and the disposal costs of that fuel. A plan for DUPIC cycle implementation
was developed.
Thesis Supervisor: Mujid Kazimi
Title: Professor of Nuclear Engineering
Thesis Supervisor: Michael Driscoll
Title: Professor of Nuclear Engineering Emeritus
Acknowledgements
This project could not have been completed without the expertise and help of many
people. I am greatly indebted to a good many people: Professor Mike Driscoll and Professor
Mujid Kazimi, my advisors, Ron Ellis of AECL, Mike McMahon of The Mackenzie Group,
Lorne Covington of STUDSVIK of America, and Xianfeng Zhao and Mike Reynard, graduate
students in the Department of Nuclear Engineering at MIT.
Many thanks go to my parents for their loving support over the years and my
Dad's willingness to read and edit my writing. You missed out completely this time, Dad.
Table of Contents
Abstract
.............................
2.................2
.......................
Acknowledgements ................................................................................................................................................
T able of Contents ........................................................................................................................
................. 4
List of Figures ........................................................................................................................................................
L ist of Tables ............. .
...................................................................................................
3
......................
8
9
List of Symbols and Nomenclature .............................................................................................................
10
CHAPTER 1
Introduction and Background..........................................................................................................................
..................
1.1 Introduction ..............................................................................................................
1.2 Background...................................................................................................................................
1.3 DUPIC Cycle Justification ............................................................................
1.4 DUPIC Cycle Description ............................................................
1.5 REPORT STRUCTURE.......................................................
12
12
12
13
14
15
CHAPTER 2
Description of Modeling Process ......................................................................
2.1 PWR Initial Conditions ........................................................
2.2 Spent Fuel Isotopic Correlations .... ..........................................
......................
2.3 AIROX Reactivity Effects.
2.4 CANDU Burnup Correlations...........................................................
2.5 C onclusions ..............................................................................................................
16
.... 16
16
18
18
.................. 19
CHAPTER 3
PWR Model Description...............................................
3.1 Plant D escription ...................................................................................................................................
.......................................
3.2 Model Description........................................................................................
..................
3.3 Conclusions ..............................................................................................................
20
20
21
22
CHAPTER 4
.......................................................................
CANDU Model Description ...................................
4 .1 Plant D escription ...................................................................................................................................
4.2 Fuel D escription ....................................................................................................................................
............................
4.3 Model Description.................................................................................................
4.4 Standard (37-Pin) Model ........................................................................................
4.5 CANFLEX Model ....................................................................
..................
4.6 C onclusions ..............................................................................................................
23
23
26
27
29
31
33
CHAPTER 5
Development of PWR Spent Fuel Isotopic Correlations.............................................
5.1 PWR Correlation Development ...................................................................................
.........
5.2 Determination of Maximum Acceptable Error ............................................
...................
5.3 Conclusions .............................................................................................................
34
34
37
40
CHAPTER 6
......... ........
Analysis of AIROX Process ....................................
.......................
6.1 Description of AIROX Process ........................................
6.1.1 Spent Fuel Material Removal........................................................41
6.1.2 Spent Fuel Material Processing....................................................42
.....................
6.1.3 CANFLEX Bundle Fabrication ......................
..........
...............................................
6.2 Estimation of AIROX Fission Product Removal
..................
...
................................................
6.3 Conclusions
41
41
43
43
46
CHAPTER 7
47
....
.........................
Development of CANDU Burnup Model ............................
47
7.1 Isotopic Reactivity Worths and Effects ..........................................
47
.......................
Worths
7.1.1 Determination of Individual Isotope Concentration-Dependent Reactivity
48
......................
.................................
7.1.2 Determination of Isotopic Test Case Reactivity Effect
48
......................
7.2 Estimation of Fission Product Worths ..........................
.......... 49
7.2.1 Determination of Total Fission Product Worth .........................................
51
.................................
Worth
Reactivity
Product
Fission
of
Reference
7.2.2 Determination
51
.................
........................................
Burnup
Discharge
of
CANDU
Prediction
7.3
7.3.1 Net Change in Reactivity due to Fission Products............................................. 51
.................. 52
7.3.2 Overall Change in Reactivity .....................................
7.3.3 Estimation of Test Case CANDU Discharge Burnup .................................................................. 52
53
..............................
7.4 Model Integration....
54
.........................................
7.5 Conclusions ...
CHAPTER 8
........... ............
Analysis and Discussion of Results.............................................
..............................................................................
8.1 PWR Fuel Enrichm ent
8.2 Description of Compared Cycles..................................................
...
............................
8.3 Determination of Reactor Ratio in DUPIC ........................................
8.4 Effects of DUPIC on Discharge Burnup..................................................................... ......................
8.5 N atural Uranium U tilization ..................................................................................................................
8.6 Spent Fuel Production ...........................................................................................................................
.........................................
8.7 Conclusions ...
55
55
56
56
58
60
63
68
CHAPTER 9
..................................................
DUPIC Cycle Implementation in North America.......
9.1 History and Background of the DUPIC Cycle ........................................................................................
...... ............................................................................................
9.1.1 History ...........................
..........
9.1.2 Motivation for DUPIC Cycle Development.........................
................
9.1.3 Current Status of DUPIC Implementation ......................................
......................
9.2 Issues Surrounding Implementation of the DUPIC Cycle ...........................
..........................
9.2.1 Reduced Fuel Cycle Costs.............................................................
..........
.................................................
9.2.2 AIROX Plant Costs and Issues
..........
................................
9.2.3 Deregulation and Competitive Improvements ..............
9.2.4 Spent Fuel Reduction.........................
9.2.5 D UPIC Fuel Disposal ......................................................................................
......................
9.2.6 Conservation of Strategic Resources and National Security.......................
...................
...............................................
Resistance..
Proliferation
9.2.7 AIROX Processing and
70
70
70
71
71
71
72
75
77
78
80
82
83
85
9.2.8 Transportation ...............................................................................................................................
...................... 85
Analyst's Perspective and Policy Options ......................................
9.3.1 Analyst's Perspective................................................................ 85
86
9.3.2 Policy Options ...................................................
................ 89
9.4 Description of Stakeholders and Decision Makers..................................
89
9.4.1 Public Positions .................. ...........................................................................
9.4.2 U tility Positions .................................................................................... 91
.......... ........ 91
9.4.3 Governm ental Positions ...........................................................
...................... 94
9.5 Proposed Policy Implementation Method..................................
................... 94
9.5.1 W inning the Public ...........................................................................................
95
9.5.2 Gaining Utility Support...............................................................
96
9.5.3 Governmental Support....................................................
97
9.6 Conclusions ............................................................
9.3
CHAPTER 10
98
Conclusions and Future Work........................................................
98
10.1 PWR Correlation Development ......................................... .................................
10.1.1 Future Work: Additional Confirmation of PWR Correlations .................................... 98
10.1.2 Future Work: Additional Correlation Development........................................................... 98
99
10.2 AIROX Process Analysis.............................................
100
..........................................
Uncertainties
Removal
AIROX
Eliminating
10.2.1 Future Work:
.................... 101
10.2.3 Future Work: The Effects of Cooling Time .......................
102
10.3 CANDU Burnup Prediction..........................................
103
10.3.1 Future Work: CANDU Modeling in CASMO .....................................
104
10.3.2 Future Work: Relation of Reactivity Worth and Isotope Concentration........................
105
10.4 DUPIC Cycle Performance...........................................
10.4.1 Future Work: DUPIC Economic Performance............................................ 105
105
10.4.2 Future Work: Alternative Cycle Comparisons .........................................
106
10 .5 P olicy Analysis ..................................................................................................................
107
..............................
Economics
and
Cycle
Values
Public
of
Analysis
10.5.1 Future Work: Additional
10.5.2 Future Work: Use of DUPIC Cycle to Burn Weapons Plutonium........................ ..................... 107
10.5.3 Future Work: Comparison of Proliferation Resistance of DUPIC Cycle ................................... 108
R eferences ...........................................
......... 109
APPENDIX A
Burnup Correlations for Fixed Cycle Length or Batch Number ................................................. 112
APPENDIX B
Sample CASMO Input and Output for Reference PWR...................................
........ 113
APPENDIX C
Sample CANDU CA SMO Models.............................................................................................................
115
APPENDIX D
Alternate AIROX Fission Product Removal Forecasts...........................................
118
APPENDIX E
Influence of Cooling Time on Spent PWR Fuel...................................................
124
APPENDIX F
Estimation of Fission Product Absorption Fraction Removal During AIROX ......................................
128
APPENDIX G
Sample Determination of (Ak/AX), for U-235
.............
.....................
APPENDIX H
Benefits from Use of Slightly Enriched Uranium in CANDUs ..............................................
................. 135
138
List of Figures
.......... 14
..
Figure 1-1: Diagram of 43-pin CANFLEX Fuel Bundle .........................
17
Figure 2-1: Model Development Process Flowchart ........................................
21
Figure 3-1: Plan View of Reference PWR Fuel Assembly ......................................................
Figure 4-1: Calandria and Pressure Tubes in a CANDU-600 (Ref. 10)............................................ 25
30
..................
Figure 4-2: Plan view of 37-pin Model Layout ........................................
32
............................................................................................
Layout
M
odel
45-pin
of
View
Plan
Figure 4-3:
Figure 5-1: Comparison of CASMO PWR Data and Safeguards Plutonium Correlation..................................... 36
....... 38
....................
Figure 5-2: Graph of K-Infinite vs Burnup for CANDU Reference Case..................
....................................... 41
.............. ...................
Figure 6-1: Flowchart of AIROX Fuel Processing....
Figure 8-1: Achievable Burnup in Component Stages of DUPIC Cycle .................................................. 59
........... ............ 62
Figure 8-2: Natural Uranium Utilization ..............................................
Figure 8-3: Spent Fuel Utilization............................................................................64
Figure 8-4: Annual Spent Fuel Production Comparison of All Cycles ................................................ 66
67
........................
Figure 8-5: Annual Spent Fuel Production for DUPIC and PWR-only Cycles ...........
114
.....................................
PWR
Reference
for
Card
Figure B-1: Sample CASMO Input
........ 115
..................
Figure C-1: Sample CASMO Input Card for 37 Pin CANDU Model .............
Figure C-2: Sample CASMO Input Card for 43 Pin CANFLEX Model..............................................117
Figure E-1: Change in K-Infinite Over Time Due to Actinide Decay and Fission Product Buildup ...................... 124
Figure E-2: Change in K-Infinite Over Time Due to Fission Product Buildup and AIROX Removal ................ 125
List of Tables
20
Table 3-1: Operating Parameters of Reference PWR ........................................
24
.....
..............
.
...
...........................
Reactors
Table 4-1: Operating Parameters of CANDU
28
Expansion..........................................................
Thermal
Without
and
With
K-Infinite
of
Table 4-2: Comparison
30
..............................
Parameters..............................
Alteration
Model
4-3:
Table
31
Table 4-4: K-Inf vs Burnup for Model and Benchmark (Ref. 13) .......................................................................
Table 4-5: Comparison of Benchmark to Model Isotopic Composition for 37-Pin Bundle (Ref. 3) ..................... 31
Table 4-6: Comparison of Benchmark to Experimental Isotopic Data for CANFLEX Bundle (Ref. 5).................... 33
Table 5-1: Range of Parameters for Spent PWR Fuel Correlation Development.............................................. 35
.......... 35
Table 5-2: Isotopic Correlations for Spent PWR Fuel ........................................
37
..................................
Correlations
Fuel
Table 5-3: Maximum Errors in Isotopic Concentrations for PWR Spent
44
......................................................
Study)
(This
AIROX
During
Table 6-1: Estimated Fission Product Removal
49
.................
Examination.............................
Accuracy
and
Cases
Correlation
AIROX
Table 7-1:
Table 7-2: Sample Integrated Model of Reference Case........................................................................................ 50
Table 9-1: Summary of PWR, CANDU, and DUPIC Fuel Cycle Costs and Savings (1) ..................................... 72
......... 113
Table B-1: Summary of Reference PWR Fuel Isotopics and K-Infinite as a Funtion of Burnup .............
Table D-1: Comparison of Fission Product Percent Removal Forecasts During AIROX................................... 118
119
Table D-2: Analysis of Fission Product Chemical Properties ........................................
130
Fuel.........
Spent
Table F-I: ORIGEN Isotopic Fractional Absorption Estimation for 4.5 w/o, 50 MWD/kg PWR
135
Table G-I: Sample Determination of U-235 (Ak/AX), .........................................................................................
Table G-2: Comparison of Case 1 and Case 2 Isotopic Concentration-Dependent Reactivity Worths................ 136
138
Table H-: Approximate Discharge Burnup of SEU-Fueled CANDU ...............................................
List of Symbols and Nomenclature
LWR
Light Water Reactor
PWR
Pressurized Water Reactor
CANDU
Canadian Deuterium Reactor
DUPIC
Direct Use of Spent PWR Fuel in CANDU
AIROX
Dry fuel refabrication process
Xp
PWR Reload Enrichment
BdL
PWR Discharge Burnup
Bor
Cycle average soluble boron concentration
n
PWR batch number
Be
Cycle burnup
MOX
Mixed Oxide fuel
KAERI
Korean Atomic Energy Research Institute
AECL
Atomic Energy of Canada, Limited
INEL
Idaho National Engineering Laboratory
Ako
Initial excess reactivity
Akavg
Maximum acceptable reactivity error
Aki
Average individual isotopic reactivity worth,
(Ak/Ax)i
Per-isotope, cycle average change in reactivity with concentration
%Axi,Max
Maximum allowable percentage isotopic variances
Akfp
Reactivity worth in a CANDU of all fission products present in spent PWR fuel
Xfis
Overall fissile content
6Akfp
Net change in reactivity due to fission products
AKi
Total isotopic reactivity change
AK
Overall change in reactivity
ABdC
Change in CANDU discharge burnup
Bdc
Total CANDU burnup
Rx
AIROX fission product absorption removal fraction for uncooled fuel
Ak'fp
Post-AIROX total worth of fission products
Akfp,ref
Difference between CANDU reference case reactivity with and without AIROX
processing
BdC,ref
Reference CANDU discharge burnup
(F/P)p
Feed-to-product ratio in PWR
(W/P)
Tails produced per unit mass of reactor fuel
Rf
Number of PWRs required to fuel one CANDU
Q
Gross thermal power
L
Capacity factor
1
Thermodynamic efficiency
Uu
Natural uranium utilization
Usf
Spent fuel utilization
CHAPTER 1
Introduction and Background
1.1 Introduction
Reducing the cost of electricity generated by nuclear power has been, and remains, a
primary goal of both industry and academia. The world trend towards privatization of electricity
production has also served to increase the pressure to reduce expenses so that nuclear energy
remains competitive with other primary fuels. The DUPIC (Direct Use of Spent PWR Fuel In
CANDU) cycle was originally proposed in Korea as a means of reducing spent fuel production.
The DUPIC cycle also has the potential to reduce electrical generation costs by extracting more
energy from the reactor fuel. This increased production efficiency, when combined with the
decrease in spent fuel production, comprise the key advantages of the DUPIC cycle and make
powerful arguments that support implementation of the DUPIC cycle.
This study was undertaken with the goal of examining the environmental and economic
performance of the entire DUPIC cycle. Furthermore, this study addresses the effects of initial
PWR fuel cycle conditions on later-stage cycle performance and determines optimum
performance values for variables such as PWR initial enrichment (Xp) and discharge burnup
(BdL).
1.2 Background
The DUPIC cycle requires the use of two different types of reactors, PWRs and
CANDUs. PWRs (Pressurized Water Reactors) are cooled and moderated by light water and are
the most common type of nuclear power plant in the world with 343 of these reactors currently
planned or in existence. CANDUs (Canadian Deuterium Reactors) are cooled and moderated by
heavy water and are the second most common nuclear power plant, with 37 plants in the world.
(Ref 1) The U.S., Japan, and many European countries rely primarily on PWRs for nuclear
power production while only Canada relies exclusively on CANDUs. Korea is a unique country
because it generates electricity from both PWRs and CANDUs. This also makes Korea uniquely
suited to implement the DUPIC cycle.
1.3 DUPIC Cycle Justification
The DUPIC cycle reuses spent PWR fuel in CANDU reactors. CANDUs have a much
greater neutron efficiency than PWRs, due primarily to the use of heavy water as the coolant and
moderator. CANDU reactors can thus operate on fuel with much lower enrichments than PWRs.
For instance, CANDUs are designed to burn fuel with the natural concentration of 0.711 w/o U235 while PWRs must have fuel with much higher enrichments, usually at least 3 w/o U-235.
Because of the high initial enrichment of PWR fuel relative to CANDU fuel and the high neutron
economy of CANDUs, even fuel that will no longer sustain a PWR can achieve significant
additional burnup in a CANDU.
There is usually sufficient fissile uranium and plutonium
remaining in the spent PWR fuel to allow a CANDU to burn the spent PWR fuel considerably
longer than fresh natural uranium fuel. For instance, natural uranium fuel is usually discharged
from CANDUs with a fissile content (isotopes U-235, Pu-239, and Pu-241) of 0.25 w/o after a
burnup of approximately 8.3 MWD/kg. PWR fuel, initially enriched to 3 w/o and burned for 35
MWD/kg, still has a significant total fissile content, often greater than 1 w/o after discharge, and
can thus be burned for an additional 12 MWD/kg in CANDU reactors. The DUPIC cycle not
only extends the use of PWR fuel but, when compared to conventional cycles that use only
PWRs or CANDUs, the DUPIC cycle also significantly decreases the amount of spent fuel
produced during the generation of a given amount of electricity.
1.4 DUPIC Cycle Description
In the DUPIC fuel cycle, PWR spent fuel is stored for a period of time after being
removed from the PWR at a certain discharge burnup. After cooling, the spent fuel is shipped to
a special plant to be remanufactured into CANDU fuel bundles using a dry processing cycle such
as AIROX (Ref. 2). At this plant, the cladding is punctured and gases that are trapped in the fuel
rods are captured and stored for later disposal. The fuel cladding is then removed and the fuel
pellets are transferred to a furnace where they are subjected to alternating reduction/oxidation
reactions at peak temperatures of 1600 C. Volatile fission products are removed during this
process while the spent fuel pellets are reduced to a fine powder. This powder is then milled,
shaped, and sintered into CANDU fuel pellets. These pellets are reclad into fuel pencils and the
pencils are assembled into a 43-pin CANFLEX bundle as shown in Fig. 1-1.
Figure 1-1: Diagram of 43-pin CANFLEX Fuel Bundle (Ref. 34)
Note: In Fig. 1-1 the center fuel pin and the surrounding ring are larger in
diameter than the outer two rings. Some studies (Ref. 3) use different fuel
compositions in the different-diameter pins. In this study the fuel compositions
were uniform throughout all rings.
This refabricated bundle of spent PWR fuel can then be burned in CANDUs in place of
the natural uranium (non-enriched) fuel that is normally used. Computer analysis and practical
experience have shown that minimal, insignificant modifications to CANDU control systems are
required to safely use spent PWR (DUPIC) fuel (Ref. 3).
1.5 REPORT STRUCTURE
This study will address the effects of the choice of PWR fuel cycle on the overall
performance of the DUPIC cycle. Chapter 2 of this report describes the overall process that was
followed to develop the complete DUPIC cycle model applied in this study. Chapters 3 and 4
describe the PWR and CANDU computer models, respectively, created to model the reactor
physics and fuel burnup behavior. The development of PWR isotope correlations is discussed in
Chapter 5 while Chapter 6 describes the AIROX process and the creation of an AIROX model.
Chapter 7 relates the use of a method, similar to that used in Chapter 5, to develop correlations
for discharge burnup in the CANDU. Chapter 8 compares the environmental and economic
performance of the DUPIC cycle to conventional PWR and CANDU fuel cycles. Chapter 9
examines the potential public policy impacts of a DUPIC cycle proposal linking the U.S. and
Canada. Chapter 10 summarizes the overall conclusions of this study and outlines areas for
future work. Appendices A through I contain detailed explanations of important procedures and
document important results in support of the main text.
CHAPTER 2
Description of Modeling Process
One of the principal goals of this project was to develop an integrated model to predict the
achievable discharge burnup in the CANDU portion of the DUPIC cycle given initial PWR
conditions.
A flowchart of the process for model development is shown in Fig. 2-1.
The
development of this integrated model can be broken into four main stages: determination of
PWR initial conditions, development of PWR spent fuel isotopic correlations, prediction of
AIROX process effects on fuel, and development of CANDU burnup correlations.
2.1 PWR Initial Conditions
Certain characteristics of the fresh PWR fuel must be known in order to accurately
predict isotopic concentrations in the spent fuel. If reload enrichment, Xp, cycle-average soluble
boron concentration, Bor, and PWR discharge burnup, BdL, are known then the correlations
developed for the next stage of the model (discussed below) can be used immediately to
determine spent fuel isotopics.
If PWR discharge burnup is not known then the reload
enrichment in addition to either the batch number, n, or the cycle burnup, Bc, can be used to
predict discharge burnup.
This parameter can be calculated using correlations developed in
another MIT study (Ref 4), as summarized in Appendix A, before proceeding to the next stage.
2.2 Spent Fuel Isotopic Correlations
Once discharge burnup, reload enrichment, and boron concentration are known the
correlations developed in this study can be used to predict the isotopic composition of
Figure 2-1: Model Development Process Flowchart
Given one of the following sets of input data for PWR fresh fuel:
Xp, Bor, BdL
Xp, Batch number (n)
Calculate
BdL from
Xp, B,
-
PWR
-
AIROX
-
CANDU
correlations (1)
Obtain spent fuel isotope
concentrations from
correlations based on BdL,
Bor, and initial Xp (2)
Use AIROX correlation to obtain
AkFp for fission product removal
Calculate Ak for each isotope (Ak)
by comparing CANDU DUPIC
reference and test cases
Sum Ak, and AkFP, to determine
total net AK
Determine ABd from Eq. (7-5)
Add ABdc to reference BdC, Ref to determine overall
CANDU DUPIC BdC
Sum Bdc and BdL to obtain total DUPIC burnup Bd
Notes:
(1) BdL correlations for PWR are developed and discussed in Ref. 4.
(2) If boron concentration is unknown, a default value of Bor = 200Xp ppm can be used
spent fuel discharged from the PWR at the specified burnup.
The development of these
correlations is described in Chapter 5 of this study. The correlations predict the weight percent
concentration of important uranium and plutonium isotopes in terms of the initial mass of heavy
metal in fresh PWR fuel.
2.3 AIROX Reactivity Effects
The next stage of the model accounts for gains in the reactivity of the fuel that result from
the processing that occurs during AIROX treatment. Because of the high temperatures that are
used during oxidation/reduction and sintering, some fission products are vaporized and removed
from the spent fuel. When these products are vaporized, the overall parasitic neutron absorption
of the spent fuel is decreased and there is a proportional increase in the reactivity of the fuel. A
correlation was developed to predict the change in reactivity during the AIROX process. This
correlation and a list of vaporized fission products can be found in Chapter 6 along with a
discussion of the analysis performed on this stage of the DUPIC cycle.
2.4 CANDU Burnup Correlations
The final stage of the model predicts the achievable discharge burnup in the CANDU
portion of the DUPIC cycle. This prediction is made by examining the reactivity effects in a
CANDU of the significant uranium and plutonium isotopes mentioned above.
We first
determined the reactivity effects of a given isotope as its concentration in the fuel is varied. The
total change in fuel reactivity can then be calculated by comparing test case concentrations of
each isotope to the concentrations of that isotope in the reference case, 4.5 w/o-enriched PWR
fuel burned to 50 GWD/MT. This change in reactivity from the reference case, when summed
with a change in reactivity due to AIROX processing, can be used to determine the achievable
burnup in the CANDU portion of the DUPIC cycle. The correlations developed in this stage of
the model, along with a more detailed description of the methodology, can be found in Chapter
7.
2.5 Conclusions
Once the correlations were developed for each stage of the model, the stages were
integrated using a spreadsheet program so that the CANDU burnup and overall DUPIC burnup
are parametrically determined from set of initial PWR conditions. This integration of submodels
obviates the need for additional physics computer code runs over a wide range of PWR initial
conditions. This model format could also be used to predict results for different reactors (PWRs
or CANDUs with higher or lower powers, different fuel types and geometries, etc.) if specific
numerical relations similar to those used in this study are developed. The spreadsheet, including
specimen input and output, is discussed and documented in Chapter 8.
CHAPTER 3
PWR Model Description
The reference PWR for this study was Yonggwang-1, a Westinghouse 950 MWe Pressurized
Water Reactor. A Korean PWR was selected because it facilitated benchmarking the results of
this study's modeling to results developed by the Koreans during their study of the DUPIC cycle.
Additionally, much of the existing literature discussing the proposed DUPIC cycle has focused
on Korean PWRs and, more specifically, plants similar in design to Yonggwang-1. Finally, a
great deal of the existing spent PWR fuel has been discharged from older-generation PWRs such
as Yonggwang- 1.
3.1 Plant Description
The important operating parameters for this plant are shown below. These parameters
were determined using data obtained from AECL (Ref. 5), Fuji (Ref 6), and KAERI (Ref. 7).
The plant, as modeled, uses a 17x17 Westinghouse fuel assembly containing 264 fuel rods. All
parameters given below are for initial conditions and all dimensions given are cold dimensions.
Table 3-1: Operating Parameters of Reference PWR
Contractor
Gross Power (MWth)
Net Electrical Power (MWe)
Fuel Assemblies
Assembly Pitch (cm)
Fuel Pins per Assembly
Pin Pitch (cm)
Specific Power (kW/kgU)
Linear Power Density (kW/m)
Coolant Temperature (OC)
Fuel Pellet Diameter (cm)
Cladding Thickness (cm)
Cladding Material
Gap Thickness (cm)
Fuel Density (gUO 2/cm^3)
Westinghouse
2775
900
157
21.5
264
1.26
41.8
17.9
307
0.7844
0.0571
Zr-4
0.008
10.33
3.2 Model Description
Both the PWR and the CANDUs in this project were modeled using the CASMO-3
reactor analysis program developed by STUDSV1X NUCLEAR, a division of STUDS1K AB,
Nykoping, Sweden.
Because CASMO is designed to model PWRs, the Yonggwang-1 plant
could be represented in CASMO without any of the modifications and adaptations that were
necessary to model a CANDU reactor. A 17-cell by 17-cell fuel array, as shown in Fig. 3-1, was
created in CASMO and used for all PWR modeling. This array was created and run in CASMO
using input derived from the parameters in Table 3-1.
Figure 3-1: Plan View of Reference PWR Fuel Assembly
~" -- ~ ~ I-:W
The cIt
PR
SO
Ne.
K.MO.
Th W cr
weref~ly
N
%I
1
iZ
-tOK
kk
~-001"
w
wereX
rserdwem
f 5'he15*K
Z;''~'2
Z
z*Z
"W
H
-;
A--9* XF@edas
pt(e
K*
Z:I R f
" X:
ZW
X
Fuli
%I.
*-
Colnt*
=z::
NX
3.01.
%-.% ....
K
*.%- K
R
1:
%
....
.. K,R :j~t~
K:
*K*
K***.K*K:*K
Ok
K
:X..0
x';
X
.A
W-1
W.....a-0.
.01 00 00,
f~ ;
5SO
5.;;
@0.l
:R:
RA
i%**",:
- SO\*
i%
-k'a......
x RZ"
A. RK*
~5ik 0
..........
..
:
Rx
.
%Z--Z$X
: .:r,,%%%%%%%:kkkkk
.
%;Pi
W.,
SO.'
asmdld tseaysathtfulpwe odiin.Al ot0lrd
wthdrwn.Celstat
wuldnorall
conainthee
-*X ros
wre
odeed
a
emty
containing only moderator). Cycle-average soluble boron was used, when desired, to control
reactivity and to examine the effects of soluble boron concentration on spent fuel isotopics.
3.3
Conclusions
The modeling of the PWR portion of the DUPIC cycle was fairly straightforward since
basic, square fuel assemblies were used. For reference, sample CASMO input, output, and output
summaries for the reference PWR are contained in Appendix B.
CHAPTER 4
CANDU Model Description
Because this study used a square-lattice reactor modeling code, several adaptations were
necessary to convert the cylindrical CANDU fuel bundles and channels into equivalent square
structures.
Reference lattices for both a CANDU-600 using natural uranium fuel and the
reference reactor using mixed-oxide fuel were created and then compared to benchmark
calculations to confirm their validity. Significant adjustments had to be made to the model to
approximate real life conditions, but the effects of these adjustments proved to be minimal as the
results of this study show that CASMO-3 can be used to model CANDU fuel and predict fuel
reactivity and isotopic composition with acceptable accuracy for current purposes. It should be
noted that the most recent version of CASMO, CASMO-4, can explicitly model CANDU and
other hexagonal fuel lattices. (Ref. 8)
4.1 Plant Description
The design of a CANDU reactor core differs significantly from that of U.S. Light Water
Reactors. CANDU reactors are designed to bum natural (non-enriched) uranium and to be
continuously refueled in order to operate economically. This requirement for online fueling is an
important influence on the design of the CANDU.
CANDU fuel is contained in rods (fuel pencils) that are approximately 1.2 cm in diameter
and 50 cm in length. These rods are fabricated into fuel bundles consisting of either 37 or 43
pencils. These bundles are the basic unit of fuel for a CANDU reactor. As shown in Fig. 4-1,
fuel bundles are located inside pressure tubes that run horizontally through the reactor. Each
pressure tube contains 13 fuel bundles and pressurized heavy water coolant. A gas-filled gap and
calandria tube surround each pressure tube. The gap and calandria tube insulate the pressure
tube and can contain leaks in the pressure tube. All pressure tubes are inside of a calandria tank
which is filled with heavy-water moderator at low pressure and temperature.
The operating parameters for the reference plants were obtained from existing literature
and, in a few cases, directly from Atomic Energy of Canada, Ltd. (AECL).
The operating
parameters for the two reactors of interest here are listed in Table 4-1. These parameters are for
the most part identical and differ mainly in detailed lattice dimension and specific power. When
burning MOX fuel using a CANFLEX bundle, however, the Bruce-i reactor produces 850 MWe
net from 2832 MWth. This power uprating from the current levels shown in Table 4-1 is the
result of using enriched fuel instead of natural uranium fuel and is possible in newer reactors or
in existing reactors that have sufficient heat-removal capacity to avoid exceeding design limits.
(Ref. 9) Analysis by AECL and Lawrence Livermore National Laboratory (Ref. 3) as well as
this study included this power uprating.
All dimensions given are cold dimensions unless
otherwise specified.
Table 4-1: Operating Parameters of CANDU Reactors
Core Thermal Power
Net Electrical Power
2180
638
Bruce - 1
NU Fuel
2510
769
Channel Pitch
27.94
27.3
cm
Avg. Fuel Temperature
900
900
K
Fuel Density
10.36
10.36
gUO2/cm^3
Avg. Coolant Temperature
560.66
560.66
K
Avg. Coolant Density
0.812
0.812
g/cm^3
Moderator Temperature
345.66
345.66
K
Moderator Density
1.111
1.102
g/cm^3
Specific Power
25.4
32
kW/kg HE
CANDU-600
Units
MWth
MWe
Figure 4-1: Calandria and Pressure Tubes in a CANDU-600 (Ref. 10)
CALANDRIA
fuel rod diameter
13.08 mm
Calandria tube
thickness
1.4mm
Pressure tube inside
diameter (103.4 mm)
Pressure tube
thickness
4.34mm
Calandria
tube outer
diameter
131.8 mm
PRESSURE TUBE (WITH 37-PIN BUNDLE)
pellet diameter
12.1 mm
cladding thickness
0.38mm
FUEL ELEMENT
4.2 Fuel Description
There are two types of CANDU fuel bundles: 37-pin and 43-pin bundles. The 37-pin
bundle is the standard CANDU natural uranium bundle, in this paper it is referred to as a
"standard" or "reference" bundle.
The pins in this bundle all have the same diameter and
composition. In both bundles the pins are set in three rings around a single center pin. The 43pin bundle is referred to as a CANFLEX bundle and contains pins with two different geometries.
The inner pins have larger diameters than the outer pins. The CANFLEX bundle was designed
for alternate fuel types such as the mixed-oxide fuel (MOX) that is produced from excess
weapon materials or from LWR spent fuel as proposed for the DUPIC cycle. The outer pins
have smaller diameters so that they operate at lower linear powers and can be subject to higher
burnups without fuel failure.
In the proposed fuel cycle the MOX fuel produced in an LWR is homogenized using the
AIROX process. (Ref 2) AIROX is a dry-processing technique originally developed in the
1960s at INEL by Aerojet International and is currently being jointly investigated by AECL, the
Korean Atomic Energy Research Institute (KAERI) and the U.S. DOE. In AIROX the PWR fuel
cladding is removed and the fuel pellets are subjected to alternating reducing and oxidizing
conditions. (Ref. 11)
This turns the fuel into a powder that can then be milled to a fine
consistency and sintered into pellets for a CANFLEX bundle.
This process is expected to
produce very homogeneous fuel pellets that perform as well as conventional natural uranium
fuel. (Ref. 12) Also, some fission products (and all volatile fission products) are removed during
the oxidation-reduction process with some additional removal during the sintering stage. Since
most of these products act as poisons, their removal increases the reactivity worth and potential
burnup of the MOX fuel over that of the LWR spent fuel.
4.3 Model Description
The reference and CANFLEX fuel bundles were modeled using the CASMO-3 assembly
analysis program developed by STUDSVIK NUCLEAR, a division of STUDSVIK AB,
Nykoping, Sweden. The bundles were analyzed at steady-state full power. No poisons or other
reactivity control methods were used or required despite the additional reactivity of MOX fuel.
Studies by AECL showed that the additional reactivity caused by MOX fuel can still be
accommodated within the safety envelope of the Bruce reactors. (Ref. 13)
Both the standard and CANFLEX CANDU bundles were modeled using a 17x17 lattice
composed of 289 cells. An array of this size was chosen to provide the highest spatial resolution
possible. It was also decided that each pin should occupy one lattice cell and should be modeled
using its real dimensions in order to maximize realism. However, the individual square lattice
cells were quite large in volume compared to the actual triangular spacing in CANDU fuel
bundles. The fuel pin-to-coolant area ratio is smaller in CANDUs than in the equivalent cell of
the square CASMO model; thus the fuel pin spacing is artificially large and the area occupied by
coolant in the CASMO model is significantly greater than in actual CANDUs.
This extra
spacing in the fuel and coolant area of the lattice decreases the space available for the moderator
in the region representing the calandria tank. Thus, without any compensation, a CASMO model
using real pin dimensions would contain too much coolant and too little moderator.
This inaccuracy was corrected by applying a density adjustment factor to the coolant and
moderator. This factor was determined by comparing the real-life liquid areas to the model areas
and then altering the coolant and moderator densities so that the model contained the correct
volume of moderator and coolant, although in densities different from those in an actual
CANDU. Also, since accurate moderator and coolant densities were so critical to the model, the
thermal expansion function in CASMO was disabled and the hot, full-power densities of the
coolant and moderator were used as the pre-adjusted input densities. This ensured that exactly
the right volume of moderator and coolant was used so that the model will have the same fuel-tomoderator ratio as in an actual CANDU (see Table 4-3).
A calculation was done with the
thermal expansion function enabled to investigate the effects of this behavior. As is shown in
Table 4-2, disabling thermal expansion caused no significant error in the model.
Table 4-2: Comparison of K-Infinite With and Without Thermal Expansion
% Error
Allowing Expansion
Expansion Disabled
[MWD/MT]
[K-Inf]
[K-Inf]
0
1.1026
1.1028
-0.02
1.0611
1.0696
1.0649
1.0529
1.0378
1.0214
1.0046
0.9880
0.9720
0.9568
0.9426
1.0613
1.0698
1.0650
1.0530
1.0378
1.0214
1.0045
0.9878
0.9718
0.9565
0.9422
-0.02
-0.01
-0.01
-0.01
0.00
0.01
0.01
0.02
0.02
0.03
0.03
Burnup
200
1000
2000
3000
4000
5000
6000
7000
8000
9000
10000
The capabilities of CASMO-3 affected the model geometry in other ways.
Because
CASMO was designed to model light water reactors, its basis angle is ninety degrees. The pins
were thus placed in a roughly square layout and air-filled cells (simulating void) were used in the
corners of the fuel matrix to approximate a more circular form. The air cells were modeled as a
square channel with an extremely thin wall of Zircaloy. Second, the pressure tube, calandria
tube, and gap were approximated by the insertion of a flow channel from a boiling water reactor.
The gas gap between the two tubes was eliminated because CASMO-3 is incapable of modeling
such a feature. It was believed that the effects of this loss would be minimal because of the small
size of the gap and the small absorption cross-section of the gas, a belief that was verified by the
accuracy of the results. Additionally, the pressure tube in a CANDU is composed primarily of a
zirconium alloy which contains a small amount of niobium. When the two tubes were merged in
the model the overall material composition had to be altered slightly to reflect the absorption
cross-section of the other metal. Since CASMO does not have niobium in its material library,
another metal with a similar absorption cross-section was chosen and its density was adjusted to
approximate the cross-section of niobium. In both models the alternative metal that was used
was Cu-63.
4.4 Standard (37-Pin) Model
The 37-pin bundle was modeled using a 7 x 7 array of fuel pins and twelve air cells to
produce a roughly circular, 37-pin fuel matrix as shown in Fig. 4-2. The air cell is actually
modeled as a square pin containing air with an extremely thin cladding thickness. The bundle
was modeled with natural uranium fuel and the bundle parameters shown in Table 4-3 were
applied to approximate real conditions. The CASMO results were then compared to reactivityversus-burnup and isotopic benchmarks provided by AECL to verify the validity of the model.
(Ref. 5) The results of this comparison are presented in Tables 4-4 and 4-5 and show that the
natural uranium, 37-pin model accurately models a CANDU reactor. The difference between the
model and benchmark reactivity was within one percent until a burnup of 5000 MWD/MT was
achieved and then the error gradually increased to 1.5 percent. Since the trend is systematic - the
model consistently underestimates k - it should be possible, if desired, to increase agreement
with the benchmark through the introduction of further simple changes of factors such as pin
diameter and fuel density. Appendix C includes a CASMO input card for the 37-pin bundle.
Figure 4-2: Plan view of 37-pin Model Layout.
-w
II ..
*Tci
;i~Iijr
-
Homogenized Pressure /
4L N
Calandria Tube Surrounded By
240 External Moderator Cells
(moderator not shown)
[.[3[31313134,
33'A
IT5
M
5f O O
22
D
Fuel Pin
0=
= Coolant
,Air
Cells
Table 4-3: Model Alteration Parameters
Standard Model CANFLEX Model
Coolant Density Factor
0.682
0.5783
Moderator Density Factor
0.9746
1.0698
Zircaloy Channel Thickness
4.28
4.98
mm
Channel Zircaloy Composition
Channel Copper Composition
99.28
99.28
w/o
0.72
0.72
w/o
(simulating Nb)
Isotopic errors were also within acceptable ranges.
Differences between actual and
expected levels of uranium-235 and uranium-238 were less than one percent and plutonium
differences were about three percent.
There are likely to be greater errors in plutonium
concentrations than uranium because of the version of CASMO used. The university version
used at MIT treats resonance integral calculations differently than the commercial version and
these calculations are particularly important to the accurate prediction of plutonium production.
Table 4-4: K-Inf vs Burnup for Model and Benchmark (Ref. 13)
Bumup
[MWD/MT]
0
200
1000
2000
3000
4000
5000
6000
7000
8000
9000
10000
Model
1.1028
1.0613
1.0698
1.0650
1.0530
1.0378
1.0214
1.0045
0.9878
0.9718
0.9565
0.9422
K Infinite
CANDU
1.1109
1.0785
1.0778
1.0724
1.0603
1.0453
1.0298
1.0140
0.9985
0.9837
0.9698
0.9570
% Error
0.72
1.59
0.74
0.69
0.69
0.72
0.82
0.93
1.07
1.22
1.38
1.54
Table 4-5: Comparison of Benchmark to Model Isotopic Composition for 37-Pin Bundle (Ref. 3)
BdC = 8000 MWD/MT
Isotope
U-235
U-238
Pu-239
Pu-240
Pu-241
Pu-242
CASMO
0.205
98.486
0.252
0.110
0.024
0.007
AECL
0.206
98.389
0.249
0.110
0.023
0.007
Error*
0.40%
-0.10%
-1.25%
-0.40%
-4.46%
-2.83%
*Calculated using four significant figures for each value
4.5 CANFLEX Model
The CANFLEX bundle was also modeled using a 7 by 7 cell fuel array. To preserve
symmetry, however, only four air cells were used. As shown in Fig. 4-3, this produced a matrix
containing 45 fuel pins instead of the 43 pins that are in an actual CANFLEX bundle. It was
decided that since the CANFLEX bundle uses homogeneous MOX fuel and the fuel composition
is input into CASMO on a weight-percent basis, the addition of two extra fuel pins would not
cause unacceptable error provided the fuel-to-moderator and fuel-to-coolant mass ratios were
adjusted to account for the extra fuel mass. Additionally, because each pin is exposed at a
different rate, it would have been extremely difficult to normalize the modeling results to a 43pin bundle. Using fewer fuel pins was considered, but it was decided that maintaining lattice
symmetry was more important than modeling the correct number of pins. This decision was
upheld by the results, also shown in Table 4-6, of modeling an asymmetric 43-pin bundle. This
model contained the same number of pins and mass of fuel as an actual CANFLEX bundle, but
the error introduced because of the asymmetry was approximately the same as the 45-pin model
and significantly greater with respect to Pu-239, an important isotope.
A natural uranium
CANFLEX bundle CASMO card is included in Appendix C.
Figure 4-3: Plan View of 45-pin Model Layout
* * * *Homogenized
Pressure /
SCalandria Tube Surrounded By
**4 .
240 External Moderator Cells
(moderator not shown)
**
S=Coolant
*
Cells
**Air
= Inner Fuel Pin
=
Outer Fuel Pin
Table 4-6: Comparison of Benchmark to Experimental Isotopic Data at Discharge Burnup
for CANFLEX Bundle (Ref. 5)
Isotope
U-235
U-238
Pu-239
Pu-240
Pu-241
Pu-242
Benchmark
Composition
w/o
0.173
96.076
0.32
0.347
0.089
0.137
45-Pin
Model
w/o
0.186
95.873
0.319
0.355
0.094
0.125
% Error
%
7.51
-0.21
-0.31
2.31
5.62
-8.76
43-Pin
Model
w/o
0.184
95.876
0.318
0.354
0.095
0.125
% Error
%
6.36
-0.21
-0.63
2.02
6.74
-8.76
4.6 Conclusions
CASMO-3 can be used to model circular-geometry lattices with acceptable accuracy.
Modeling a CANDU reactor with a square-geometry program required adjustments to coolant
and moderator densities and approximation of real-life CANDU fuel-bundle geometries. These
adjustments allowed the creation of 37-pin and 43-pin CANDU fuel bundle models that
predicted spent fuel isotopic composition within reasonable bounds of error.
CHAPTER 5
Development of PWR Spent Fuel Isotopic Correlations
Both the PWR and CANDU correlations were developed using CASMO-3. These CASMO-3
simulations were created and analyzed in large part by X. Zhao, an MIT student, during the
course of a project examining the re-use of AIROX-processed fuel in PWRs.
The PWR
correlations allow the prediction of isotopic composition in the spent PWR fuel based on three
inputs: reload enrichment, Xp, discharge burnup, BdL, and cycle-average boron concentration,
Bor. Maximum acceptable levels of error in isotopic prediction from the PWR correlations were
then developed. Bounds on acceptable error were developed in terms of the reactivity worth of a
given isotope as it affects achievable burnup in the CANDU stage of the DUPIC cycle.
5.1 PWR Correlation Development
The development of PWR correlations involved the creation and execution of a
significant number of CASMO simulations. Reload enrichment, burnup, and cycle-average
soluble boron concentration were varied over current and potential ranges in a series of CASMO
calculations.
The ranges that were examined are shown in Table 5-1.
The results of the
CASMO simulations were compiled and the post-burnup heavy metal isotope concentrations
were then systematically linearized about the variables Xp, Bor, and BdL.
When necessary,
quadratic order terms were added to reduce the residual errors to within acceptable limits. The
expressions for PWR isotope concentrations that were derived in this manner are shown in Table
5-2. The concentrations given by the correlations in Table 5-2 are in terms of w/o of initial
heavy metal in fresh PWR fuel.
Table 5-1: Range of Parameters for Spent PWR Fuel Correlation Development
Notation
XP [w/o]
Enrichment
Burnup
Cycle Average
Boron Concentration I
Low Value
3
20
20M]
B [MWD/kg
BM1
Bor [ppm]
High Value
8
96
Step Size
1
10
1600
1600
200
0
I
Table 5-2: Isotopic Correlations for Spent PWR Fuel
Nuclide
Correlation [w/o]
U - 234
X24 = 0.0084 + 0.0066Xp - 0.00041B
U - 235
X25 = 0.26 + 0.579X - 0.008Xp2 - 0.038B + 0.065x10-3Bor
U - 236
X26 = 0. 197Xp - 0. 164X25 - 0.003B
U - 238
X28 = 99.12 - 0.693Xp - 0.074B - 0.153x10-3Bor
Pu - 238
X48 = -0.0011X, + 0.0001B + 1.1x10-5B 2 + 0.0026x10"3Bor
Pu - 239
X49 = 0.325 + 0.056Xp - 0.0005B - 1.45x10-5 B2 + 0.053x10 3 Bor
Pu - 240
X40 = 0.112 - 0.0155X + 0.0055B - 2.0x10-sB2 + 0.0029x10 3Bor
Pu - 241
X41 = 0.0237 + 0.004Xp + 0.00288B - 1.3x10-sB 2 + 0.016x103 Bor
Pu - 242
X42 = -0.0127 - 0.021Xp + 0.00404B - 7x10-6B 2
Table 5-2 Notes
(1) Xaz is the weight-percent concentration of the uranium or plutonium isotope of atomic number
ending in 'a' and mass number ending in 'z,' initial heavy metal basis.
(2) B is the fuel burnup at the desired analysis point, [MWD/kg].
(3) Bor is the soluble boron concentration, use 200 times Xp as default, [ppm].
The equations in Table 5-2 show the sensitivity of isotopic composition to soluble poison
concentration - an important factor that can explain differences between this study's results and
others which may not consider this variable. This also suggests that fuel management strategies
and assembly design details may introduce significant uncertainties that will exceed the
uncertainties associated with the use of these correlations. If this is the case, requisite accuracies
should be obtainable by re-normalization of the correlations to compensate for core-specific
variances such as leakage and usage of gadolinium compared to IFBA or WABA poisons.
It is interesting to note that there were past efforts to develop isotopic ratio correlations
for safeguard purposes.
A correlation of the form given by Eq. (5-1) was developed in a
materials safeguard study to predict the ratio of Plutonium 242 and 240 as a function of
plutonium-241 and -239 in PWR spent fuel. (Ref. 14)
xPu-2 42
= 3.3
Pu- 240
2
(5 -1)
Pu-241
XPu-
2 39
)
Both the CASMO output and the correlations from Table 5-2 satisfy the above isotope
ratio relation to within a maximum discrepancy of four percent and an average error of about
2%. Fig. 5-1 shows the consistent agreement of the results of this study's PWR CASMO data to
the above plutonium isotope ratio correlation.
Figure 5-1: Comparison of CASMO PWR Data and Safeguards Plutonium Correlation
0.43
Xp = 2
0.4 ------0.37 -----
0.34 .....
0.31
SXp 2.5
A Xp = 3
* Xp=4
e
Xp=6
= 5
Xp
+ Xp = 7
- Xp 8
_0.28
Correlation from Eq. 5-1
o 0.25
.. 0.22
0 .19
------------------ -
-- ----
0.16
0 .16 -.---. .---.....-. .. -..
---. . . . . .. ...------... ... .....
---. .. -----... ... ..-----... ... ..
------- --------.. .... ..---..-..-..--..--..-..---...
. -------------...
......
0.13
0.19
0.21
0.23
0.25
0.27
Pu-242/Pu-240
0.29
0.31
0.33
0.35
5.2 Determination of Maximum Acceptable Error
Table 5-3 summarizes the calculated maximum allowable error and the achieved
accuracies. The maximum error determinations were based on CASMO data calculated for the
CANDU reference case as defined in Chapter 3. The error in terms of the relative burnup effect
was then determined from the CASMO data for the CANDU reference case and is plotted in Fig.
5-2.
Table 5-3: Maximum Errors in Isotopic Concentrations for PWR Spent Fuel Correlations
Nuclide
Allowable Error
Correlation Max.
I%1
Error [%] (1)
7.3
1.2
1.6
0.04
6.4
U- 234
U- 235
U- 236
U- 238
Pu - 238
100
3.4
70
6
100
Pu - 239
4.0
0.77
Pu - 240
Pu - 241
Pu -242
47
8
100
0.86
2.40
1.76
Table 5-3 Notes
(1) Error is for range of enrichments 4 < X, < 7.
(2) Extending range of enrichments to 2 < Xp, 8 approximately doubles the error.
(3) The error cited = (correlation - CASMO) / (CASMO) x 100
(4) All data points fall within the cited bounds. The average absolute correlation error is
approximately one-half the maximum values quoted.
(5) 3.4 % error in U-235 gives approximately 2 % error in discharge burnup for CANDU DUPIC fuel.
Other isotopes are scaled according to Eq. (5-4) as discussed in Section 5.2.
An examination of Fig. 5-2 reveals that the total change in initial reactivity between the
initial state and loss of criticality is approximately 0.21.
After decreasing the total change in
reactivity by 0.04 to allow for leakage, the initial excess reactivity, Ako, can be determined to be
0.17. For a linear burnup history as shown in Fig. 5-2, the increase in relative burnup capability
can be approximated by Eq. 5-2 where Ak = 0.17:
Figure 5-2: Graph of K-Infinite vs Burnup for CANDU Reference Case
1.3
...... ...- - - - . . .. . . ... .. .. .. . . . . . .......
...
--..
...... ... .....-
. :'. -..
.;.- ....
-
. - .- . . . . .. . . . .-..
'---;--'- .. . . . ..--
1.2
-Delta k, I
1.1
Lealage Allowance = 0.04
1
0.9
--..
--------
. . -. 1------------------
1 + reactivity
- k-inf
08
0
2
4
10
8
6
12
14
16
Burnup (MWD/kg)
Ad
BdCRe f
"Akvg5Ak
(5-2)
0
Next the maximum acceptable relative error in burnup was set at 2%. This is equivalent
to an error of 0.4 MWD/kg of fuel over a total burnup of 20 MWD/kg. Substituting two percent
for the left-hand term of Eq. (5-2) gives us the maximum acceptable reactivity error, Akavg =
0.02/5 = 0.004.
A series of CANDU DUPIC runs were then made, systematically varying each nuclide
concentration individually to determine (Ak/Ax)i ,the per-isotope, cycle average change in
reactivity with concentration. Chapter 7 and Appendix G detail the methods used to determine
(Ak/Ax)i for the examined isotopes.
Substituting the (Ak/Ax)i value for U-235 into Eq. (5-3) and using the previously
determined maximum allowable reactivity error (0.004) gives the allowable variance in isotopic
composition for U-235. Dividing this value by the initial concentration of U-235 gives the
maximum allowable percentage variation in composition.
(5-3)
x =g
The maximum allowable percentage isotopic variances for the other examined nuclides,
%Axi,Max, were then determined by comparison to the maximum allowable percentage error for
U-235. For each isotope, %Axi,M,ax can be determined by dividing the concentration-dependent
reactivity worth of U-235 by that of the test nuclide and then multiplying by the maximum
allowable percentage variation in composition for U-235, %Axu-235,Max.
This calculation is
shown in Eq. (5-4).
%AX,,M=
Ak
(AkJ
AxU
--
235
,Mav
(v4)
The results of these calculations to determine the maximum allowable isotopic
composition error for the reference case are shown in the "Allowable Error" column of Table 53. The far right column of Table 5-3 shows the maximum error between the CASMO case and
the correlation prediction for each tested combination of reload enrichment, boron concentration,
and discharge burnup.
As is evident in Table 5-3, the nuclide which most severely challenges the accuracy
targets is U-235. While each individual isotope clearly meets its individual limits on error, care
must be taken that the cumulative individual errors do not sum to create an excessive overall
error. Individually, the accuracy of these correlations ranges from adequate to far more than
sufficient. Also, closer examination of the detailed numerical comparisons reveals no definite
remaining systematic bias that could be corrected in any simple fashion.
5.3 Conclusions
The PWR spent fuel composition correlations can be very useful because of their
simplicity and accuracy.
These correlations can account for a range of initial conditions and
allow the user to predict spent fuel isotopic composition without using a CASMO-equivalent
computer program or interpolating data from a graph.
While the correlations developed
specifically for this study are not necessarily applicable to all reactors, the development process
is straightforward and can be done for any other reactor of interest.
CHAPTER 6
Analysis of AIROX Process
In the proposed process spent PWR fuel, after cooling, will be shipped to an AIROX plant to be
refabricated into CANDU fuel. Unlike the British and French PUREX plants that use "wet"
separation techniques where spent fuel is dissolved and separated in acid baths, AIROX uses a
dry process to refabricate the spent PWR fuel into CANFLEX fuel bundles. During the dry
processing, some fission products are vaporized and removed from the fuel. The processed PWR
fuel will thus have a higher neutronic reactivity after AIROX processing. The gain in reactivity
from AIROX processing depends on the amount of fission products in the spent fuel and thus on
the discharge burnup of the spent PWR fuel.
6.1 Description of AIROX Process
A flowchart describing the overall AIROX processing path is shown in Fig. 6-1.
Figure 6-1: Flowchart of AIROX Fuel Processing
Spent PWR Fuel
Receiving
Pelletization
Sintering and
a
Fuel Pin Fabrication
6.1.1
Fuel Disassembly
Fuel Rod Decladding
4-- Powder Conditioning 1-- Fuel Oxidation/Reduction
Bundle Fabrication
DUPIC Fuel Shipping
Spent Fuel Material Removal
After the spent PWR fuel assemblies are cooled and shipped to the processing plant the
first step in the AIROX process is spent fuel disassembly. In this step the spent PWR fuel rods
are removed from the assembly structures and the structures are treated as solid waste. The fuel
rods are then punctured to remove fission gases and then declad. Any fission gases that are
removed are collected and stored for decay or disposal. The fuel cladding is removed from the
spent fuel pellets and mechanically cleaned to maximize recovery of spent fuel material. The
cladding, once cleaned, is then treated as solid waste. The removal of assembly structural
materials and fuel cladding is expected to generate approximately 0.3 MT/MTHM of solid waste.
(Ref. 15)
6.1.2
Spent Fuel Material Processing
The next step of AIROX processing involves treating the spent fuel pellets and debris
with alternating oxidation and reduction cycles to convert the pellets into a fine powder. The
peak temperature reached during the oxidation/reduction stage is approximately 1200 C. (Ref.
16) The spent fuel powder can then be blended with other processed batches of powders,
recycled, or milled to produce powder of the desired consistency and fissile content. After a
batch of powder meets specifications of size and content uniformity, it begins the pelletization
process. The fuel powder is compacted and granulated; a lubricant is then added to the powder
to facilitate pelletization.
Following the completion of a thermal treatment to remove the
pelletization lubricant, the green pellets are sintered at a temperature of 1600 C in a reducing
environment. (Ref. 16) The sintered pellets are then ground to dimension and surface finish
specifications and quality tested. Any defective pellets and all scrap materials from earlier pellet
and powder processing steps are recycled into the oxidation/reduction stage.
6.1.3
CANFLEX Bundle Fabrication
The final stage of AIROX fuel processing begins with the fabrication of the CANFLEX
fuel pins. The pellets are loaded into pins and the end caps are then welded onto the pins.
Quality testing will identify defective pins for fuel material recycling while the satisfactory pins
are assayed for fissile content. After measuring, pins of two different sizes are assembled into
CANFLEX bundles. Depending on the recycled fuel fissile content, some plans require that the
center pins of some CANFLEX bundle contain dysprosium poison mixed with natural uranium.
These poisoned center pins will be fabricated at a separate facility and shipped in a finished,
tested form to the AIROX plant. Finished bundles are examined to ensure they meet dimension
and weld quality specifications and satisfactory bundles can then be shipped to CANDU reactors.
Once again, unacceptable bundles are recycled by removal of fuel material, which is recycled,
and cladding and assembly structures are treated as solid waste.
6.2 Estimation of AIROX Fission Product Removal
Fission products are removed at three points during the AIROX process: cladding
puncture, pellet oxidation/reduction, and pellet sintering.
When the cladding is punctured,
gaseous fission products such as xenon, krypton, and some carbon are released. The greatest
removal of fission products occurs during the oxidation and reduction of the spent fuel pellets.
Temperatures up to 1200 C vaporize most of the remaining volatile fission products. While the
pellets are sintered at an even higher temperature (1600 C), there are few fission product oxides
with boiling points between 1200 C and 1600 C and it is expected that most of the volatile
fission products will have been removed prior to this step.
This study's estimation of the AIROX process's percent fission product removal is shown
in Table 6-1. Different predictions are cited for some nuclides in studies by AECL (Ref. 3), U.S.
DOE (Ref 17), and SCIENTECH. (Ref 15) Appendix D contains the removal predictions of
these three sources. It was decided that for the purposes of this study a prediction amalgamating
the forecasts from all available sources, with added consideration of other factors, was the most
reasonable approach to developing an estimation of fission product removal.
Table 6-1: Estimated Fission Product Removal During AIROX (This Study)
Fission Product
% Removal
14C
Cd
Cs
3H
I
In
Kr
Mo
Ru
Se
Te
Xe
All others
100
80
99
100
99
75
99
80
80
99
99
100
0
Table 6-1 was created by first examining existing literature for fission product removal
forecasts. These forecasts were then considered relative to these elements' chemical properties.
Factors such as boiling point and preferred oxidation state were used to forecast whether an
element and/or its oxide vaporize during exposure to the oxidation/reduction or sintering process.
If the element was expected to vaporize, then the removal forecasts given by other studies were
used to predict the percentage removal predicted in Table 6-1. Further discussion and tables
detailing the results of the removal estimation process can be found in Appendix E.
It is important to note that this study's final forecast for fission product removal differs
from forecasts by other sources. This is partially due to differences between fuel processing
methods: AECL based their forecasts on small-scale tests of the OREOX process while this
study considered the known temperatures of volatilization of the significant fission product
elements and their oxides.
Compared to our estimates, using AECL forecasts would
approximately double the fission product absorptions that are removed during AIROX.
For
instance, AECL removes most palladium, rhodium, silver, and technetium while this study's
forecast removes none.
Additionally, some removal forecasts could not be validated when
considered against fission product chemical properties and it was decided that a conservative
estimate of reactivity gain would be more desirable.
The removal fractions given in Table 6-1 then had to be converted to absorption fractions
to permit integration of these results into the model.
ORIGEN simulations were completed
which analyzed the neutron absorption fractions of various isotopes in spent PWR fuel. The total
absorption percentage removed during AIROX was then determined by multiplying each
isotope's percent removal fraction (given in Table 6-1) by the percent absorption fraction of that
isotope in spent PWR fuel (as determined in the ORIGEN runs).
These calculations are
discussed further in Appendix F.
This study determined that nuclides accounting for 32.8% of all fission product
absorptions in uncooled (i.e. no cooling time) spent fuel are removed during AIROX processing.
The removed fission products include xenon isotopes that would otherwise quickly decay. If the
xenon isotopes are not included in the removal estimate (i.e. prior to AIROX processing the
spent fuel is cooled just long enough for all xenon to decay), the fraction of fission product
absorptions removed drops to 14.4%. This study estimates that in uncooled fuel, Xe-135 and
Xe-131 account for 12% and 6% of all fission product absorptions, respectively.
Since DUPIC fuel is unlikely to be processed before at least a year or two of cooling, the
32.8% removal estimate does not accurately reflect the true removal of the AIROX process.
After the fuel cools for 10 years and the absorption fractions of various isotopes change due to
decay or yield, the AIROX process is expected to remove only 20.4% of all fission product
absorptions. This estimate closely agrees with a similar study by INEL (Ref 17). For further
discussion of the effects of cooling time on spent fuel composition and the AIROX process, see
Appendix E.
6.3 Conclusions
The determination of the fission product amounts that are removed during AIROX
processing has been discussed and shown to be a significant unresolved issue. Three different
expert groups use three different forecasts. Because fission products and fission product removal
will affect the reactivity of the CANFLEX fuel and thus affect DUPIC cycle burnup forecasts,
standardizing fission product removal estimates should be a primary concern of all parties
performing DUPIC analysis.
CHAPTER 7
Development of CANDU Burnup Model
The development of correlations to predict CANDU discharge burnup of AIROX-recycled spent
PWR fuel constituted a key part of this study.
These correlations allow the prediction of
discharge burnup based on the difference in isotopic composition between a test case and the
CANDU reference case.
7.1 Isotopic Reactivity Worths and Effects
7.1.1
Determination of Individual Isotope Concentration-Dependent Reactivity Worths
The first step in developing CANDU correlations required the determination of the
reactivity worths of individual isotopes. The development of these worths was briefly described
in Chapter 5 and will now be examined in some detail. A series of CASMO simulations were
created which perturbed isotopic concentrations from the CANDU reference case.
In each
simulation case, one isotope's concentration was increased by a factor sufficient to cause a
noticeable change in K-infinite. For each isotope, an increase of 1 w/o (i.e. x w/o becomes (x+1)
w/o) was used for the first case and in the second case the isotope's concentration was increased
by 1.5 w/o.
After the simulations were completed, case-by-case comparisons of the change in Kinfinite with isotopic concentration for the CANDU DUPIC spent fuel were made between the
reference and test cases. For each isotope at each burnup step between 0 and 14 MWD/kg, the
reference and test case K-Infinite values were compared and the differences recorded. The KInfinite differences at each burnup step were then integrated using Simpson's Rule to obtain the
average individual isotopic reactivity worth, Aki.
This value was next divided by the initial
isotopic perturbation to finally obtain each individual isotope's concentration-dependent
reactivity worth, (Ak/Ax)i, as shown in column 5 of Table 7-2. Appendix G contains a sample of
this calculation process for the isotope U-235.
7.1.2
Determination of Isotopic Test Case Reactivity Effect
The previously-determined (Ak/AX)i values are then used to estimate each individual
isotope's effect on reactivity, Aki, by comparing the difference in concentration between the
reference case and test case PWR spent fuel isotopic concentrations for a given isotope. The test
case concentrations can be obtained from the PWR correlations discussed in Chapter 5 or from
CASMO computer models of the reference PWR. The following equations can then be used to
calculate individual (Aki) and total (AKi) isotopic reactivity changes:
Ak, = (Xes, - X,f
AK,=ZAk,
(7-1)
(7-2)
The values obtained from these equations can then be used to calculate CANDU
discharge burnup as described in Section 7.3.
7.2 Estimation of Fission Product Worths
In order to estimate the reactivity gain from fission product removal during the AIROX
process and the effect on CANDU burnup, it was first found necessary to develop a correlation
expressing the reactivity worth in a CANDU of all fission products present in spent PWR fuel
(Akfp). The worth of fission products in the CANDU reference case then had to be determined.
7.2.1
Determination of Total Fission Product Worth
A series of CASMO simulations varying both PWR reload enrichment and discharge
burnup as shown in Table 7-1 were completed. These simulations were chosen to isolate and
examine the relationships between burnup, reload enrichment, and reactivity gain due to fission
product removal. The first part of the simulation calculated the spent fuel characteristics of PWR
assemblies subject to the initial conditions shown in Table 7-1. Varying amounts and types of
fission products were then removed from sets of CANFLEX bundles created from the uncooled
PWR spent fuel generated in the previous simulation step.
Table 7-1: AIROX Correlation Cases and Accuracy Examination
BdL
[GWD/MT]
45
Akfp
AIROX Correlation
% Error
1
Xp
[w/o U-235]
4
0.127
0.127
0.39
2
3
4
5
5
5
45
50
60
0.126
0.128
0.149
0.127
0.134
0.149
-0.40
-4.69
0.00
5
6
65
0.158
0.157
0.95
Case
Note: The basis of estimate in Table 7-1 is 100% of fission products removed after 0 years cooling time.
CASMO simulations were again run, this time examining the CANFLEX bundles created
in the previous step.
The k-effective results of these CANFLEX simulations were then
linearized with respect to PWR discharge burnup, PWR reload enrichment, and overall fissile
content Xfis. The results of this linearization are given by Eq. (7-3) and Table 7-2.
The linearization of these results, examining Akfp with respect to Xp and BdL, revealed a
very strong, linear dependence of Akfp on PWR discharge burnup.
This relation should be
expected: as burnup increases in the PWR so does the fission product inventory. This increase in
fission product concentration steadily depresses reactivity in the PWR and will similarly affect
CANDU reactivity. The correlation given by Eq. (7-3) was compared to other correlations of the
same data that also accounted for Xp and fissile uranium and plutonium concentrations, Xfn. The
correlation in Eq. (7-3) was found to have comparable accuracy, and minimal complexity, when
compared to the other examined correlations. Equation (7-3), with BdL in terms of MWD/kg, is
used in the integrated model to predict the residual reactivity worth in a CANDU of all fission
products present in the uncooled spent PWR fuel.
Table 7-2: Sample Integrated Model of Reference Case
Initial PWR Xp [w/o
Bd, PWR
Boron [ppm] (1)
BdC,ref
Delta ko
Delta Kfp
IROX FP Remova
Delta Kfp, ref
Isotope
24
25
26
28
48
49
40
41
42
Total Ki
Delta Delta Kfp
Delta BdC
CANDU BdC
DUPIC Burnup
4
45
800
14.40
0.173
0.1265
20.37%
0.10073
w/o
MWD / kg
ppm
MWD / kg
Composition [w/o]
Reference Case Correlation Delta xi
0.017
0.778
0.521
92.959
0.025
0.522
0.193
0.174
0.079
0.016
0.790
0.523
92.896
0.024
0.540
0.259
0.156
0.071
-0.001
0.012
0.002
-0.063
-0.001
0.018
0.066
-0.018
-0.008
Delta k /
Delta xi
-0.031
0.150
-0.011
-0.001
-0.076
0.194
-0.044
0.276
-0.031
Delta ki
0.00002
0.00180
-0.00003
0.00005
0.00004
0.00340
-0.00291
-0.00502
0.00025
-0.00240
0.00000
-0.20 MWD / kg
14.20 MWD / kg
59.20 MWD / kg
Notes:
(1) Default Boron value of 200 Xp is assumed.
(2) The values of reference case composition (column 2), Dk/DXi (column 5),
BdC,ref, and AIROX removal are the same for all cases; all other values
change with test case conditions.
Akf = pf = 0.059+ 0.0015B
(MWD/kg)
(7-3)
7.2.2
Determination of Reference Fission Product Reactivity Worth
The reference worth of remaining, post-AIROX fission products, Akfp,ref, was determined
using a CASMO simulation of the reference CANDU case. This simulation removed fission
products, in the amounts shown in Table 6-1, from the PWR reference case (Xp = 4 w/o, BdL
45 MWD/kg) and then burned the post-AIROX fuel in the reference CANDU.
=
The initial
reactivities of the reference CANDU case with and without AIROX removal were then
compared and the difference between the two initial reactivities, Akfp,ref, was recorded. This
value is shown as "Delta Kfp Ref' in Table 7-2.
7.3 Prediction of CANDU Discharge Burnup
The total fission product worth, when combined with AIROX fission product removal
forecasts from Chapter 6 and compared to the reference post-AIROX fission product worth,
allows prediction of CANDU burnup extension due to AIROX fission product removal. After
the net change in reactivity due to fission products, 8Akfp , is determined, this value can be
combined with the total isotopic reactivity change, AKi , to give the overall change in reactivity,
AK.
The total change in reactivity can then be converted to give the change in CANDU
discharge burnup, ABdc ,and finally permit calculation of the total CANDU burnup, BdC.
7.3.1
Net Change in Reactivity due to Fission Products
The first step in calculating the net change in reactivity due to fission products,
Akfp ,
requires adjusting the total worth of fission products from Eq. (7-3), Akfp , to account for the loss
of fission products during AIROX processing. This adjustment is shown in Eq. (7-4) where Rx is
the AIROX fission product absorption removal fraction for uncooled fuel as given in Chapter 6.
Ak' = (1- Rx)Ak
(7-4)
The post-AIROX total worth of fission products, Ak'fp , can then be compared to Akfpef
to get the net change in reactivity due to fission products, 6Akfp , as shown in Eq. (7-5). This net
change is due to fission product buildup and removal differences between the test case and
reference case.
AkfP = Akp,,, - Ak', = Ak,,ref - (I- R x ) Ak
7.3.2
(7- 5)
Overall Change in Reactivity
The overall change in CANDU reactivity can now be predicted by summing the results
obtained from Eq. (7-2) and Eq. (7-5), the total isotopic reactivity change and the net change in
reactivity due to fission products, as shown in Eq. (7-6).
AK = AK, + SAkfp
7.3.3
(7 - 6)
Estimation of Test Case CANDU Discharge Burnup
If K-infinite and reactivity are assumed to be approximately linear functions of burnup,
the results of Eq. (7-6) can then be used to calculate the net change in CANDU burnup by
substituting the appropriate values into Eq. (7-7):
A
dC,Re
_ AK
(7-7)
Ako
Once the net change in CANDU burnup is determined this value can be added to that of
the reference CANDU discharge burnup, BdC,ref, to determine the test case CANDU discharge
burnup, BdC, as shown in Eq. (7-8).
BdC= BdC,ref +
Bd
(7-8)
7.4 Model Integration
Once all stages of the model creation process were completed, the model was integrated
using a standard spreadsheet program. A sample calculation of the integrated model, using the
reference case, a PWR initially enriched to 4 w/o and burned to 45 MWD/kg, is shown in Table
7-2.
Please note that 8Akfp , given in Table 7-2 as 'Delta Delta Kfp', should be zero because
the model test case is actually predicting the results of the CASMO reference case. Because
fission product inventories and AIROX removal are the same, there is no difference between
Akfp for the test case and Akfp,ref for the reference case. Examination of Table 7-2 also reveals the
error resulting from use of the PWR correlations instead of CASMO modeling to predict
uranium and plutonium isotope concentrations in the PWR spent fuel. These isotopic errors
cause CANDU burnup to be shortened by 0.20 GWD/MT. This is equivalent to a burnup error
of 1.4% in the CANDU stage and an overall DUPIC burnup error of 0.3%, which is quite
acceptable for present purposes.
The analytic model was also examined for accuracy over the given range of PWR reload
enrichments and discharge burnups; the results of DUPIC CASMO models agreed with the
predictions of the analytic model over a range of examined burnups. Error in CANDU discharge
burnup prediction ranged from approximately (-8%) at the lower range of examined burnups to (2%) in the mid-region and (-4%) at the high end of the range of examined burnups. It is believed
that the analytic model's accuracy can further be increased by reexamining the actinide worth
ratios for the CANDU portion of the model (Dk/Dx values in Table 7-2), particularly since the
model consistently underestimates achievable CANDU burnup.
7.5
Conclusions
The determination of concentration-dependent isotopic and fission product reactivity
worths completed the development of the CANDU modeling stage. The CANDU stage was then
integrated with the PWR and AIROX stages to create a complete, integrated model. The model
was, in turn, benchmarked to DUPIC cycles operating at low, medium, and high reload
enrichments. This proved that the integrated model permits analytical determination of CANDU
discharge burnup with a minimum error of (-1.5%) and a maximum error of (-8%). It should be
possible to further reduce these errors by recalculating the concentration-dependent reactivity
worths in CANDU of the uranium and plutonium isotopes.
CHAPTER 8
Analysis and Discussion of Results
The development of an integrated, parametric model to predict discharge burnup from the
CANDU portion of the DUPIC cycle simplifies analysis of the entire DUPIC cycle.
Examination of DUPIC cycle fuel efficiency was particularly interesting and revealed general
guidelines for maximizing the fuel efficiency of the cycle.
8.1 PWR Fuel Enrichment
As discussed above, PWR fuel differs from CANDU fuel in that PWR fuel is enriched in
U-235. Enriched fuel contains more fissile uranium (U-235) than is present in natural uranium;
enriched fuel generally is composed of 3 - 5 w/o U-235 while natural uranium (NU) contains
approximately 0.711 w/o U-235.
Enrichment plants produce enriched fuel by preferentially
concentrating the U-235 from a large amount of natural uranium into a smaller amount of
enriched uranium. Increasing the concentration of U-235 requires increasing amounts of NU
feed. Enrichment, consequently, creates significant amounts of depleted uranium waste (tails)
with much lower concentrations of U-235 than are present in natural uranium (0.2 - 0.3 w/o
compared to the 0.711 w/o in NU).
A measure of the natural uranium required to produce one mass unit of reactor fuel is the
feed-to-product, (F/P)p , ratio which is readily computed from total NU and U-235 material
balances. This ratio is directly proportional to enrichment and is given in Eq. (8-1) for tails with
a U-235 concentration of 0.25 w/o. The (F/P) ratio can also be used to determine the tails
produced per unit mass of reactor fuel, W/P.
F/P = 2.17Xp - 0.54 = 1+ W/P
(8-1)
As shown in Eq. (8-1), both natural uranium consumption and tails production increase
with increasing PWR enrichment. CANDUs, on the other hand, do not currently use enriched
fuel and thus produce no tails. While tails are much less toxic than spent fuel, the volume of tails
required to produce a given volume of reactor fuel is significantly greater than the volume of fuel
produced.
8.2 Description of Compared Cycles
By varying initial PWR conditions, the efficiency of the DUPIC cycle was determined as
measured by tails production, natural uranium consumption, and spent fuel production per unit
electrical power generation. These efficiencies were then compared to the efficiencies of other
cycles with the same power generation. Cycles that produced all of their electricity from PWRs
(PWR-only)
or
natural
uranium-fueled
CANDUs
(CANDU-only)
were
considered.
Additionally, the efficiencies of a Parallel cycle were determined. This Parallel cycle produces
power by the same mix of natural-uranium CANDUs and PWRs as the DUPIC cycle but uses
fresh natural uranium fuel in the CANDU instead of AIROX-processed spent PWR fuel. Only
the DUPIC cycle uses MOX-fueled CANDUs.
8.3 Determination of Reactor Ratio in DUPIC
An important factor in determining the efficiencies of the DUPIC and Parallel cycles was
the reactor ratio, or the number of PWRs required to continuously fuel a CANDU reactor.
Because PWRs burn fuel longer and have slightly greater thermal efficiency than CANDUs,
PWRs produce much less spent fuel per unit of electrical power than a CANDU. For instance, a
PWR that burns its fuel to 35 GWD/MT produces approximately 11 GWDe per metric ton of
spent fuel while a natural uranium-fueled CANDU produces only about 2.7 GWDe per metric
ton of spent fuel. Because CANDUs obtain less power per unit fuel than PWRs, for a given
amount of power production a CANDU must use more fuel. It follows from the above example
that CANDUs will also consume more fuel per day of operation than PWRs; a 950 MWe PWR
will burn < 0.1 metric tons of fuel per day while a 750 MWe CANDU will burn approximately
0.25 metric tons of fuel per day. In this study DUPIC CANDUs were assumed to exclusively
burn AIROX-processed PWR spent fuel and thus, because of these differences in fuel
consumption, multiple PWRs were required to fuel one CANDU. Rf in Eq. (8-2) gives the
number of PWRs needed to produce sufficient spent fuel to fulfill the fueling requirements of
one CANDU. Qc and Qp are the gross thermal power and Le and Lp are the capacity factors of
the DUPIC CANDU and the PWR, respectively.
R = QBdLLC
(8-2)
As can be seen in Eq. (8-2), the number of PWRs per CANDU changes with both PWR
and CANDU burnup; thus each different PWR burnup will have a different achievable CANDU
burnup in the DUPIC cycle. Replacing Q with 1/rl (where r is thermodynamic efficiency) in Eq.
(8-2) gives the electrical power ratio (ratio of MWe) of the two reactor types in question.
Section 8.6 describes the method used to predict spent fuel production and the electricity
produced by the different fuel cycles.
The DUPIC cycle reactor ratio facilitates planning the needed reactor mix when
considering implementation and optimization of the DUPIC cycle. If a utility or country decides
to fuel all of its CANDUs with spent PWR fuel, it is a simple exercise to use the integrated
DUPIC model and Eq. (8-2) to determine other requirements for implementation of the DUPIC
cycle. For instance, if Ontario Hydro desired to fuel 8 Bruce CANDUs using U.S. PWRs,
approximately twenty-four 950 MWe PWRs with reload enrichments of 4 w/o and discharge
burnups of 44 GWD/MT would be required to fully fuel these CANDUs (based on the
characteristics of the reference reactors.) If PWR reload enrichment is fixed by a regulatory limit
or executive decision, the number of PWRs required to continuously fuel the CANDUs can be
determined. If Xp and BdL are flexible and the number of PWRs contributing to the DUPIC cycle
are fixed, Xp and BdL can be adjusted so that the PWRs produce the amount of spent fuel
required to fuel the CANDUs. This planning is constrained by the available range of reactor
ratios, as determined from the available reactor mix and characteristics.
8.4 Effects of DUPIC on Discharge Burnup
The DUPIC cycle effectively extends the burnup of PWR fuel by 10 to 20 MWD/kg. A
significant part of this burnup gain is due to the reactivity increase associated with the removal of
fission products during the AIROX process. However, the main additional reactivity gain results
from the higher neutronic efficiency of the CANDU design.
Fig. 8-1 shows the achievable burnup in the PWR and CANDU portions of the DUPIC
cycle, calculated with the integrated model as described in Chapter 7. Note that there are three
data sets for each fuel cycle; each set is differentiated by a factor that gives the PWR discharge
burnup for that data set. For instance, the 10 Xp CANDU data set gives the discharge burnup for
the CANDU stage of the DUPIC cycle where the PWR discharge burnup (in GWD/MT =
MWD/kg) was ten times the PWR reload enrichment (in w/o U-235). For a PWR using 5 w/o
enriched fuel burned to a 10 Xp discharge burnup of 50 MWD/kg, the achievable burnup in a
DUPIC CANDU is approximately 20 MWD/kg. (For comparison, the NU-fueled Bruce reactors
have a discharge burnup of 8.3 MWD/kg.)
Fig. 8-1 is significant in that it clearly shows the dependence of total DUPIC cycle
burnup on the PWR discharge burnup. Since changing PWR enrichment over the considered
Figure 8-1: Achievable Burnup in Component Stages of DUPIC Cycle
100
90
80
30
70
to7.
chainga~..n
- U-
slihtl
change ache
(5 M~W
c
" ---CANDU
"-nu- w
-----------.---------..
Xp CANDU - 0-
-
sinfcnl
----Xp CANDU
m11
MWD/kg effect on PWR), it is important to optimize burnup strategy for the PWR rather than
overall
CANDUPIC cycle
burnup while significantly
to 7 w/othus
U-235)
appears
onlytoslightly
be a ke(3
changesto
optimizing
evable
rangiven enrichment
increases
reativity
the penalty
er
burnups, as initial enrichment
that although
driving low t PWR
Fig. shows
8-1
are increasing gains in CANDU burnup and thus increasing gains in overall DUPIC burnup. At
higher initial enrichments and high (12 Xp) PWR burnup, though, achievable CANDU burnup
remains essentially constant.
Beyond approximately 5.5 w/o U-235, gains in DUPIC cycle
burnup are due strictly to gains from the PWR. For the high-burnup (12 Xp) CANDU case,
achievable CANDU burnup is actually maximized at an enrichment of 6 w/o although there is
very little decrease from this maximum as Xp is further increased.
8.5 Natural Uranium Utilization
The measure of efficiency used in most of the following comparisons is utilization.
Utilization of a resource is usually described in terms that express the effective use of the
resource. In the case of nuclear power reactors, this measure of effectiveness is defined as the
amount of electricity produced per unit mass of initial resource.
Utilization can thus be
determined by calculating the number of MWDe produced by a reactor or a cycle from a given
amount of resource, natural uranium. Utilization is given on the charts in this Chapter as a
function of reload enrichment, Xp.
Utilization was calculated for three different discharge
burnups per each Xp. Differing burnups are denoted by different line types: a discharge burnup
of 10 times reload enrichment (10 Xp) is denoted by a solid line, 11 Xp is given by a dash-dot
line, and 12 Xp by a dotted line.
The natural uranium utilization, Uu, for the DUPIC cycle can be calculated as shown in
Eq. (8-3) from values of BdL and Bdc determined by the integrated model. "f' in Eq. (8-3) is the
fraction of fissile material mass lost during PWR burnup and AIROX processing.* For the
purposes of this study, f is assumed to be zero. rp and rc in Eq. (8-3) are the PWR and CANDU
thermodynamic efficiencies, respectively.
*A simple estimate of the fraction of heavy metal fissioned during burnup is given by Bd [MWD/kg] / 1000. Heavy
metal losses during AIROX are expected to be small: less than or equal to one percent. (Ref. 15)
Uu[MWDe/kgUnat] = /BdL + (1- f)rcBdc
(F/P)PWR
(8-3)
Note that this utilization factor depends on the thermodynamic efficiency,r , of the
reactors employed in the DUPIC cycle.
This efficiency factor must be adjusted when
determining Uu for the Parallel case because, as noted earlier, this study assumes natural
uranium-fueled CANDUs to have lower thermal efficiencies. The above correlation must also be
adjusted to account for the additional natural uranium requirements of the CANDU in the
Parallel cycle. These adjustments can be made by increasing the mass flow in the denominator
by the factors shown in Eq. (8-4) where BdN is the natural uranium-fueled (NU) CANDU
discharge burnup and
1CN
is the NU CANDU thermal efficiency.
UjMWDe/kgUnat] =
Bd + (1- f))7cBdc
(F/P),
(8-4)
+(1- JBdc * 17c
It is interesting to note that Eq. 8-4 reduces to the PWR-only or CANDU-only cases if the
case-appropriate substitutions are made. For instance, in the PWR-only case, f =1 because there
is effectively a complete loss of the spent fuel between the PWR and CANDU stages. Equation
(8-4) can be modified for a CANDU-only cycle by setting BdL and (F/P)PWR equal to zero.
Equation (8-3) can also be modified to describe a parallel cycle using slightly enriched
uranium (SEU) CANDU as described in Appendix H. Appendix H also discusses the advantages
of a Parallel cycle using an SEU CANDU.
Fig. 8-2 shows the natural uranium utilization for four different cycles: DUPIC, PWRonly, CANDU-only, and Parallel. There is only one line for the CANDU-only Uu because
natural uranium CANDUs burn all of their fuel to approximately the same discharge burnup,
independent of initial PWR enrichment.
-
-
Fig. 8-2 also shows that at low to medium enrichments in the PWR, the DUPIC cycle is
even more efficient than a natural uranium-fueled CANDU, which alone is better than a PWR in
natural uranium efficiency. At low to medium ( 4 %) enrichments, the savings from fueling the
CANDU with spent PWR fuel rather than NU outweigh the PWRs' low uranium utilization.
-
Also, extending the PWR burnup at a given Xp increases DUPIC Uu even more and allows a
high-burnup (12 Xp) DUPIC cycle to achieve the same Uu as the CANDU up to 5 w/o
-
enrichment.
Figure 8-2: Natural Uranium Utilization
3.00
--DUPIC 10Xp
12Xp
- + - DUPIC.......
- U - PWR Only 11Xp
CANDU Only
---
C 2.75
D
J
S..
2.50
2 .50
-..
-.....
---...
-
-A-- Parallel 11Xp
--
- - DUPIC 11Xp
- -- PWR Only 10OXp
- --- PWR Only 12Xp
--Parallel 10OXp
- -k.-Parallel 12Xp
------------------------...
..
...
..
..
...
..
...
..
..
...
..
..
----------------..
-
2.25
0
i
i-
E
1.75
1.75
---
.
1.50
2.50
-A
-Ar
U-7
.....
"-----
.. -
-- ' . ..
---
------ -------
---
". .. .
.
3.00
3.50
4.00
4.50
5.00
5.50
6.00
6.50
7.00
7.50
Initial PWR Enrichment, Xp, w/o
PWRs have lower natural uranium utilizations because of the large amount of natural
uranium consumed during the enrichment process. With increasing enrichment, the (F/P) ratio
increases more rapidly than achievable burnup in a PWR. The resulting decrease in utilization is
evident in the downward trend of the PWR utilization curve.
The Parallel cycle is only slightly more efficient than the PWR-only cycle because of the
high reactor ratio of PWRs to CANDUs in the DUPIC cycle. With the same high ratio, the
overall Uu of the Parallel cycle is dominated by the Uu of the PWRs; the single CANDU can
only increase the overall Uu by a slight amount. This is most evident in the decreasing Uu
benefit of the Parallel cycle at high burnups.
8.6 Spent Fuel Production
The benefits of the DUPIC cycle can also be considered in terms of spent fuel production
efficiency, defined in terms of the electricity produced from one kilogram of spent fuel material.
Since the DUPIC CANDU is fueled solely with spent PWR fuel, the total volume of spent fuel
produced by the DUPIC cycle is essentially the same as that produced by the PWRs alone.
There will be some additional spent fuel waste generated in the AIROX process, but for the most
part the extra burnup in the CANDU is "free" in terms of spent fuel generation.
There is
significant additional solid waste produced during AIROX due to the PWR cladding and
structural materials but this additional waste is not included in the results shown in Fig. 8-3.
This figure compares the spent fuel utilization, Usf (in terms of MWDe/kg spent fuel), of the
four cycles. DUPIC Usf is given by Eq. (8-5) and depends only on the cycle thermal efficiency
and discharge burnup.
U,[MWDe/kg spent fuel ] = qB
+ cBdc
(8- 5)
The DUPIC cycle has the highest Usf of the four cycles. This high efficiency, greater
than even a PWR, results from burning the spent PWR fuel an additional 10-20 MWD/kg in the
CANDU reactor. It is interesting to note that the additional gain in utilization from higher
burnups in the PWR is largely offset by the decreased achievable burnup in the CANDU portion
of the cycle. This cancellation effect is noticeable when the DUPIC Usf is compared to the PWR
Usf, the next- highest-efficiency cycle. At low enrichment, approximately 2.5 MWDe/kg spent
fuel is gained by pushing PWR discharge burnup from 10 Xp to 12 Xp. At high Xp this gain is
even greater, approximately 5 MWDe/kg. The gain from doing the same in the DUPIC cycle,
however, is much smaller, only about 40% of the PWR gain at any point.
The achievable
DUPIC CANDU burnup is thus affected by the fission product build-up in high burnup fuel.
This effect will be examined below in greater depth.
Figure 8-3: Spent Fuel Utilization
I
35.00
-DUPIC 10Xp
-- DUPIC 11Xp
•--
30.00
DUPIC 12Xp
-----------------------
-E---PWR Only 10Xp
-- PWR Only 11Xp
-PWR Only 12Xp
--*
---CANDU Only
-Parallel 10Xp
A Parallel 11Xp
-...
- A- - Parallel 12Xp
25.00
20.00
. .
.
..................
- --.........
15.00
10.00
---
A-ii
t--i"
4-...
5.00
4.
p~re.?rr.....~.~..~..
I
n on -I2.50
2.50
;
:
I
3.00
3.50
4.00
4.50
---------------i-.'.-----------^~ ----
p
p
5.00
5.50
p
p
6..00
6.50
p.~.
7.00
7.50
Initial PWR Enrichment, Xp, wlo
The NU CANDU utilization is the lowest of the examined cycles, primarily because the
achievable burnup with natural uranium fuel is very low (8 MWD/kg). Again, the Usf of a NU
CANDU is constant because all CANDU fuel is discharged at approximately the same burnup,
independent of PWR enrichment. The Usf of the Parallel cycle is given by Eq. (8-6), similar in
64
form to Eq. 8-5, but again corrected for the use of an NU CANDU rather than a DUPIC
CANDU. Eq. (8-6) can also be used to calculate Usf for CANDU-only and PWR-only solutions
if a similar limiting-case procedure is followed as for Eq. (8-4).
U, [MWDe/kg spent fuel] =
FB
+ (1- f ) cBdc
*d C
BdN 1CN
(8-6)
1 + (1 - f)
There is significant gain in Usf with the Parallel cycle although its efficiency is
significantly lowered by the NU CANDU. NU CANDUs produce such large amounts of spent
fuel that just one CANDU per two or more PWRs is able to significantly lower the Usf of the
multiple PWRs used in the Parallel cycle. CANDUs, in general, produce more than 100 MT/yr
of spent fuel while PWRs generate between 34 and 15 MT/yr of spent fuel, depending on
enrichment and discharge burnup.
These comparisons are not exactly equivalent, however,
because the reactors produce different amounts of electricity and have different thermodynamic
efficiencies.
The sheer volume of waste produced by a NU CANDU can be put in better perspective
by Fig. 8-4 which shows the mass of spent fuel produced by DUPIC, CANDU-only, and PWRonly cycles over the course of a year. A Parallel cycle is not included for comparison because
there is no unambiguous value for an equivalent Parallel cycle; since the NU CANDU in the
Parallel cycle produces less electricity than the DUPIC CANDU, the remaining generation gap
can be made up with either additional NU CANDUs or PWRs. The Parallel cycle will thus
produce more spent fuel annually than the PWR-only cycle but significantly less than the
CANDU-only cycle.
Fig. 8-5 shows the DUPIC and PWR-only cycles in closer detail. Because the different
cycles use different numbers of reactors and have different efficiencies and outputs, the results in
Fig. 8-4 and Fig. 8-5 were normalized to the amount of electricity produced by the DUPIC cycle
at the same Xp. This was done by computing the yearly electrical production of a DUPIC cycle
operating at a given Xp and BdL. The number of PWRs or NU CANDUs that are required to
produce the same amount of electricity was then determined and these numbers were multiplied
by the annual spent fuel generation of each type of reactor, where the annual spent fuel
production, m, of a reactor is given by Eq. (8-7) and the annual electricity production, E, is given
by Eq. (8-8).
365.25QL
m[MT/yr] =
B
(8-7)
E[MWDe/yr] = mBry = 365.25QLy
(8-8)
Figure 8-4: Annual Spent Fuel Production Comparison of All Cycles
1200.00
---
DUPIC 10Xp
- + - DUPIC 12Xp
-*
1000.00
----
--PWR Only 11Xp
CANDU Only 10Xp
-- --DUPIC 11Xp
-
PWR Only 10Xp
-
-- PWR Only 12Xp
- - -CANDU Only 11Xp
- CANDU Only 12Xp
800.00
- - .---....
--
o--
---------------
6.50
7.00
_ 11A-
600.00
400.00
,*
"
-
-... . °i
-----------------
-o°
-
-------
i_
--
200.00
n n 2.50
2.50
3.00
3.50
4.00
4.50
5.00
5.50
PWR Reload Enrichment, Xp
6.00
7.50
To determine the annual energy production of a DUPIC cycle, the cycle's Xp and BdL
must first be chosen. Bdc can then be obtained from the integrated model described in Chapter 7.
Eq. (8-2) can now be used to determine the number of PWRs needed to fuel one CANDU and
the annual DUPIC electricity production is determined as follows in Eq. (8-9). Ep and Ec are
obtained by substituting the appropriate PWR and DUPIC CANDU values into Eq. (8-8).
(8-9)
ED[MWDe/yr] = EpRf + E c
Figure 8-5: Annual Spent Fuel Production for DUPIC and PWR-only Cycles
140.00 -e
---.--
CANDU Only spent fuel production varies
between 430 MT/yr for 3 w/o, 10Xp case to
1000 MT/yr for 7 w/o, 20Xp case.
DUPIC 10Xp
- DUPIC 11Xp
- DUPIC 12Xp
--in--PWROnly i10Xp
20.00 1
- --
.-
PWR Only 11Xp
- PWR Only 12Xp
00.00 1
-
-
------- --
80.00
60.00
-
--
-
-
----
--
I----------------- -------------
---------
------------0
----------------------------------- --------------40.00
2.50
*
3.00
3.50
4.00
5.00
4.50
5.50
6.00
6.50
7.00
7.50
Initial PWR Enrichment, Xp, w/o
The number of reactors, N, that are required to produce the same amount of electrical
power as the DUPIC cycle in a CANDU-only or PWR-only cycle is now calculated by dividing
Eq. (8-9) by Eq. (8-8) for the appropriate CANDU-only or PWR-only case. Finally, the annual
spent fuel generation of any fuel cycle can be calculated by multiplying the reactor-specific
67
results of Eq. (8-7) by the applicable reactor mix for any cycle. For instance, for a PWR-only
cycle, Eq. (8-10) gives the required number of PWRs, Np and Eq. (8-11) gives the annual spent
fuel production of this PWR-only cycle.
Np -
E EEpR, +Ec
=
E
(8-10)
E+
Mp [MT spent fuel/yr] = mNp = mp,
pfEc
(811)
It is interesting to note that there appears to be less gain from increasing enrichment (and
thus burnup) in the high-burnup (12 Xp) DUPIC case. At high enrichments this curve is much
more level than the lower-bumup cases and M does not continuously decrease as enrichment
increases. There is actually a minimum in the 12 Xp DUPIC curve, indicating that the optimum
enrichment, in terms of spent fuel generation, is approximately 6 w/o. The increasing spent fuel
production at the high-enrichment end of this curve shows that the effects of fission product
buildup on CANDU reactivity become very significant at high PWR bumup. For both the 10 Xp
and 11 Xp DUPIC cases the effects of fission product buildup are not as pronounced and do not
limit additional CANDU burnup.
8.7
Conclusions
The DUPIC cycle is more efficient than all other examined cycles (PWR-only, CANDU-
only, and Parallel) in terms of spent fuel minimization. The PWR portion of the cycle is the
dominant contributor to spent fuel efficiency; extending burnup of the spent PWR fuel in the
CANDU only contributes a small, virtually enrichment-independent gain to the DUPIC cycle. In
terms of uranium utilization, the DUPIC cycle is more efficient than the CANDU-only cycle up
to reload enrichments of 5 w/o. At high burnups in the DUPIC cycle, CANDU discharge burnup
is essentially independent of enrichment above 5 w/o. Because the PWR is dominant in terms of
burnup contribution and the elasticity of CANDU burnup in terms of original PWR enrichment is
relatively small, DUPIC cycle characteristics should be chosen that optimize PWR performance.
CHAPTER 9
DUPIC Cycle Implementation in North America
As the nuclear industry has matured, different methods have been proposed to enhance the
environmental and economic performance of nuclear power. The DUPIC cycle has the potential
to do both by decreasing the amount of spent nuclear fuel that is produced and by producing
more electricity from a given amount of uranium. The proximity of U.S. LWRs to Canadian
CANDUs in North America creates a potential partnership that can implement the DUPIC cycle.
While there are numerous advantages to implementation of a DUPIC cycle between the U.S. and
Canada, there are also significant barriers to that implementation that must be addressed.
Because this concept is in its infancy, there is little concrete evidence as to the future path that a
proposal for implementation would have to travel.
It should be possible to extrapolate the
general nature of the path, however, from the historical treatment of issues relating to nuclear
power.
9.1 History and Background of the DUPIC Cycle
9.1.1
History
The DUPIC cycle was originally proposed by Korea in 1989 and accepted by the U.S. in
1990. Korea (KAERI), Canada (AECL), and the U.S. (Los Alamos National Laboratory) began
a feasibility study in 1991 that continued for a year before an experimental program was initiated
in 1992. This program is currently ongoing and is divided among the parties such that Canada is
in charge of designing, irradiating, and testing DUPIC fuel pins and bundles. KAERI's role is to
prepare the facilities to implement a test-scale DUPIC cycle using spent PWR fuel from Korean
reactors and Korean facilities to process and refabricate the DUPIC fuel. The U.S. has been
examining and developing materials safeguards procedures designed to prevent nuclear weapons
proliferation.
9.1.2
Motivation for DUPIC Cycle Development
The DUPIC cycle was originally proposed by Korea as a means of reducing waste
generation and decreasing natural uranium consumption.
Korea is a small, site-limited, and
resource-poor country and appreciates the possibility of decreasing dependence on foreign
sources for natural uranium and space requirements for spent fuel disposal. While the above
three issues have been motivators for consideration of the DUPIC cycle in the U.S., an important
additional motivator has been the prospect of decreased fuel cycle costs and finding a method of
recycling spent fuel that has improved proliferation resistance over the wet processing techniques
currently in use.
9.1.3
Current Status of DUPIC Implementation
Korea, Canada, and the U.S. are continuing research into implementation of the DUPIC
cycle in Korea. AECL is beginning irradiation of three DUPIC fuel pins and the first tests of
larger DUPIC fuel samples are expected to occur in Korea in 2000. (Ref. 27) There are no
formally published studies of potential U.S. - Canadian implementation of the DUPIC cycle.
Some individuals are attempting to initiate discussions with U.S. and Canadian utilities regarding
the potential of the DUPIC cycle, but no official (i.e. governmental or even utility) decisions
have been made to examine the feasibility and advantages of the DUPIC cycle.
9.2 Issues Surrounding Implementation of the DUPIC Cycle
Implementation of the DUPIC cycle will affect many areas of nuclear energy production
and thereby national policy-making. The DUPIC cycle may lower fuel cycle costs and thus the
costs of generating electricity and disposing of spent fuel, thereby making nuclear generating
stations more competitive.
A more competitive nuclear industry will affect national policy
regarding electricity deregulation, radioactive waste disposal, nuclear energy regulation, and
strategic resource planning. Any resurgence or reorientation of the nuclear industry will attract
the interest and concern of many public and special interest groups.
9.2.1
Reduced Fuel Cycle Costs*
Both U.S. and Canadian operators can gain significantly from implementation of the
DUPIC cycle. Table 9-1 summarizes the cost and savings estimates that are discussed below in
further detail. The cost estimates are primarily from (Ref 25, 26, 27, and 31). Note, however,
that many of the unit costs in question vary considerably among different analysts: See (Ref. 29)
for a definitive survey of international studies.
Table 9-1: Summary of PWR, CANDU, and DUPIC Fuel Cycle Costs and Savings (1)
Process
Uranium Fuel Procurement
Fuel Fabrication
Spent Fuel Storage
Spent Fuel Disposal
Total
PWR Savings
(4)
[$/kg]
175
370
545
CANDU Savings
(3)
DUPIC Cost
Net Savings (2)
[$/kg
[$/kg]
[$/kg]
151
113
20
98
382
510
175
200
885
151
-397
20
268
42
Notes:
(1) Basis of estimates is $/kg of spent reference DUPIC fuel, Chapter 2.
(2) Savings are determined by subtracting DUPIC Cost from the sum of PWR and CANDU Savings.
(3) CANDU savings represent the dollars saved by displacing NU fuel and are adjusted to account for greater fuel
consumption and spent fuel production when fueled with NU as opposed to spent PWR fuel. The general
formula is CANDU Savings = Rsf * NU CANDU value.
* Because DUPIC cycle implementation in the U.S. and Canada has not been examined in significant detail, there
are few known reliable economic analyses on the lifecycle costs and benefits of DUPIC implementation in the U.S.
Many economic references this study cites have been developed for Korea while very few have been developed for
Canada. For the purposes of this study, estimates developed for Korea will be used when necessary for the U.S. and
Canada. Actual costs in the U.S. in Canada can be expected to vary in a variety of ways and due to a variety of
factors. There is one recent study (Ref. 31) of spent fuel disposal costs in Canada. It only examines direct
repository costs, however, and does not include transportation and storage costs. It is one of the present study's
recommendations for further work that a complete, life cycle economic analysis be completed of DUPIC cycle
implementation in the U.S. and Canada.
(4) PWR savings represent the back-end costs avoided by sending spent fuel from NR PWR units to one CANDU
U.S. operators and the DOE will save primarily on spent fuel storage and disposal costs
while Canadian operators, by using spent PWR fuel to fuel the CANDUs, will not have to buy
uranium ore. The U.S. Department of Energy currently assesses utilities a 1 mill/kwhr fee that is
earmarked for spent fuel disposal costs.
For the reference PWR with fuel burned to 45
MWD/kg, this fee translates to a disposal cost of $370/kg of spent fuel. Provided that Canadian
utilities take responsibility for disposing of the spent DUPIC fuel, it seems reasonable to expect
that U.S. utilities or the Department of Energy will recognize most of these spent fuel disposal
fees as savings.
This fee also does not account for the cost that utilities incur from storing spent fuel on
the premises because it currently cannot be shipped to a spent fuel repository and spent fuel pool
space is exhausted, requiring the purchase of dry storage casks. Some utilities have been storing
fuel on-site for more than 20 years and some decommissioned power plants have even had to
ship their spent fuel to other reactor sites for storage. One estimate of PWR spent fuel storage
costs is as high as $260/kg. (Ref 26) Since spent PWR fuel can be expected to be stored for
some amount of time before AIROX processing, U.S. utilities should not expect to save all of
these costs.
This study conservatively (i.e. high end) estimates that complete DUPIC fuel
storage (PWR fuel prior to processing and CANDU DUPIC fuel after irradiation) may cost
$175/kg. Storage savings, when combined with PWR spent fuel disposal savings, might allow
U.S. utilities to save as much as $545/kg from implementation of the DUPIC cycle.
While the Canadian utilities will still have to dispose of the DUPIC spent fuel, they will
have to dispose of a much smaller volume of waste. A significant portion of actual spent fuel
disposal costs will be proportional to fuel volume; since the DUPIC cycle can be expected to
produce approximately half the waste of the equivalent NU CANDU cycle, there should be
significant spent fuel disposal savings for Canadian utilities as well. For instance, the ratio of
NU CANDU spent fuel production to DUPIC CANDU spent fuel production normalized per unit
of electrical power generation, Rsf, is given by Eq. (9-1).
R =B
-
BdC7C
(9-1)
Applying Eq. (9-1) to the conditions of the reference case would give a spent fuel
production ratio of 1.89; this is the ratio used in Table 9-1 to adjust the CANDU cost estimates.
Canadian operators would enjoy additional savings because they would not have to
purchase and fabricate NU fuel for their CANDU reactors.
Canadian studies estimate fuel
procurement and fabrication costs to be approximately $140/kg (Ref. 25). NU CANDU spent
fuel storage is estimated to cost $10/kg in Korea (Ref. 26), but a figure of at least twice that
($20/kg) seems more realistic and is used in this study. NU spent fuel disposal is estimated to
cost $52/kg in Canada while DUPIC disposal is forecast to cost about $200/kg for the reference
case. (Ref. 31) When the savings from spent fuel production are added to the avoided additional
costs of using of NU fuel, Canadian savings from implementation of the DUPIC cycle can be
estimated to be greater than $300/kg. For example, in the reference case, where every kilogram
of DUPIC fuel reduces spent fuel production by 0.89 kg, the total savings from DUPIC
implementation may be expected to approach 1.89*$140/kg (fresh NU fuel savings) +
0.89*$52/kg (avoided spent fuel savings) a $350/kg. If the NU spent fuel disposal costs are
adjusted upwards, more in line with estimates from the rest of the world (for instance, disposal is
expected to cost $400/kg in Sweden and Finland), total Canadian savings could be expected to
exceed $600/kg.
Combining the savings in the U.S. and Canada would increase the total savings in the two
countries to approximately $900-1000/kg.* Even greater savings could be projected if estimates
by U.S. nuclear power critics were used: One critic estimates that the real eventual cost of LWR
fuel disposal will be 3 mills/kwhr or $1050/kg of spent fuel in the reference case. (Ref 30)
9.2.2
AIROX Plant Costs and Issues
The aforementioned economic savings forecasts do not account for the added cost of the
AIROX processing plant. A 1996 cost evaluation of a conceptual AIROX facility estimated that
a plant capable of processing 400 MT/yr of PWR spent fuel would cost $1.1 billion with a
levelized unit (of spent fuel) cost of $510/kg. (Ref 15) This plant is capable of processing the
spent fuel from 17.5 reference PWRs and fabricating fuel for 5.4 CANDUs using Eq. (8-2) and
(8-7) as described in Chapter 8. When the costs of the DUPIC fuel fabrication is subtracted from
net U.S. and Canadian savings as detailed in Table 9-1, the net savings from implementation of
the DUPIC cycle can be estimated to be about $40/kg of spent fuel.
This estimate is extremely variable, however, due to the uncertainties regarding plant
construction costs. Estimates of the cost of the plant are extremely preliminary and were given
with an error range of +50% to -30%. (Ref. 15) Part of the wide range in cost estimates is
because this plant will be the first of its kind utilizing unique processes, a condition that makes
an accurate estimate extremely difficult. This plant will potentially be subject to the extreme
price escalation seen previously in nuclear power plant construction. Additionally, the licensing
*This savings estimate assumes DUPIC spent fuel disposal costs to be approximately 0.5 mills/kwhr. These costs
include cooling of the spent DUPIC fuel for 50 years. This substantial cooling period significantly reduces costs,
making it, on paper, less expensive (in constant dollar terms) to dispose of DUPIC fuel than NU fuel. Comparison
between the cited NU and DUPIC spent fuel disposal costs, as cited, is not quite accurate. The DUPIC fuel will
have a higher storage cost, though, and it is hoped that this higher cost may make the comparison more accurate.
and litigation process will also be complex and likely costly due to the groundbreaking nature of
this plant.
The location of the AIROX plant will also affect construction costs due to differing labor
and materials costs as well as other factors such as government support and public opposition (or
lack thereof). It will probably be less expensive to construct an AIROX plant in Canada than in
the U.S., although it may be difficult to convince Canada to both dispose of the spent fuel and
build the AIROX plant, despite the fact that such a plant would most likely be subject to
substantial litigation in the U.S. Regardless of location, though, it is also possible that the plant
costs in the U.S. or Canada would be higher than the estimated cost in Korea. Hence, the net
savings given in Table 7-1 from a DUPIC cycle are to be considered extremely speculative
estimates.
The economic savings of the DUPIC cycle are very dependent on the AIROX plant costs
and the Canadian spent fuel disposal costs.
Most current consideration of DUPIC
implementation in North America predicts the construction of the AIROX plant and DUPIC
spent fuel disposal in Canada. Negotiations regarding these two issues are likely to be the most
contentious in the case of DUPIC implementation between any two countries. Given the current
problems of spent fuel disposal in the U.S., it seems that disposal of spent DUPIC fuel in Canada
is the more feasible option. The Canadian estimate of $200/kg for DUPIC spent fuel disposal
seems low compared to estimates for disposal of different types of spent fuel in other countries,
especially since DUPIC fuel has a high fission product inventory.+
+Fission
product inventory is approximately proportional to total burnup. DUPIC fuel will thus contain about 8
times the fission product mass of NU CANDU fuel. However, the volume of CANDU spent fuel that is produced
will be about twice that of DUPIC spent fuel.
9.2.3
Deregulation and Competitive Improvements
The electric power industry in the U.S. is currently being deregulated to increase
competition. One of the most-contested areas of deregulation involves capital investments that
utilities made expecting to recover the cost in their service contracts, so-called "stranded costs."
Many utilities have nuclear power plants that, because of cost escalation during construction or
operating problems, have very high capital costs that have not yet been repaid by the utilities'
captive ratepayers. Had the electricity market continued to be regulated, these utilities would
have eventually recovered their capital investment in these plants through regulated electricity
rates set specifically to repay the utilities for their expenditures. However, as electricity markets
are deregulated and lower-cost independent generators are allowed to compete with the utilities
for customers it is expected that the market price of electricity will fall below the minimum level
needed by utilities to allow them to completely recover their capital investments in the nuclear
power plants. Nuclear plants (or other facilities) where only part of the initial investment can be
recovered are called stranded assets.*
Stranded cost recovery is the policy that compensates utilities for the unrecovered portion
of the investment. Utilities claim they are entitled to stranded cost recovery because they were
assured under past regulations to recover their capital investment over the life of the plant.
Utilities then claim that, because they were encouraged to believe that the nuclear plants could be
amortized over 30 or more years, they should be allowed to pay for the stranded assets over a
long period of time and these costs should be passed on to their former ratepayers.
As deregulation progresses, the U.S. states have been inclined to approve stranded cost
recovery through electricity surcharges to all customers for several years into the future. The
amount of the surcharge and length of time of recovery vary from state to state. If utilities do, in
the end, save money through implementation of the DUPIC cycle, the allowed amount of
stranded cost recovery should be adjusted. In general, the portion of stranded costs attributable
to on-site spent fuel storage should only be a small fraction of the total, although there will be
additional savings if the DOE surcharge is removed from spent PWR fuel that is used in the
DUPIC cycle.
9.2.4
Spent Fuel Reduction
While spent fuel issues are inextricably tied to economic issues, the above discussion
dealt with spent fuel savings in dollar terms as a part of overall fuel cycle costs. There are other,
more qualitative issues regarding spent fuel production and savings that also need to be
addressed including the timing and feasibility of spent fuel repository disposal and the
subsequent inaccuracy of spent fuel disposal costs.
The estimates for spent fuel disposal costs discussed in Section 9.1.1 are based on the
eventual construction and use of a spent fuel repository within budget and time constraints. The
fate of the repository program has always been somewhat uncertain, however, because of the
very politically sensitive nature of the decisions that have been and will be made.
The
framework for the establishment of a geologic repository for spent fuel disposal was first created
by the Nuclear Waste Policy Act of 1982. The original act called for the opening of the first
repository by January 1998 and the opening of a second repository in the 2001 timeframe. The 1
mill/kwhr disposal fee (with provisions for annual review) was also established in the 1982 act.
In 1987 the Nuclear Waste Amendments Act designated Yucca Mountain, Nevada as the
repository site and provided for the establishment of an interim storage site outside of Nevada.
Since the 1987 act, Nevada has obstructed the establishment of a repository to the extent that a
Supreme Court order was required to begin characterization of the Yucca Mountain site.
Because of this and other delays, the opening of the Yucca Mountain repository has been delayed
at least until 2010. The interim waste storage facility siting provision in the 1987 act has been
reversed; any interim storage site that is established will likely be in Nevada, close to the Yucca
Mountain site. The site characterization of Yucca Mountain, originally estimated to cost $100
million, will cost more than $6 billion to complete. The licensing of the repository will cost an
additional $1-2 billion after the NRC research program and regulatory expenses are added up and
before any repository construction begins.
When the history of the repository program is considered, serious questions regarding its
feasibility and likelihood of implementation are raised. Given the escalation of cost estimates,
questions are also raised regarding the adequacy of the 1 mill/kwhr fee to completely pay for the
lifecycle repository costs. (Ref. 30) Since the future of the repository program in regards to cost
and even existence are so questionable, any evaluator should be very critical of using economics
as the sole criterion when examining the advantages of the DUPIC cycle. Indeed, since it is not
definitely known how we will dispose of the spent fuel we already have, it is clear that
minimization of spent fuel production is an important consideration in the evaluation of any
nuclear fuel cycle policy. The fact that, as discussed in Chapter 8 and shown in Fig. 8-4, the
DUPIC cycle produces significantly less spent fuel per unit electricity than CANDU-only cycles
and moderately smaller amounts than a PWR-only cycle should make the DUPIC very attractive
regardless of most economic considerations.
9.2.5
DUPIC Fuel Disposal
It is important to consider the issues surrounding the choice of country for final disposal
of DUPIC fuel. It is likely that there will be significant political discussion of this issue. In the
case of a regional DUPIC cycle (between different countries), it is in both countries' interests to
dispose of the DUPIC fuel in the other country: no country wants to be the final resting site of
more waste than is necessary, whether or not the waste is nuclear.
It is thus important to
consider the potential savings from DUPIC implementation in terms of country-dependent
benefits (and costs).
For instance, if DUPIC fuel is disposed of in Canada, the benefit to the U.S. could be
very significant. It is expected that future spent fuel production in the U.S. will exceed the
planned capacity of the Yucca Mountain repository; implementation of the DUPIC cycle might
reduce future spent fuel volumes such that a second repository is not required. There would be
significant savings in the U.S. associated with avoiding the construction of a second repository.
Congress did not have the political will to site a second repository in the 1980s and it is unknown
if this situation has changed. Indecisiveness or fragmented support would most likely slow
project completion. This can be seen in the history of the current Yucca Mountain repository:
The resultant project delays and increased litigation have drastically escalated costs. There is no
reason to believe that the political environment in the U.S. has significantly changed and there is
thus no guarantee that the second repository could actually be created if needed. Utilities would
then be on their own to dispose of the spent fuel or the federal government would have to store
the fuel aboveground indefinitely; either option will be costly.
Canada would benefit from DUPIC cycle implementation with spent fuel disposal in the
U.S. in two ways. Not only would Canada not have to pay spent fuel disposal fees but their
spent fuel transport and storage costs should be lower under DUPIC due to the smaller volumes
of spent fuel produced. Even if the spent DUPIC fuel is disposed of in Canada, Canada will
benefit because the spent fuel volume generated by their reactors will be lowered by half. If the
U.S. must dispose of the DUPIC spent fuel, the DUPIC fuel should be no more expensive to
dispose of than spent PWR fuel.
One recent study (Ref 31) indicated that spent PWR and
DUPIC fuel have similar decay heats over time, partially due to fission product removal during
AIROX processing and partially due to increased plutonium and actinide destruction
(particularly Pu-239, Pu-241, and Am-241) during burnup in the CANDU (Ref 31).
Recent news indicates that Canada may have as much trouble disposing of their spent
fuel as the U.S., though. Canadian utilities were expecting a decision "regarding the future
direction of used fuel disposal" in 1998, but in February the government instead announced that
another panel would be appointed to fully reconsider the issues involved in siting a spent fuel
repository (Ref. 32). Due to failures in the original analysis to adequately consider the social
issues of spent fuel disposal, the Canadian government created a new agency to study the issue
and announced that the decision to build a spent fuel repository and begin construction was at
least 20-25 years away (Ref. 32). A repository delay such as this should encourage the Canadian
government to seek means of reducing the amount of spent fuel that must eventually be stored.
Should either country be able to permanently dispose of spent fuel, the different political
climates of the two countries will affect DUPIC fuel disposal costs and this should also be
considered in determining the siting of DUPIC fuel disposal. Due to different Canadian political
and geographical qualities, the costs of disposing of a given amount of spent DUPIC fuel in
Canada might not be as large in the U.S. For instance, a recent Canadian study (Ref. 31) expects
DUPIC fuel disposal to cost only $200/kg once the spent DUPIC fuel is cooled for 50 years
aboveground. Disposal of the same fuel in the U.S., if subject to the 1 mill/kwhr surcharge,
would cost $400/kg based on total accumulated burnup.
It thus appears particularly
advantageous to dispose of DUPIC fuel in Canada because their costs are lower and they are
planning on a lengthy cooling period for spent DUPIC fuel which should allow sufficient time
for the construction of a repository.
9.2.6
Conservation of Strategic Resources and National Security
The prospect of conserving natural and strategic resources is another motivator for
implementation of the DUPIC cycle.
As shown in Chapter 8, the DUPIC cycle has better
uranium utilization than both PWR and CANDU cycles up to 4 or 5 w/o initial PWR fuel
enrichments. Above 5 w/o, the DUPIC cycle is still more efficient than the PWR fuel cycle. In
general, conservation of any resource is desirable politically and strategically. A reduction in the
required amount of uranium ore reduces the destruction and disruption of the environment that
occurs during mining of uranium ore.
Uranium reserves are of great potential importance to national energy security because of
the demands that will be placed on the world energy supply in the future. While the shortages of
low-cost energy that have been forecast for the past 40 years have yet to, and may never,
materialize, the continued population growth and industrialization of less-developed countries
will require ever-increasing amounts of energy. While uranium, when used in breeder reactors
and recycled, can almost be considered a renewable resource, deployment of this system will be
costly and contentious. The uranium saved from early DUPIC cycle implementation could thus
postpone the time when the existing light water reactor (LWR) infrastructure becomes
outmoded. A complicating factor of this analysis is that Canada is the world's leading supplier
of uranium and the U.S. now imports most of its uranium from Canada (Ref. 33).
Implementation of the DUPIC cycle would result in a significant reduction in the amount of
uranium mined in Canada, an outcome that the uranium miners would not support.
Because the DUPIC cycle consumes significant amounts of plutonium during the
CANDU stage of the cycle (as mentioned above and in Ref 31), the DUPIC cycle will produce
spent fuel that is even more proliferation-resistant than PWR spent fuel. This is an advantage
because there will be much less temptation to excavate a DUPIC repository in an attempt to
recover weapons material. The use of the DUPIC cycle will make the repository more secure
and increase the safety of all countries from groups attempting to obtain weapons material to
make a nuclear weapon.
9.2.7
AIROX Processing and Proliferation Resistance
One of the primary concerns and goals of the U.S. since the 1970s, with respect to any
nuclear issue, has been non-proliferation. The U.S. Department of State is participating in a joint
DUPIC development project with Korea and Canada because they believe that the AIROX-cycle
meets the U.S. requirements for proliferation resistance.
The main reasons for the AIROX
cycle's U.S. support derive from the use of dry-processing rather than the "wet" processing that
is used in fuel recycling plants in France and the U.K.
The U.S. supported uranium recycle in LWRs and was also developing breeder reactors
until President Carter implemented the U.S.'s current stand against these fuel cycles during his
term in office. While President Reagan rescinded the official order forbidding reprocessing,
since the Carter administration many in the U.S. have been opposed to the construction of any
sort of "wet" recycling plant due to of the near-complete separation of uranium and plutonium
isotopes from the fuel material that is done during wet reprocessing. Additionally, utilities have
come to view reprocessing as uneconomical and have not pursued a recycling strategy.
In the U.S. view, the near-complete separation of isotopes prior to the preparation of
mixed oxide fuel is thought to facilitate weapons material diversion and thus too risky in terms of
proliferation, even though France, Japan, and the U.K. are all actively pursuing reprocessing
programs. The quite recent testing of nuclear weapons in India and testing threats by Pakistan
will almost surely harden opinion against wet reprocessing and any treatments not perceived as
"proliferation-resistant." The use of a dry recycle during AIROX does not involve the separation
of uranium and plutonium isotopes from the rest of the fuel material. Additionally, during the
AIROX spent fuel processing the gamma-emitting fission products continue to provide selfprotection to the spent fuel. Also, AIROX processing is done in heavily shielded enclosures
called hot boxes that prevent contamination of the plant from released fission products. These
hotboxes will all be monitored by safeguard systems developed by the U.S. It would be very
difficult to covertly remove any spent fuel material from a hotbox even if the hotboxes weren't
monitored.
The most significant disadvantage of the AIROX process, as with any recycling process,
is that radioactive waste is generated during the process. All of the removed fission products
discussed in Chapter 6 in addition to the PWR spent fuel cladding and structural materials must
be treated and disposed of as nuclear waste. While some of the removed fission products can be
stored at the AIROX plant until they decay to safe radioactivity levels, other fission products,
such as iodine, are captured in a solid medium and then disposed of similar to other solid waste.
AIROX processing, as opposed to wet reprocessing, is characterized by low-level waste streams
that are small in volume. And while there is still a large net decrease in waste generated by the
DUPIC cycle, even including waste generated by the AIROX process, difficulties with public
citizens and environmental groups may still arise simply because new waste is created during the
DUPIC cycle.
9.2.8
Transportation
Transportation of DUPIC fuel between the various stages of the cycle is also an issue.
Transportation of spent fuel must be relatively easy and very safe otherwise the implementation
of the cycle is threatened. Fortunately, there is already substantial experience in this area as
spent fuel is currently transported between reactor sites; more experience will also be gained as
spent fuel is shipped in the coming decade to an interim storage site which may or may not be at
Yucca Mountain.
While current spent fuel transportation methods are extremely safe and the
procedure is well-established there is the possibility that transportation schedules can be affected
by litigation and public protest.
Incidents of public intervention and delay of spent or
reprocessed fuel shipments have recently occurred in Germany and Japan. Interruptions in the
flow of spent fuel to the AIROX plant have the potential to drastically increase the lifecycle costs
of DUPIC cycle implementation.
9.3 Analyst's Perspective and Policy Options
9.3.1
Analyst's Perspective
The potential qualitative gains of DUPIC cycle implementation, particularly in an region
such as North America where there are significant numbers of both types of reactors in a good
ratio (4-5 LWRs to each CANDU), far outweigh the possible negative consequences of cycle
implementation (Ref. 1) There is the potential for large reductions in Canadian spent fuel
volume, the conservation of significant amounts of natural uranium, and the avoided disposal of
large amounts of spent fuel in the U.S. Two possible drawbacks or barriers are higher than
expected AIROX costs that negate all potential savings from the cycle and a potential spent fuel
transportation accident. The former is significantly more likely than the latter, and the latter is
not a significant threat in terms of radiation release. The containers used to transport spent fuel
have been conservatively designed and tested to withstand even train crashes. Additionally,
DUPIC fuel will always be protected and contained by cladding during transport to and from the
AIROX plant. The potential risk from most any transportation accident is probably smaller than
most people incur when they commute in their cars to work.
There would be significant regulatory and pubic opinion barriers that would need to be
overcome in order to implement the DUPIC cycle between the U.S. and Canada, but it is the
author's belief that many of these barriers might be overcome with complete and open education
of the public as to the benefits, disadvantages, and risks associated with the DUPIC cycle. All
parties will have different values and interests; it is very likely that some parties will decide that
their interests and values are not served by implementation of the DUPIC cycle. If a significant
part of the populace decides that DUPIC cycle implementation is in their best interests, however,
it is very likely that partisan opposition lacking broad-based support could be overcome.
9.3.2 Policy Options
The DUPIC cycle offers the potential for some reduction in spent fuel production and
increase in uranium utilization over the fuel cycles currently used, whether implemented in a
country or region. The potential gains from the DUPIC cycle are far more dramatic in a regional
implementation, such as between the U.S. and Canada or Brazil and Argentina.
These are
considered regional implementations because each country is operating on a different fuel cycle
and can make great improvements in their fuel cycles, either by avoiding spent fuel disposal
costs (the country with PWRs) or by avoiding fuel purchase costs and significantly decreasing
generated spent fuel volumes (the country with CANDUs).
In an intra-country DUPIC
implementation, such as is being planned in Korea, there are still significant savings, but fresh
PWR fuel must still be purchased and DUPIC spent fuel must still be stored and disposed of in
some manner.
However, the DUPIC cycle is by no means the only and not necessarily even the most
effective way of addressing some of the issues raised in Section 9.2. There are numerous options
in addition to DUPIC cycle implementation that can be pursued to solve the spent fuel disposal
problems in the current U.S. LWR-based fuel cycle. The first option would be to fast-track the
establishment of the Yucca Mountain repository and increase the planned capacity or even
establish a second repository.
Another possible solution would involve the use of a self-
sustaining fuel cycle. Evaluation of multiple recycle LWR fuel cycles based on the AIROX or
other non-aqueous processes is currently being done at MIT; this type of innovative fuel cycle
may prove to be both feasible and easily implementable.
National energy security can be improved without DUPIC implementation by pursuing
energy conservation and renewable energy programs.
Conservation will reduce total energy
demand and thus demand on foreign energy sources and an increase in energy production from
renewable resources would do the same. Increased use of renewable energy sources will also
enhance national security because this use will save other domestic resources for use in the
future. Neither of these options is likely to provide sufficient electricity, either through savings
or increased production, to eliminate the need for expansion of the energy supply from fossil
and/or nuclear sources, however.
There are also other options for effective implementation of the DUPIC cycle. Many
DUPIC studies assume that the associated AIROX plant will be located in a region or country
that is implementing the cycle and that the plant will be used exclusively by the region or
country that builds the plant. It is possible that a non-associated AIROX plant could be built that
would buy spent PWR fuel from utilities in one country and then sell AIROX-processed
CANFLEX fuel to other countries with CANDU reactors. The exact cash flows associated with
"purchases" and "flows" will depend on the cycle economics; an electricity generator may
actually pay the plant owner to accept the generator's spent fuel in order to avoid spent fuel
disposal costs.
Other synergistic applications of an AIROX plant are possible. For example, the AIROX
process could be used for refabrication of accelerator targets used for "electric breeding." This
process uses a proton accelerator to power a spallation neutron source that transmutes a fissile
isotope (U-238 or Th-232) into fissile isotopes (Pu-239 or U-233) for LWR fueling. Weapons
grade plutonium could also be blended into batches of spent PWR fuel during AIROX
processing. This could be used to increase the reactivity of spent PWR fuel that has been cooled
for significant periods of time, permitting its use as DUPIC fuel and thereby decreasing existing
volumes of spent fuel.
This process would thus also help the U.S. decrease its plutonium
stockpiles and could even save the multi-billion dollar cost of a special purpose facility that
would be constructed solely for purpose of disposing of these stocks of plutonium.
9.4 Description of Stakeholders and Decision Makers
A significant number of residents (acting as electricity consumers) in the United States
and Canada could be affected by implementation of the DUPIC cycle between these two
countries, particularly if there are savings to utilities and changes to stranded cost recovery
policies.
Other groups that will be affected by DUPIC implementation are electric utilities and
nuclear power plant operators and taxpayers. The decision makers involved in implementation
of a DUPIC cycle include Congress, executive agencies such as the President and Department of
Energy, and federal agencies such as the Nuclear Regulatory Commission, Environmental
Protection Agency, and Department of Energy. Environmental and public interest groups can be
expected to have significant influence during DUPIC policy negotiations.
9.4.1
Public Positions
The public can fill three separate and independent roles at the same time: ratepayer,
taxpayer, and concerned citizen. As a ratepayer, the public should generally be pleased with any
decrease in electricity rates and would be likely to support the DUPIC cycle on a purely
economic basis if implementation will produce net savings for the utilities and these savings are
passed along to the consumer. As a taxpayer, the public is also likely to support the DUPIC
cycle. As discussed earlier, with the significant delays and uncertain future of the repository
project, it is very possible that storage and repository costs will exceed forecasts and may exceed
the proceeds from the 1 mill/kwhr charge to utilities. If this actually happens, some combination
of utility and government funding will have to pay for the budget overruns.
Either as a
ratepayer, taxpayer, or both, the public will have to fund spent fuel disposal. If implementation
of the DUPIC cycle reduces even a portion of these cost overruns then the public will again gain
from implementation of the DUPIC Cycle.
In the concerned citizen role, the public can take a variety of stands regarding DUPIC
cycle implementation. In an absolutist, anti-nuclear role, the public could oppose any change to
the existing nuclear fuel cycle, especially since the new AIROX plant would produce nuclear
waste. The spent fuel transportation requirements would also be cause to oppose the DUPIC
cycle because of the threat of accidents and radiation releases. On the other hand, as mothers
and fathers concerned about the future state of the world, it is possible that the public will decide
to support the AIROX plant. This might be because the public recognizes that there will be a net
decrease in spent fuel production and that the transportation issue is a moot point. (Spent PWR
fuel will eventually have to be transported to a repository or storage site if it is not recycled;
CANDU fuel bundles must originate at either the AIROX plant or another fuel fabrication plant.)
Additionally, the public could recognize that their risk from a fuel transportation accident is
extremely minimal because of the margins of safety inherent in the transportation containers.
Because of the extreme uncertainty associated with predicting the public's eventual views
on the DUPIC cycle, and because of the perceived general negative public sentiment toward
nuclear power, education will be essential to helping the public make an informed, rather than
irrational, decision that hopefully supports implementation of the DUPIC cycle. Unfortunately,
this type of dialogue has never been successfully created with the public (thus creating the
general fear of and negative sentiment towards nuclear power and nuclear waste).
If
implementation of the DUPIC cycle in the U.S. and Canada is to succeed the public in both
countries must not be opposed to DUPIC cycle implementation.
While gaining the public's
support of DUPIC implementation will not guarantee implementation, alienating the public will
certainly prevent DUPIC cycle implementation because of the political nature of much long-term
decision making in the U.S.
9.4.2
Utility Positions
It is likely that most nuclear generators will support DUPIC implementation. Some
utilities may be deterred from attempting implementation by the substantial litigation that might
arise from interest groups seeking to block implementation. There are other utilities, however,
that would probably vigorously pursue the creation of a U.S.-Canada DUPIC cycle.
As
mentioned above, the difficulties and delays associated with the federal waste repository have
raised significant doubts as to the likelihood of its opening within the next twenty years, or ever.
Although on-site dry storage is possible for older spent fuel, this option requires additional
licensing, litigation, and money. In order to avoid dry storage and further complications with the
federal repository it is extremely likely that there will be sufficient U.S. utility interest to pursue
implementation of a joint U.S.-Canadian DUPIC cycle. Additionally, many of the operators
whose spent fuel stores are approaching the storage capacity of spent fuel pools are in the
northeast. (Ref. 27) This would facilitate implementation because transportation distances to
Canada are shorter.
9.4.3
Governmental Positions
It is difficult to assess the position of the U.S. federal government regarding nuclear
power and spent fuel disposal.
Both issues are so politically charged that Congress, the
President, and associated federal agencies often seem reluctant to explicitly address the subject.
The site selection process for the spent fuel repository was impossibly complicated;
Congress was unable to act in an organized, efficient, and analytical way when siting the federal
repository. Although there has been a change in the attitudes of some Congressmen towards
nuclear power and its importance to national interests (most notably Senator Domenici), most of
Congress seems uncomfortable taking a stand for or against nuclear power. For this reason it is
difficult to assess the potential positions of Congress regarding DUPIC cycle implementation. A
reduction in the volume of spent fuel that will have to be dealt with in the future should appeal to
Congress and the President, as should the prospect of saving voters (taxpayers and ratepayers)
money through a reduction in spent fuel storage and disposal fees.
Implementation of the
DUPIC cycle will also alienate some number of voters; this is motivation to oppose DUPIC
implementation.
The executive branch and its agencies don't have as significant a stake in DUPIC
implementation as the legislative branch, the public, and the utilities. The executive branch's
importance to DUPIC implementation results from the decision-making power that resides in the
federal agencies such as the NRC and the Department of Energy. The NRC, in recent years, has
shown a great deal of interest in working with nuclear generators to revise licensing and other
procedures in order to make nuclear electricity generation more competitive with other forms of
power production. A principal motivation of this interest is believed to originate with the belief
that only viable generators will be able to fully fund plant decommissioning at a reactor's end-oflife. This can be seen in the NRC's decision to allow utilities to pursue higher-enrichment PWR
strategies and to streamline licensing procedures for the construction of new reactor plants. It
thus seems likely that the NRC would support DUPIC implementation in order to increase
nuclear generators' competitiveness.
The Department of Energy stands to gain significantly from implementation of the
DUPIC cycle because of the resultant decrease in future spent fuel inventories that would have to
be stored.
proliferation.
Additionally, these volumes are decreased in a way that does not encourage
The Environmental Protection Agency would similarly benefit from DUPIC
implementation, although the EPA would also be concerned about the waste generated by the
AIROX plant. The net decrease in spent fuel inventories, though, would probably earn the
EPA's approval. The NRC, DOE, and EPA, probably the three agencies with the most decisionmaking power, would thus likely support implementation of the DUPIC cycle.
Comparable agencies in Canada would also be likely to support DUPIC implementation
from the standpoint of reducing spent fuel production, especially since their repository has now
been significantly delayed. There will most likely be some regulatory agency concern about
using reactor fuel with more reactivity, but there has already been substantial testing in Canada
examining the effects of using enriched fuel and ensuring that it is safe for use in CANDUs.
Non-stakeholders will also try to influence any policies considering DUPIC cycle
implementation. Groups that could be expected to support DUPIC implementation would be
lobbying and special interest groups representing nuclear power generators and consumer
interest groups. Some environmental groups and special interest groups representing competing
power generators (natural gas, oil, and coal) would likely vocally oppose implementation of the
DUPIC cycle. It is also possible that some environmental groups would support DUPIC cycle
implementation because of the net spent fuel reduction while some consumer groups would
oppose DUPIC implementation.
9.5 Proposed Policy Implementation Method
Section 9.4 has outlined the various interests of the some of the stakeholders and
influential parties that will be affected by implementation of the DUPIC cycle. Any proposed
implementation of the DUPIC cycle would need to approach these parties and gain their support
by attempting to understand their value system and emphasize the advantages of the DUPIC
cycle that the parties value.
9.5.1
Winning the Public
In this case, the most effective method of approaching the public might be to emphasize a
combination of economic (should they be present) and environmental advantages. It is a littlecontested fact that the American public "votes their pocketbook" in Presidential elections;
economic savings through lower electricity bills might have a similarly powerful influence on
their perception of the advantages of the DUPIC cycle. This argument, of course, requires that
the DUPIC cycle save the utilities (and hence the public) money.
It should also be possible to promote the environmental advantages of the DUPIC cycle.
As discussed in Chapter 8, the DUPIC cycle has a higher natural uranium utilization and spent
fuel efficiency than the PWR-only cycle. The DUPIC cycle thus conserves a natural resource
and produces less waste. The prospect of reducing the total amount of nuclear waste (in terms of
spent fuel) and of not disposing of U.S. spent fuel in the U.S. (if disposed of in Canada) should
also appeal to the public. The DUPIC cycle could thus be promoted in terms of making the
environment safer for future generations.
It was previously mentioned that education is believed to be of primary importance in
garnering public support. Each parent, ratepayer, and environmental activist will have different
beliefs, values, and agendas.
Teaching these stakeholders the relative merits of the DUPIC
cycle, or making the information easily available, will not necessarily convince any stakeholder
to support implementation of the cycle. The information, particularly regarding transportation
and processing safety, must be readily available, however, so that any perception of "hiding the
facts" is prevented and so that concerned individuals can decide whether they will see a savings
in their electricity rates or if they need to worry about a transportation accident. It is important
that the public base its decisions on values rather than fears; fear is what has dominated public
sentiment regarding nuclear power in the past and it is this fear that must be avoided or
prevented to the greatest extent possible. Sincere attempts to address the public's concerns,
adequate information, and sufficient emphasis of the advantages of DUPIC cycle implementation
are the only ways to overcome this fear. While it is not necessary that a significant portion of the
public enthusiastically support the DUPIC cycle, it is critical to successful implementation that
some of the public support the DUPIC cycle and that the majority be at least ambivalent towards
the process. If the public is generally ambivalent (perhaps because in their minds the economic
and environmental benefits are offset by the safety questions) then it is possible that the DUPIC
cycle may be successfully implemented if other major players such as the government and
utilities support the cycle.
9.5.2
Gaining Utility Support
It is probable that power generators will be most easily swayed by economic arguments.
The most powerful argument promoting DUPIC implementation would be the prospect of
savings for both the U.S. and Canadian operators. To this end, DUPIC expenses must be divided
proportionally to the benefits. The U.S.'s gains from saving on spent fuel disposal fees will
probably be greater than Canadian fresh fuel savings. If Canada is responsible for disposing of
the spent DUPIC fuel, they should gain additional consideration when the costs and savings are
apportioned. Allocating the costs of the AIROX plant relative to the savings enjoyed by each
country and the relative costs of DUPIC implementation would be an efficient method of
ensuring that each country benefits fairly from DUPIC implementation.
Without economic arguments it may be difficult to gain the support of the utilities. It
may be possible to convince some U.S. utilities to support the DUPIC cycle as a definite means
of disposing of spent fuel in an uncertain environment but it would be difficult to gain
widespread support without economic advantages.
Depending on the size of the parties,
however, the support of one or two generators may be sufficient to begin successful
implementation.
9.5.3
Governmental Support
It may be easiest to gain the support of the U.S. and Canadian governments for DUPIC
cycle implementation.
Spent fuel disposal is a significant problem and may be a significant
expense for both countries. The DUPIC cycle's primary benefit, apart from theoretical economic
savings, is waste efficiency. A significant reduction in spent fuel volumes may allow the U.S. to
avoid construction of a second repository and may allow the Canadians to build a smaller
repository. The U.S. could realize additional savings by using the DUPIC cycle to dispose of
surplus weapons plutonium instead of building a special-purpose facility solely for this purpose.
The DUPIC cycle should not raise any proliferation concerns among either government and the
national security arguments for implementation of the DUPIC cycle are compelling, too.
The governments also have sufficient power to ensure implementation should the DUPIC
cycle prove uneconomical to the utilities. Since the spent fuel disposal and national security
arguments are strong, U.S. and Canadian governments may find it in their best interests to gain
the power generators' support by making the cycle economic. This could most effectively be
accomplished by reducing the cost of the plant, either by subsidizing construction or by
providing low-interest loans. The greatest potential barrier to implementation will be political in
both countries. Even if the government would support the cycle on technical grounds, if the
voters are opposed to cycle implementation then government approval is very unlikely.
9.6 Conclusions
The DUPIC cycle offers a proliferation-resistant solution for reducing PWR spent fuel
inventories and the amount of plutonium that must be placed in storage. This is particularly
important in light of the questionable future of the federal waste storage and repository program.
There are additional benefits in the areas of conservation and national security and
implementation of the DUPIC cycle may even help nuclear generators (and consumers) to save
money.
The DUPIC cycle must not be considered solely in terms of economics, however,
because of the number and importance of other issues. Public support, or at least widespread
ambivalence, toward the DUPIC cycle is key to successful implementation.
CHAPTER 10
Conclusions and Future Work
The successful development of an analytical model for prediction of overall DUPIC cycle
burnup facilitated analysis of the DUPIC cycle and its comparison to other cycles. While
this study fulfilled its objectives, the scope and depth of the study was necessarily limited
by time constraints. Over the course of the study, numerous additional topics were found
that merit re-examination or further investigation.
10.1 PWR Correlation Development
A series of CASMO-3 simulations permitted the development of a series of
analytical correlations that predicted PWR spent fuel isotopic composition.
These
predictions can be made based on knowledge of simple PWR fuel cycle parameters such
as reload enrichment, discharge burnup, soluble boron concentration, and cycle length.
The correlations accurately predicted PWR spent fuel burnup over a broad range of
reload enrichments and discharge burnups.
10.1.1 Future Work: Additional Confirmation of PWR Correlations
It would be valuable to confirm the accuracy of the PWR spent fuel isotopic
correlations using a different, perhaps more accurate modeling code and cross-section
library such as CASMO-4, HELIOS, or MOCUP.
10.1.2 Future Work: Additional Correlation Development
Analytical correlations to predict isotopic concentrations in spent fuel are very
useful and can save significant time and computing power. It could be very useful to
future studies if similar correlations were developed for reactors and lattices other than
the reference case such as the ABB/CE lattice used in System 80+ reactors or BWR
lattices.
10.2 AIROX Process Analysis
Analysis of the AIROX process shows that there is a non-negligible reactivity
gain due to the removal of certain fission products during the oxidation / reduction and
sintering steps. Experimental hot cell tests of the AIROX process have been completed
that examined fission product removal, yet there is still disagreement among the
published removal forecasts. The lack of consensus regarding the types and amounts of
fission product removal among other published AIROX studies is confusing and provides
strong motivation for a second round of hot cell AIROX tests that seek to specifically
quantify fission product removal. Fission products such as palladium, rhodium, silver,
and technetium that account for approximately 13, 2, 2, and 5% of all fission product
absorptions, respectively, are removed in some studies and not in others. 80% removal of
these fission products in addition to the fission products already forecast to be removed
would remove nearly 40% of all fission product absorptions during AIROX.
If further quantitative analysis of the AIROX process reveals that these elements
are not removed, it may be possible to remove palladium, rhodium, and technetium using
other dry processes. Removal of these strong poisons will further increase achievable
burnup in the CANDU portion of the DUPIC cycle.
The cooling time of DUPIC fuel before and after AIROX processing significantly
affects achievable burnup in the CANDU portion of the DUPIC cycle. PWR spent fuel
that is cooled for longer than 5 years begins to lose significant reactivity due to Pu-241
decay and Gd-155 and Am-241 buildup. Cooling times should thus be kept as short as
possible in order to extend burnup in the CANDU to the greatest extent possible.
The reactivity gain due to fission product removal during AIROX processing is
time-dependent.
The decay and buildup of fission products creates absorption cross-
sections that do not change significantly with cooling time with the notable exception of
Gd-155. It is the decay of Pu-241 and buildup of Am-241 and Gd-155 that cause most of
the reactivity decrease of spent fuel with time.
10.2.1 Future Work: Eliminating AIROX Removal Uncertainties
An effort should be made by all interested parties to determine a consistent fission
product removal forecast through additional hot-cell tests of AIROX processing on PWR
spent fuel, analysis of existing results of small-scale AIROX tests (provided they are
sufficiently detailed), or the use of computer modeling to predict fission product removal.
The existence and use of a consistent removal forecast would greatly reduce the
uncertainty associated with predicting the reactivity effects of AIROX processing,
especially when comparing the results of different studies.
For instance, using the
process described in Appendix F, AECL forecasts remove approximately 35% of all
fission product absorptions. Use of this removal fraction in the reference case integrated
model would increase achievable CANDU burnup by more than 10%, or 1.5 MWD/kg, to
15.75 MWD/kg.
100
10.2.2 Future Work: Increasing Fission Product Removal During AIROX
It may be possible to modify the AIROX process to remove additional strong
fission product poisons. Palladium, rhodium, ruthenium, and technetium form metallic
inclusions in the fuel matrix during burnup. At the end of the AIROX process, before
sintering, where the DUPIC fuel is in a fine powder form, these elements could possibly
be separated from the DUPIC fuel powder using a gas-fluidized bed or cyclone separator.
While UO2 has a density of approximately 10.4 g/cm3 , palladium, rhodium, ruthenium,
and technetium have densities of 11.97, 12.4, 12.6, and 11.5 g/cm3 , respectively. If the
size of the metallic inclusions is an appreciable fraction of the size of the UO2 particles at
this stage, the density differences of these fission products should be sufficient to permit
separation from the DUPIC fuel powder.
There may be additional dry methods to separate fission products from the
DUPIC fuel during the AIROX process such as raising temperatures during the
oxidation/reduction and sintering steps. Any additional fission product removal will
increase achievable CANDU burnup, thereby making the DUPIC cycle even more
efficient in terms of uranium utilization and spent fuel efficiency.
10.2.3 Future Work: The Effects of Cooling Time
It is also necessary to account for the effects of cooling time on DUPIC fuel
reactivity in order to generate accurate, consistent CANDU burnup predictions.
The
effect of cooling time is significant; one study estimates a decrease in k-infinite of 0.04
over a 10-year cooling period (Ref. 24). Such a decrease in k-infinite would decrease
CANDU burnup by more than 3 MWD/kg.
It may be possible to determine analytical correlations that predict both the timevariable reactivity gain from AIROX processing and the change in DUPIC fuel reactivity.
Since the reactivity change during cooling is due primarily to the concentration changes
of Pu-241, Gd-155, and Am-241, it may be possible to accurately correlate reactivity
change to cooling time considering only these three isotopes (plus one decaying generic
absorber) as per unpublished work by X.F. Zhao at MIT. If these correlations cannot be
developed it is necessary to continue using computer models to predict the time-variable
AIROX reactivity gain and fuel reactivity decrease.
10.3 CANDU Burnup Prediction
It is possible to model a circular fuel assembly using a square lattice modeling
program such as CASMO-3.
Isotopic concentrations were predicted with a maximum
error of less than 10%.
Using a data base computed using this model, it is possible to analytically predict
the discharge burnup of the CANDU portion of the DUPIC cycle. This prediction can be
made by correlating the reactivity gain from fission product removal during AIROX and
the reactivity effects of isotopic composition with a change in CANDU burnup from a
reference case. The reactivity worths of some uranium and plutonium isotopes vary nonlinearly with concentration; it is important to refine current estimates of these ratios over
a more appropriate range of concentration changes.
102
10.3.1 Future Work: CANDU Modeling in CASMO
The excess plutonium produced by this study's CANDU models (relative to
benchmarks) might be indicative that the resonance integral of U-238 is too large. It may
then be possible to improve the accuracy of the CASMO-3 models used in this study by
artificially altering the model to decrease the resonance integral of the U-238 in the fuel
pins. This could be done by slightly decreasing the fuel pellet surface area and increasing
pellet density (to conserve mass) because the resonance integral of U-238 is proportional
to the square root of pellet surface area divided by pellet mass.
The use of other
advanced computer codes that explicitly model circular lattices such as CASMO-4 (Ref.
8) or MOCUP should also increase the accuracy of the CANDU modeling.
It would also be interesting to examine the effects of using different materials in
the corner cells that were filled with air in the CANDU models. The use of these cells,
and the use of air instead of void, caused geometric anomalies in the model and the
neutronic spectrum. It may be instructive to use solid Zircaloy or void in place of these
air cells and examine the results.
10.3.2 Future Work: Relation of Reactivity Worth and Isotope Concentration
It should be possible to further increase the accuracy of the CANDU discharge
burnup predictions of this study by redetermining the values used for concentrationdependent isotopic reactivity worth, (Ak/Ax)i.
This study used a one-range-fits-all
approach that, in retrospect, was inappropriate in that the Axi values that were used were
much larger than necessary. For each isotope, varying the perturbation magnitudes over
103
ranges closer to those predicted by the PWR correlations in this study should allow the
determination of reactivity worth values more appropriate to the test cases of interest.
10.4 DUPIC Cycle Performance
Overall DUPIC cycle performance does not improve significantly once PWR
reload enrichments increase past 5.5 or 6 w/o U-235, corresponding to burnups around 60
to 70 MWD/kg. As PWRs operate at higher burnups, PWR spent fuel production is
decreased and more units are required to fuel a single CANDU. While DUPIC spent fuel
production is reduced as PWR enrichment increases, the overall reduction is much
smaller than the initial savings in spent fuel production from switching to a DUPIC cycle.
The incremental gains in spent fuel savings are thus small and might be easily
outweighed by increasing natural uranium consumption and higher PWR fueling costs for
the multiple PWRs. By itself, then, the DUPIC concept thus does not provide significant
motivation for the development and use of ultra-high PWR burnup capabilities.
In terms of resource utilization, there does not appear to be an incentive to change
PWR fuel management practices to enhance CANDU performance in the DUPIC cycle.
The requirement of multiple PWRs to fuel one CANDU reactor causes the PWR
characteristics to dominate the DUPIC cycle in terms of uranium utilization and spent
fuel efficiency. In general, the DUPIC cycle simply increases PWR spent fuel efficiency
and PWR uranium utilization by a fixed value over the range of reload enrichments.
Large reductions in spent fuel production volumes are possible for both parties
should a CANDU-only country use and dispose of AIROX-processed fuel from a PWRonly country. A win-win situation may develop since the PWR-only country would not
104
have to dispose of spent PWR fuel used in the DUPIC cycle and since the CANDU-only
country would see a factor of two reduction in spent fuel generation.
10.4.1 Future Work: DUPIC Economic Performance
Analysis of DUPIC performance with respect to economics instead of resource
utilization and waste savings might reveal additional optimal performance points.
It
would also be useful to agree upon a common set of cost parameters for uranium
procurement, fuel fabrication, and spent fuel storage and disposal.
This would allow
more accurate and more consistent comparison of economic benefits of DUPIC
implementation. The most recent comprehensive IAEA study on the economics of the
nuclear fuel cycle (Ref 29), would appear to be a good starting point for this analysis.
The wide range of projected spent fuel storage and disposal costs, in particular,
makes economic analysis and comparison of the DUPIC cycle difficult. It is important to
consider that in a Canadian/U.S. DUPIC cycle spent DUPIC fuel disposal costs might
vary with the country of disposal. If the U.S. were to treat DUPIC fuel the same as spent
PWR fuel, disposal costs would be much higher in the U.S. than in Canada, thereby
making an even stronger case for Canadian disposal of DUPIC spent fuel.
10.4.2 Future Work: Alternative Cycle Comparisons
It may be instructive to compare DUPIC performance to another parallel cycle
that uses slightly enriched uranium (SEU) in a CANDU as an alternative to NU fuel.
SEU fuel (with about 1.1-1.3 w/o enrichment) would confer many of the benefits of
DUPIC fuel (extended burnup, power uprating, decrease in waste generation) without as
105
significant an increase in fuel fabrication costs.
SEU fuel cycles have already been
examined in some detail by AECL but have not been formally compared to DUPIC cycle
performance.
10.5 Policy Analysis
There are numerous potential stakeholders when considering implementation of
the DUPIC cycle in North America.
As shown in Chapter 9, the public, nuclear
generators, and the government might all benefit substantially from the DUPIC cycle.
These gains would be most quantifiable in economic terms should the DUPIC cycle save
money, but it is difficult to estimate the true economic costs and savings of any
implementation of the DUPIC cycle, particularly at such an early stage of analysis. There
are additional environmental and national security gains from DUPIC implementation but
the economic motivator would be the most powerful factor in winning approval of the
DUPIC cycle.
The public is the most influential group regarding DUPIC implementation: its
political power can sway government policy and the prospect of extensive litigation
would likely deter nuclear generators from DUPIC cycle implementation should the plan
be opposed by a significant part of the populace. If the DUPIC cycle does not incur
significant public
opposition, it is likely the government
would approve
of
implementation and could then encourage nuclear generators to follow suit should the
DUPIC cycle prove uneconomical without government subsidies.
106
10.5.1 Future Work: Additional Analysis of Public Values and Cycle Economics
As public opinion and lack of significant opposition will be extremely important
to successful implementation of the DUPIC cycle in North America, additional policy
analysis focusing on public opinion and methods of gaining public support will be
extremely critical in creating a plan for implementation.
It is critical to identify the
values of the public and to determine how these values can be served through
implementation of the DUPIC cycle. As saving money is a common value and almost
every individual will be affected as an electricity ratepayer, additional economic analysis
to quantify the total savings (or expense) of the DUPIC cycle is crucial to making an
effective economic argument supporting DUPIC implementation.
10.5.2 Future Work: Use of DUPIC Cycle to Burn Weapons Plutonium
Studies by the U.S. and Canada (Ref 3) have shown that it is technically feasible
to burn weapons-grade plutonium in CANDU reactors by mixing the plutonium with
depleted uranium. Additional policy analysis should examine the possibility of using the
DUPIC cycle and spent PWR fuel to bum weapons plutonium as well.
It should be
possible to enrich and burn in CANDUs spent PWR fuel that has cooled for a long time
and would otherwise be uneconomical for use in the DUPIC cycle. This would make the
plutonium extremely proliferation-resistant and avoid generation of significant amounts
of CANDU spent fuel.
In addition, one could consider building an AIROX facility, similar to that used in
the DUPIC cycle, that would blend weapons plutonium into spent PWR fuel and fabricate
107
PWR fuel assemblies instead of CANFLEX fuel bundles. Recycling spent PWR fuel in
this manner would confer the benefits of fuel recycling that were discussed in Chapter 9
and increase national security by making the weapons plutonium proliferation-resistant.
If all surplus weapons plutonium can be disposed of using the DUPIC cycle or PWR
recycle instead of a specialized disposal facility, the U.S. taxpayers could directly save
several billion dollars in avoided plant construction costs and additional monies in
avoided spent fuel disposal costs.
10.5.3 Future Work: Comparison of Proliferation Resistance of DUPIC Cycle
While preliminary evaluation of the decrease in spent fuel volumes has been
made, other measures of proliferation resistance should be better evaluated.
This
includes the self-protection characteristics of the spent fuel and the plutonium isotope
mix in the spent fuel. The DUPIC cycle appears to be advantageous in both of these
areas: extending the spent PWR fuel burnup in the CANDU significantly increases
protective fission product inventories while burning significant amounts of fissile
plutonium.
108
References
[1]
AMERICAN NUCLEAR SOCIETY, "World List of Nuclear Power Plants,"
Nuclear News, Vol. 41-3, March 1998.
[2]
FEINROTH, HERBERT, "Research, Development, and Demonstration Plan for
AIROX Dry Recycle of Nuclear Reactor Spent Fuel," Gamma Engineering
Corporation, GAMMA 8540-2, August 1993.
[3]
AECL TECHNOLOGIES, INC., "Plutonium Consumption Program - CANDU
Reactor Project," Final Report, July 1994.
[4]
MACLEAN, H.J., McMAHON, M.V., AND DRISCOLL, M.J., "Uranium and
Separative Work Utilization in Light Water Reactors," Massachusetts Institute of
Technology Department of Nuclear Engineering, MIT-NFC-TR-009, January
1998.
[5]
ELLIS, R.J., "DUPIC Fuel Cycle Physics Calculations and Benchmark Studies,"
AECL Report in Preparation, 1998.
[6]
FUJII, H. and MORISHIMA, A., eds. Directory of Nuclear Power Plants in the
World. Tokyo: Japan Nuclear Energy Information Center Co., Ltd., 1994.
[7]
Personal communication with Kookjong Lee, Engineer at KNFC, July 1997.
[8]
KNOTT, D., EDENIUS, M., PELTONEN, J., and ANTTILA, M., "Results of
Modeling Hexagonal and Circular Cluster Fuel Assembly Designs using
CASMO-4," Proceedingsof the American nuclear SocietyTopicalMeeting Advances in NuclearFuel ManagementII, Myrtle Beach, March 23-26, 1997,
EPRI TR-107728-V1.
[9]
TORGERSON, D.F., BOCZAR, P.G., and DASTUR, A.R., "CANDU Fuel Cycle
Flexibility," AECL Chalk River, AECL-11129, June 1994.
[10]
TANG, J.R., DRISCOLL, M.J., AND TODREAS, N.E., "Physics Considerations
in a Passive Light Water Pressure Tube Reactor (PLPTR)," MIT-ANP-TR-009,
November 1991.
[11]
SULLIVAN, J.D. AND COX, D.S., "AECL's Progress in Developing the DUPIC
Fuel Fabrication Process," 4 International Conference on CANDU Fuel,
Pembroke, Ontario, October 3-5, 1995.
[12]
JAHSHAN, S. N. and McGEEHAN, T.J., "An Evaluation of the Deployment of
AIROX-Recycled Fuel in Pressurized Water Reactors," Nuclear Technology, Vol.
106, June 1994.
109
[13]
FRANKEL, A.J. and SHAPIRO, N.L., "Appraisal of PWR-HWR Tandem Fuel
Cycles," Combustion Engineering, Inc., NPSD-45, February 1977.
[14]
CHRISTENSEN, D.E. AND PREZBINDOWSKI, D.L., "Isotopic Correlation
Safeguards Techniques: Ratios and Their Properties," Transactionsof the
American Nuclear Society, Vol. 18, June 1974.
[15]
"Conceptual Design and Cost Evaluation of the DUPIC Fuel Fabrication Facility
Final Report," SCIENTECH, Inc. and Gamma Engineering Corporation, SCIECOM-219-96, May 1996.
[16]
Personal communication with Herbert Feinroth of Gamma Engineering Corp.,
March 1998.
[17]
MAJUMDAR, D., et al, "Recycling of Nuclear Spent Fuel with AIROX
Processing," U.S. Department of Energy Idaho Field Office, DOE/ID-10423,
December 1992.
[18]
LIDE, D.R., ed., "CRC Handbook of Chemistry and Physics," Ann Arbor: CRC
Press, 1994.
[19]
DEAN, J.A., ed., "Lange's Handbook of Chemistry, Thirteenth Edition," New
York: McGraw-Hill Book Company, 1985.
[20]
REID, R.C., PRAUSNITZ, J.M., SHERWOOD, T.K., "The Properties of Gases
and Liquids,Third Edition," New York: McGraw-Hill Book Company, 1977.
[21]
KUBASCHEWSKI, 0. and ALCOCK, C.B., "Metallurgical Thermochemistry,
Fifth Edition" New York: Pergamon Press, 1979.
[22]
REID, R.C., PRAUSNITZ, J.M., SHERWOOD, T.K., "The Properties of Gases
and Liquids, Fourth Edition," New York: McGraw-Hill Book Company, 1985.
[23]
McMAHON, M.V., DRISCOLL, M.J., and PILAT, E.E., "Natural Uranium and
Separative Work Utilization in LWRs," Transactionsof the American Nuclear
Society, Vol. 76, June 1997.
[24]
SANDERS, T.L. and WESTFALL, R.M. "Feasibility and Incentives for Burnup
Credit in Spent-Fuel Transport Casks," Nuclear Science and Engineering,Vol.
104, January 1990.
[25]
BOCZAR, P.G. and DASTUR, A.R., "CANDU / PWR Synergism," IAEA
Technical Committee Meeting on "Advances in Heavy Water Reactors," Toronto,
Canada, June 1993.
110
[26]
RIM, CHANG S., "Fueling CANDUs with PWR Spent Fuel," Lecture at MIT,
April 1995.
[27]
FEINROTH, HERBERT, Personal communication with Dr. Neil Todreas,
February 1996.
[28]
Korean Atomic Energy Research Institute, DUPIC cyle Background and
Information Pages, http://websys.kaeri.re.kr/nuclear/nu-f3-7.html.
[29]
NUCLEAR ENERGY AGENCY, "The Economics of the Nuclear Fuel Cycle,"
Organisation for Economic Co-Operation and Development, 1994.
[30]
COHN, S.M., "Too Cheap to Meter: An Economic and Philosophical Analysis of
the Nuclear Dream," State University of New York Press, 1997.
[31]
BAUMGARTNER, P., ATES, Y., ELLIS, R.J., TAYLOR, P. and BOCZAR,
P.G., "Disposal Costs for Selected Advanced CANDU Fuel Cycles," 6th
International Conference on Nuclear Engineering, May 10-15,1998.
[32]
SILVER, RAY, "Lack of Public Support Stymies Canadian Waste Disposal
Effort," Nucleonics Week, March 19, 1998.
[33]
"Uranium 1995: Resources, Production, and Demand," Paris: OECD, 1996.
[34]
CHOI, H., RHEE, B.W., and PARK, H., "Physics Study on Direct Use of Spent
Pressurized Water Reactor Fuel in CANDU (DUPIC)," Nuclear Science and
Engineering,Vol. 126, May 1997.
APPENDIX A
Burnup Correlations for Fixed Cycle Length or Batch Number
Maclean et al. in (Ref. 4) show that, for a generic LWR, correlations can be used to
predict discharge burnup (BdL) when knowing reload enrichment (Xp) and either constant cycle
burnup (BcL) or batch number (n) such that 1/nh' of the reactor core is replaced each refueling.
If reload enrichment and batch number are known and fixed then the averaged end of
cycle burnup of all fuel in the core, Bi (equivalent to batch reload burnup), can be determined
using Eq. (A-1).
B, = 28 X, -0.88 -19.2,MWD/kgU
(A -1)
The discharge burnup can then be directly determined using Eq. (A-2).
B,=
2n B, MWD/kgU
(A-2)
If X, and BcL are known instead, as in the case when a fixed cycle length is more
important than a fixed batch number, discharge burnup can still be calculated. B 1 should first be
determined, again using Eq. (A-1). Discharge burnup can then be determined from Eq. (A-3).
Bd = 2B
-BcL, MWD/kgU
(A- 3)
Once BdL is determined the integrated model can be used to predict Bdc based on Xp, BdL,
and Bor.
BcL can be determined from Eq. (A-4) where p is the rated core specific power [kW/kgU],
L is the cycle-average capacity factor from startup to startup, and Tc is the cycle length from
startup to startup [calendar days].
BC =
MWD/kgU
PL( 1000,
112
(A-4)
APPENDIX B
Sample CASMO Input and Output for Reference PWR
This appendix contains a CASMO input card and a summary output table. Figure B-1
shows a sample CASMO input card for the reference PWR lattice using fresh fuel enriched to 3
w/o U-235 and burned to 35 GWD/MT. The results of this modeling are shown in Table B-1 as
a summary of spent fuel isotopic concentration and K-infinite values.
Table B-1: Summary of Reference PWR Fuel Isotopics and K-Infinite as a Funtion of Burnup
Brmup
(IVIW /kg)
0
0.5
1
5
10
15
20
25
30
35
WT(%)
92234
0.024
0.024
0.024
0.022
0.021
0.019
0.018
0.016
0.015
0.014
K-inf
92235
3
2.941
2.883
2.46
2.007
1.621
1.291
1.012
0.779
0.587
92236
0
0.011
0.021
0.096
0.174
0.238
0.291
0.332
0.364
0.387
92238
96.976
96.946
96.916
96.67
96.349
96.01
95.652
95.275
94.878
94.46
94238
0.000
0.000
0.000
0.000
0.001
0.002
0.004
0.006
0.010
0.014
113
94239
0.000
0.019
0.044
0.200
0.321
0.391
0.430
0.450
0.457
0.457
94240
0.000
0.000
0.001
0.020
0.058
0.099
0.139
0.176
0.207
0.234
94241
0.000
0.000
0.000
0.004
0.020
0.044
0.068
0.090
0.109
0.123
94242
0.000
0.000
0.000
0.000
0.002
0.006
0.014
0.026
0.040
0.057
1.2900
1.2762
1.2696
1.2213
1.1611
1.1073
1.0578
1.0114
0.9677
0.9270
Figure B-1: Sample CASMO Input Card for Reference PWR
*
*
*
CASMO-3 Input
17x17 Westinghouse PWR Assembly
WABA IN
*
3.0 U-235 Enrichment
*
*
10 Sep 97
Chad Bollmann
*
Massachusetts Institute of Technology
*
Changes
= 0
TIT,TFU=1000.0,TMO=580.0,BOR=0.0,VOI .0 *No Boron
*0.711 enrich, no burnable
SIM,'STANDARD',3.0,0,0,0,0.0
absorber
*ZR4
MI1,6.550,.7000E-05/302=100.0
*SS347
MI2,7.900,.1800E-04/347=100.0
*INC718
MI3,8.200,.1800E-04/718=100.0
*INC750
MI4,8.200,.1800E-04/750=100.0
4 8 0 0 0= 5 0 0 0
00
*AGINCD
.
=15.00,
MI5,10.16,.2250E-04/47000=80.00,490
0
37
*BORSHIM
.87, 1 3 0 0 0 = 3 .4 4
MI6,2.260,.3250E-05/8000=54.81,14000=
5010=.7100,5011=3.170
*IFBA-1.5X
MI7,0.23595/5010=100
33
*Pt DET
,1 3 0 0 0 = 1 3 .85
MI8,2.2543/718=73.82,8000=12.
7 64
7
MI9,4.3,,/13000=34.324,8000=30.530,5000=2 .506,6000= . 0 *WABA
MATERIAL
*ZR4
BOX,6.550,.7000E-05/302=100.0
*FUEL
PIN,1,.3922,.4001,.4572
*GUIDET
PIN,2,0.5817,.6020/"COO","BOX"
*INSTR THIM
.6147/"AIR","MI3","MI8"
.5690,
.5042,
.4166,
.2553,
.1727,
PIN,3,
"MI3", "COO", "BOX"/2,4,6/1, "DET"
PIN,4, .1727,.2553, .4166,.5042,.5690, .6147/"AIR","MI3","MI8" *INSTR THIM
"MI3", "COO", "BOX"/2,4,6/1
*RCCA
PIN,5,.4331,.4369,.4839,.5690,.6147/"MI5","AIR","MI2"
"COO", "BOX"/1,3,5/1, "AIC"
*RCCA
PIN,6,.4331,.4369,.4839,.5690,.6147/"MI5","AIR","MI2"
"COO", "BOX"/1, 3,5/1, "BTH"
PIN,7,0.2858,0.3391,0.3531,0.4039,0.4178,0.4839,0.5613,0.6020/"COO"
*WABA
"MI1", "AIR", "MI9" , "AIR", "MI1", "COO", "MI1"/1, 6, 8/1, "WAB"
9 347
8
7
=15. 4 1
SPA,13.99,.1800E-04,,8.154/ 18= 4.5 ,
*Pressure
PRE,158
*Power Dens Korean
PDE,41.8
*No Box,21.5 Pitch
PWR,17,1.260,21.50,,,,,8
35
30
25
DEP,0.0 0.5 1 5 10 15 20
*Need Desnsity
FUE,1,10.33/3.0
7300 6
Neglect Term
= .0
*FUE,2,10.1315/3.100,
*20 BA
LPI
2
1 1
1
1 1
1 2
2 1
1 1 1
1 1
1 1 1 2
1 1
1 2 1 1 1
2 1
1 1 1 1 1 1 1
1
1 1 1 1 1 1 11
1
LST,STA
114
APPENDIX C
Sample CANDU CASMO Models
This appendix contains sample CASMO input cards for both the 37 pin and 43 pin
CANFLEX bundle models. Figure C-1 shows a sample CASMO input card for the 37-pin
CANDU bundle model. This model was developed before the CANFLEX model in order to
verify that the approximation techniques employed in this study would be effective.
It was
thought it would be easier to create an accurate model with the 37-pin bundle since this model
uses natural uranium fuel (instead of more complicated mixed-oxide fuel) and the pin layout
approximates a circular, symmetric bundle.
*
*
*
*
*
Figure C-1: Sample CASMO Input Card for 37 Pin CANDU Model
CAS 3
37 ELEMENT CANDU FILE
MODIFICATION TO ZIRC DENSITY, XE, AND NOT ALLOWING THERMAL EXPANSION
CHAD BOLLMANN
MIT 31 JULY 97
TIT,,TFU=900,TMO=345.66,,IDE='CAN1'
SIM, 'CANTEST',0.711,0.0,0,0,0
FUE,1,10.36, ,,/0.711
CAN, 6.440, 0.7E-5,,
BOX,6.479,0,560.66/29063=0.72 302=99.28
*MOD DENS ADJUSTED
MOD,1.111,,,/1102=20.097 8000=79.894 1001=0.009
COO,0.81212,,560.66/1102=19.973 8000=79.949 1001=0.078
* COO DENS
ADJUSTED
PIN,1,0.605,0.616, 0.654
PIN,2,0.8214,0.8215/'AIR', 'CAN'//-1
PRE, 100.0
BWR,7,1.6435,11.5045,0.428,7.7895,7.7895, 4
LPI
1111
1111
1112
1 1 2 2/'F'
PDE, 25.4
XEN, 0
THE, 0
0.0
DEP,0.0,0.2,1.0,2.0,3.0,4.0,5.0,6.0,7.0,8.0,9.0,
LST, STA
END
115
Figure C-2 shows sample CASMO input for the reference case CANFLEX model. The
model layout as shown in Figure C-2 is the same as was used in the benchmark CANFLEX
model cases, but for the reference case model the fuel composition has been changed to reflect 4
w/o, 45 MWD/kg spent PWR fuel that has been cooled for 10 years and refabricated using the
AIROX process as described in Chapter 6.
116
Figure C-2: Sample CASMO Input for Reference Case CANFLEX Model
*
*
*
*
*
*
*
CAS 3
CANFLEX CANDU FILE
BASELINE CALCS FOR DK/DK DETERMINATION
USES YongWang-1 4 w/o 45 GWD
CHAD BOLLMANN
MIT 16 MAR 98
EQUILIBRIUM XENON AT STARTUP
= '
TIT,,TFU=900,TMO=345.66,,IDE CAN1'
SIM,'CANTEST',0.711,0.0,0,0,0
FUE,1,10.358,,,/0.77765 401=1.6430E+00 402=3.4850E-01 36083=4.5782E-05
45103=4.3563E-02 45105=9.5056E-05 47109=6.7293E-03 53135=5.5337E-07
54131=0.0000E+00 54135=0.0000E+00 55133=1.2877E-03 55134=1.5688E-04
55135=3.9663E-04 60143=9.2812E-02 60145=7.6875E-02 61147=1.9117E-02
61148=1.0989E-04 61149=1.4423E-04 61248=1.7557E-04 62147=8.2062E-03
62149=2.5629E-04 62150=3.5542E-02 62151=1.4281E-03 62152=1.2350E-02
63153=1.3967E-02 63154=3.8900E-03 63155=2.2487E-03 64155=1.3975E-05
92234=1.6818E-02 92236=5.2104E-01 92238=9.2959E+01 92239=5.8949E-05
93237=6.4275E-02 93239=8.5409E-03 94238=2.5359E-02 94239=5.2170E-01
94240=1.9326E-01 94241=1.7424E-01 94242=7.8614E-02 95241=5.2004E-03
95242=7.8164E-05 95243=1.7714E-02 96242=2.5498E-03 96244=5.6797E-03
FUE,2,10.358,,,/0.77765 401=1.6430E+00 402=3.4850E-01 36083=4.5782E-05
45103=4.3563E-02 45105=9.5056E-05 47109=6.7293E-03 53135=5.5337E-07
54131=0.0000E+00 54135=0.0000E+00 55133=1.2877E-03 55134=1.5688E-04
55135=3.9663E-04 60143=9.2812E-02 60145=7.6875E-02 61147=1.9117E-02
61148=1.0989E-04 61149=1.4423E-04 61248=1.7557E-04 62147=8.2062E-03
62149=2.5629E-04 62150=3.5542E-02 62151=1.4281E-03 62152=1.2350E-02
63153=1.3967E-02 63154=3.8900E-03 63155=2.2487E-03 64155=1.3975E-05
92234=1.6818E-02 92236=5.2104E-01 92238=9.2959E+01 92239=5.8949E-05
93237=6.4275E-02 93239=8.5409E-03 94238=2.5359E-02 94239=5.2170E-01
94240=1.9326E-01 94241=1.7424E-01 94242=7.8614E-02 95241=5.2004E-03
95242=7.8164E-05 95243=1.7714E-02 96242=2.5498E-03 96244=5.6797E-03
FUE,3,10.358,,,/0.77765 401=1.6430E+00 402=3.4850E-01 36083=4.5782E-05
45103=4.3563E-02 45105=9.5056E-05 47109=6.7293E-03 53135=5.5337E-07
54131=0.0000E+00 54135=0.0000E+00 55133=1.2877E-03 55134=1.5688E-04
55135=3.9663E-04 60143=9.2812E-02 60145=7.6875E-02 61147=1.9117E-02
61148=1.0989E-04 61149=1.4423E-04 61248=1.7557E-04 62147=8.2062E-03
62149=2.5629E-04 62150=3.5542E-02 62151=1.4281E-03 62152=1.2350E-02
63153=1.3967E-02 63154=3.8900E-03 63155=2.2487E-03 64155=1.3975E-05
92234=1.6818E-02 92236=5.2104E-01 92238=9.2959E+01 92239=5.8949E-05
93237=6.4275E-02 93239=8.5409E-03 94238=2.5359E-02 94239=5.2170E-01
94240=1.9326E-01 94241=1.7424E-01 94242=7.8614E-02 95241=5.2004E-03
95242=7.8164E-05 95243=1.7714E-02 96242=2.5498E-03 96244=5.6797E-03
CAN,6.440,0.7E-5,,
7
BOX,6.479,0,560.66/29063=0. 2 302=99.28
97
*MOD DENS ADJUSTED
8000=79.894 1001=0.009
MOD,1.1590,,,/1102=20.0
* COO DENS
COO,0.4696,,560.66/1102=19.973 8000=79.949 1001=0.078
ADJUSTED
PIN,4,0.5370,0.542,0.575/'3','AIR','CAN'
PIN,3,0.5370,0.542,0.575/'2','AIR','CAN'
PIN,1,0.8214,0.8215/'AIR','CAN'//-1
PIN,2,0.6345,0.6395,0.675/'1','AIR','CAN'
PRE,100.0
7 53 4
98 7 5 3
, . 14, . 1 ,,1
BWR,7,1.6059,11.2412,0.4
117
*AIR GAP PINS*
APPENDIX D
Alternate AIROX Fission Product Removal Forecasts
The fission product removal forecasts for this study differed from those of the three other
sources included in this appendix. Table D-1 compares the forecasts of this study and three
additional studies.
Table D-l: Comparison of Fission Product Percent Removal Forecasts During AIROX
INEL (Ref. 17)
Scientech (Ref. 15)
This Study
(1)
(2)
(3)
Ag
Cd
Cs
I
In
Ir
Kr
Mo
100
100
80
0
99
99
0
0
99
80
100
100
0
75
90
100
75
0
100
0
100
100
0
75
100
100
0
75
100
0
100
100
0
80
99
99
75
0
99
80
Pd
80
0
0
0
Rh
Ru
80
80
0
90
0
100
0
80
Se
Tc
Te
80
80
99
0
Discrepancy
75
0
0
75
99
0
99
Xe
100
100
100
100
Nuclide
14C
3H
AECL (Ref. 5)
Notes:
(1) INEL, in Ref 17, states that hot cell tests by Atomics International in the 1960s
show that "small" amounts of technetium are removed during sintering. However,
in later quantitative discussions no forecast is made regarding expected technetium
removal.
(2) The Scientech forecasts were developed to predict released fission product waste
streams and thus may be biased upwards in some cases to produce a conservative
(in terms of waste-handling) estimate.
118
This study's removal predictions, as outlined in Table D-1, were developed from an
analysis of the chemical properties of important fission products. This analysis was performed
during the course of this study by Michael Reynard, a graduate student at MIT. This chemical
analysis focused primarily on the possible oxides and other species formed during the AIROX
process and the melting and boiling points of these compounds. If elements were found to form
oxides or hydrides with boiling points below that of the operating temperature of the oxidation /
reduction stage (1200 C) then a certain portion of the element was assumed to be removed.
Removal estimates from other sources were then compared and this study's removal forecast was
established based on the removal forecasts for similar elements. Then the boiling points of
various oxides in Table A-2 were compared to the sintering temperature (1600 C) to determine
which, if any, fission product oxides might be removed during sintering. This study's final
forecasts for removal during sintering were also based on forecasts for similar elements from the
other studies.
Table D-2: Analysis of Fission Product Chemical Properties
Element
Known
Oxides
Ag
AgO
Ag2O
Known
Hydrides
Source
1
2
1
2
1
Boiling Pt
(del C)
2212
2164
Melting Pt
(deE C)
962
960.15
dec> 100
dec 100
dec 230
dec 200
Boil During
Redux / Oxid
no
no
no
no
no
no
Boil During
Sintering
no
no
no
no
no
no
2607
2607
994
994
no
no
no
no
no
no
no
no
2
1
2
Am
AmO2
Am203
1
1
119
Element
Known
Known
Oxides
Hydrides
Ba
BaO
BaO2
BaO2.8H20
BaH2
Cd
CdO
Cm
Cs
CsH
Cs20
Source
1
2
1
2
4
1
2
4
1
1
2
1
2
4
2
1
2
1
2
1
2
1
Eu203
2
1
2
1
2
4
1
1
2
1
1
2
5
1
1
2
2
4
1
2
1
4
1
2
2
Gd203
4
1
2
1
Cs202
Cs203
Cs7O
D2
HD
D20
HDO
Dy
Dy203
Er
Er203
Eu
Gd
Boiling Pt
Melting Pt
Boil During
Boil During
(del C)
1640
1849
(deg C)
725
725
3088
2013
1920
800, -02
2750
450
Sintering
no
no
no
no
no
no
no
no
no
yes
no
yes
yes
subl 1497
1345 +-40
1350
28.4
28.8
dec
dec
Redux /Oxid
no
no
no
no
no
no
no
no
no
no
no
yes
yes
no
no
no
no
yes
yes
no
no
490 (in N2)
no
no
490 (in N2)
400
400
400
400
502
3
-254
-253
-257
4
3.82
no
no
no
no
no
no
no
yes
yes
yes
no
yes
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
yes
yes
yes
no
yes
yes
no
no
no
no
no
no
no
no
no
yes
yes
no
no
no
no
no
no
1400
767
770
1497
669
678.5
dec 400
650, -02
650, -02
-250
-248
-251
101.43
101.46
2567
2600
2868
2900
1527
1440
3273
3000
dec > 1000
321
321
1412
1500
2340
2380
1529
1497
infusable
2400
822
826
2050
623
1313
1306
2330 +-20
2340
2
120
yes
yes
no
no
yes
yes
no
no
Element
K Know
Oxides
Known
Hydrides
H2
H20
H202
Ho
Ho203
12
102
I205
1409
HI
Kr
La
La203
Mo
Mo203
MoO2
MoO3
Pd
PdO2.xH20
PdO
PdO.xH20
PdO.xH20
Pd2H
Pm
Pu epsilon
Pu
PuO
PuO2
Pu203
Rh
RhO2
RhO2.2H20
Rh203
Rh203.5H20
Ru
RuO2
RuO4
Source
1
1
1
2
1
2
1
1
2
1
1
2
1
2
1
2
3
1
2
1
2
2
1
2
1
1
1
1
2
1
1
2
1
2
1
1
1
2
2
2
2
1
2
1
1
1
2
1
1
2
1
1
2
Boiling Pt
(deg C)
-252.8
100
150.2
151.2
2700
2600
Melting Pt
(deg C)
-259.34
0
-0.41
-0.4
1474
1461
184
184
130
113.5
113.6
-35.4
-35.35
-35.5
-153
-153
3464
3470
4200
4639
4646
1155
2970
2940
3000
3230
dec 2800
3727 +-100
3727
3900
4119
dec 108
40
dec 300-350
dec 275
dec. 75
dec. 75
-51
-51
-51
-157
-157
918
920
2320
2623
2610
dec - 1100
801
1554
1550
dec, -H20, -O
870
dec 870
dec
dec
dec
1042
640
1900
2390 (in He)
2085 (in He)
1966
1966
dec
dec 1100-1150
dec 1100
dec
2310
2427
dec
25.5
25
Boil During
Redux / Oxid
yes
yes
yes
yes
no
no
no
yes
yes
yes
no
no
no
no
yes
yes
yes
yes
yes
no
no
no
no
no
no
no
yes
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
yes
Boil During
Sintering
yes
yes
yes
yes
no
no
no
yes
yes
yes
no
no
no
no
yes
yes
yes
yes
yes
no
no
no
no
no
no
no
yes
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
yes
Element
Known
Oxides
Known
Hydrides
Sb
Sb205
Sb204
Sb203
SbH3
Se
Se
Se
SeO2
Se3
Sm
Sn gray
Sn white
Sn brittle
Sn
SnO
SnO.xH20
SnO2
Sn2
SnO2.xH20 (alpha)
SnO2.xH20 (beta)
SnH4
Sr
SrO
SrO2
SrH2
Tb
Tc
Tc2
Tc207
Te
Tet2
Tee3
Te205
Xe
Source
Boiling Pt
(deg C)
Melting Pt
(deg C)
Boil During
Redux / Oxid
Boil During
Sintering
1
2
2
2
2
1750
1635
630.5
630.5
dec 380
dec 930
no
no
no
no
yes
no
no
no
no
yes
1
-17.1
-88
no
no
2
2
1
1
2
1
2
1
2
-18.4
684
685
685
685
subl 315
subl
dec 180
-91.5
217
170-180
60-80
221
subl316*,340-350
340*
118
118
yes
yes
yes
yes
yes
no
no
no
no
yes
yes
yes
yes
yes
yes
no
no
no
656
1
1790
1072
no
no
2
1803
1072
no
no
1
1
1
2
1
2270
2260
2260
2623
232
stable 13-161
stable >161
stable 13-161
no
no
no
no
no
no
no
no
no
no
subl 1800-1900*
subl 1900*
1630
1630
-52
-52
1382
1381
dec - 150
-150
777
769
2532
2665
215 dec
215 dec
1050
d>1000
1359
1356
2157
2250+-50
subl 1000
119.5
449.51
450
no
no
no
no
no
yes
yes
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
no
yes
yes
yes
yes
no
no
no
no
no
no
no
no
no
no
no
yes
yes
yes
yes
ves
yes
yes
1
1
2
1
1
1
2
1
2
1
2
1
2
1
2
1
2
1
2
2
2
1
2
3221
2800
4265
4567
310.6
988
1009
1
1245
733
no
yes
2
1
2
2
1
2
subl1790?
732.6
430
dec 400
dec 450
-112
-112
no
no
no
no
yes
yes
no
no
no
no
yes
yes
-107
-108
It is unknown why both CRC and Lange make this same apparent contradiction regarding SnO2 and SeO2.
122
Key to Sources:
1 = CRC (Ref. 18)
2 = Lange's Handbook of Chemistry (Ref. 19)
3 = "Properties of Gases and Liquids," Reid (Ref. 20)
4 = "Metallurgical Thermochemistry," Kubaschewski (Ref. 21)
5 = "The Properties of Gases and Liquids" (1977), by Reid (Ref. 22)
123
Key To Abbreviations:
"subl" = sublime
"dec" = decompose
"+-" = +/- denoting temperature variability
APPENDIX E
Influence of Cooling Time on Spent PWR Fuel
The cooling time of spent PWR fuel has a significant effect on the reactivity of spent
PWR fuel. Two effects, actinide decay and the resultant fission product buildup, are the primary
cause of a decrease in k-infinite of approximately 0.04 over a ten-year cooling time, as shown in
Figures E-1 and E-2. This decrease represents slightly less than half of the excess reactivity in a
batch of spent PWR fuel modeled by Sanders and Westfall. (Ref. 24)
Figure E-1 shows k-infinite, the infinite multiplication factor for PWR spent fuel having
cooling time (t) when loaded into an appropriate CANDU lattice model. Figure E-2 shows the
effects of fission product buildup and AIROX removal on k-infinite.
Figure E-1: Change in K-Infinite Over Time Due to Actinide Decay and Fission Product Buildup
2
3
1
k-infinite(t)
4
5
O+E
0
5
10
Cooling Time, t [Years]
124
15
20
Fig. E-1 Notes:
(1) Gain due to Xe-135 decay.
(2) Loss due to Pu-241 decay.
(3) Loss due to Am-241 buildup.
(4) Loss due to Gd-155 buildup.
(5) Loss or gain due to all other fission products and minor actinides.
Figure E-2: Change in K-Infinite Over Time Due to Fission Product Buildup and AIROX Removal
5
10
15
Cooling Time, t [Years]
Fig. E-2 Notes:
(6) Gain due to removal of all fission products.
(7) Gain due to removal of fission products during AIROX processing.
Approximately 20% of this reactivity decrease is due to the beta decay of Pu-241 and
subsequent formation of Am-241, a fairly strong poison. (Shown by Note 1 and Note 2 in Fig. 71). The effects of this decay are even more pronounced than other decays because, as seen in
125
Figure 7-2, Pu-241 has the greatest concentration-based reactivity effect of the examined
isotopes.
Another significant cause of the decrease in spent fuel reactivity is the buildup of Gd- 155
from beta decay of Eu-155 (Note 3 of Figure 7-1). This effect is particularly insidious because
Eu-155 has a short half-life (approximately 4.7 years) and Gd-155 has an extremely large
neutron-capture cross section (approximately 61000 barns). Gd-155 alone increases the total
fission product absorptions by nearly 10% over 10 years. (Ref 24)
The remaining decrease in reactivity is due to buildup of actinide absorbers (Note 4 in
Figure 7-1).
Because of the significant decrease in reactivity as cooling time increases, it is extremely
advantageous to use "fresh" spent PWR fuel in the DUPIC process. Spent fuel that has cooled
less than 5 years loses relatively little reactivity and is very suitable for use in the DUPIC cycle.
Spent fuel that has cooled for 20 or more years (of which a significant inventory exists in the
U.S.) is much less attractive from a DUPIC standpoint. Additional delays in the implementation
of the DUPIC cycle in the U.S. will only serve to further decrease the attractiveness of a majority
of the spent fuel stocks.
This study used computer modeling (ORIGEN) to cool spent PWR fuel for varying
periods of time. ORIGEN accounts for fission product decays and yields to estimate the fraction
of absorptions due to an individual isotope out of the total absorptions due to all fission products.
The fission product absorption fractions after 10 years of cooling were used to estimate the total
fission product absorptions removed during the AIROX process as further discussed in Appendix
F. An inconsistency arises because in the present work, Ak6 was determined at time t = 0 but
126
the fraction of fission product absorptions removed, given by Ak7 /Ak6 , was determined at time
t = 10 years and applied as discussed in Chapter 7.
Additional work by Zhao at MIT on this topic is progressing. He has been able to
develop a simple analytic correction for the Ak (t) components discussed above. This
correlations allows the user to estimate k -infinite (t) after a certain period of time and obviates
the need for new ORIGEN runs to estimate this value for any time frame of interest. These
results will be documented in an internal MIT report.
127
APPENDIX F
Estimation of Fission Product Absorption Fraction Removal
During AIROX
Table F-1 shows the individual isotopic absorptions as a percent of all fission production
absorptions at three different times: immediately after removal from the PWR, 1 year after
removal, and 10 years after removal. Note that the columns labeled "% Absorptions [out of 1]"
give the fractional absorption of each fission product at the corresponding cooling times. These
results were calculated using the ORIGEN computer modeling program. The total percentage of
fission product absorptions that is removed during AIROX can be calculated by applying the
results outlined in Table 6-1 to Table F-1 as follows:
(1)
Choose a cooling time.
(2)
For each nuclide that is removed during AIROX, multiply that nuclide's fractional
removal during AIROX by the nuclide's individual absorption fraction to obtain the
"absorptions removed."
(3)
Sum all entries in the "absorptions removed" column to obtain the fraction of total fission
product absorptions that are removed during AIROX, Rx.
(4)
Return to the integrated model with the calculated Rx and apply that removal fraction as
described in Chapter 7. Note that the correlation for removal of all fission products (as
given by Eq. 7-3) must be modified to correspond to fuel having the cooling time
specified in Step (1) above.
It is important to make calculations that are consistent in terms of cooling time. It is
important to be consistent in time because there will be a significant change in fission product
absorption fractions as certain nuclides decay and other nuclides are formed.
Gd-155, for
instance, has a very large neutron absorption cross-section and has a fairly high yield;
128
concentrations and the resultant absorptions of this strong poison increase dramatically over the
course of 10 years.
In this study the fraction of all fission product absorptions removed in the AIROX
process for the reference CANDU, 20.4%, was calculated for spent PWR fuel that was cooled for
10 years (Note 7 in Fig. 7-2).
The reactivity worth of all fission products in spent PWR fuel,
Akfp, was determined from uncooled spent PWR fuel (Note 6 in Fig. 7-2). This inconsistency
can be expected to cause errors in this study's estimation of achievable CANDU discharge
burnup.
The total fission product reactivity worth at 10 years should be greater than for
uncooled fuel due to isotopic concentration changes such as the buildup of Gd-155. Further
discussion of the spent fuel reactivity changes with cooling can be found in Appendix E. Thus,
this study's correlation for Akfp will underrepresent the actual worth of all fission products.
Work by X. Zhao has developed a correlation to predict the changing reactivity worth of all
fission products with cooling time for PWRs.
Because of the cooling time inconsistency, this study will underestimate fission product
removal and the results of this study should be expected to consistently underrepresent
achievable CANDU and DUPIC burnup. The error induced by this inconsistency should be
relatively small, however: discharge burnup should be consistently underestimated by no more
than 10%, or 1.5 MWD/kg of CANDU burnup and hence less than 3% of DUPIC burnup. This
will introduce a systematic bias so that all relative trends will be the same and all relative
comparisons should be valid.
129
Table F-1: ORIGEN Isotopic Fractional Absorption Estimation for 4.5 w/o, 50 MWD/kg PWR Spent Fuel
Cooled
Nuclide
GD155
RH103
ND143
SM149
CS133
XE131
SM152
SM151
EU153
ND145
SM150
SM147
AG109
EU154
KR 83
EU155
CS135
PM147
CD113
CS134
GD158
DY164
PM148M
PM148
1135
PM149
RH105
XE135
TC 99
MO 95
RU101
PR141
PD105
MO 98
PD108
LA139
ZR 93
GD157
1129
EU151
MO 97
PD107
0 Years
1 Year
10 Years
% Absorptions [out of 1]
9.77E-04 2.49E-02 1.32E-01
9.70E-02 1.18E-01 1.12E-01
9.14E-02 1.03E-01 9.80E-02
4.87E-02 9.31E-02 8.84E-02
6.55E-02 7.29E-02 6.92E-02
5.95E-02 6.66E-02 6.33E-02
4.03E-02 4.45E-02 4.23E-02
3.64E-02 4.06E-02 3.59E-02
3.19E-02 3.54E-02 3.37E-02
2.93E-02 3.23E-02 3.07E-02
2.16E-02 2.38E-02 2.26E-02
6.87E-03 1.06E-02 1.87E-02
1.78E-02 1.97E-02 1.87E-02
2.35E-02 2.39E-02 1.10E-02
6.53E-03 7.21E-03 6.84E-03
2.34E-02 2.25E-02 6.06E-03
4.69E-03 5.18E-03 4.92E-03
4.30E-02 3.85E-02 3.39E-03
2.42E-03 2.77E-03 2.64E-03
1.39E-02 1.09E-02 5.03E-04
1.61 E-04 1.78E-04 1.69E-04
5.90E-05 6.51E-05 6.19E-05
1.20E-02 2.89E-05 2.99E-29
7.48E-03 9.48E-08 9.80E-32
3.79E-09 0.00E+00 0.00E+00
6.60E-04 0.00E+00 0.00E+00
8.53E-03 0.00E+00 0.00E+00
1.22E-01 0.00E+00 0.00E+00
5.01E-02 5.55E-02 5.27E-02
2.12E-02 2.60E-02 2.47E-02
1.60E-02 1.76E-02 1.67E-02
1.00E-02 1.15E-02 1.09E-02
9.20E-03 1.02E-02 9.66E-03
7.70E-03 8.50E-03 8.07E-03
6.71 E-03 7.41 E-03 7.04E-03
6.33E-03 6.98E-03 6.63E-03
5.77E-03 6.37E-03 6.05E-03
4.01 E-03 5.40E-03 5.12E-03
4.68E-03 5.21E-03 4.94E-03
3.38E-05 5.14E-04 4.41 E-03
4.06E-03 4.49E-03 4.26E-03
3.84E-03 4.24E-03 4.02E-03
% Removal
Our Forecast
0.99
1
0.99
0.99
0.8
0.99
0.99
1
0.8
0.8
0.8
0.99
0.8
130
Removed
Absorptions
0.00E+00
0.00E+00
0.00E+00
0.00E+00
6.49E-02
5.95E-02
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
6.46E-03
0.00E+00
4.64E-03
0.00E+00
1.94E-03
1.37E-02
0.00E+00
0.00E+00
0.00E+00
0.00E+00
3.75E-09
0.00E+00
0.00E+00
1.22E-01
0.00E+00
1.69E-02
1.28E-02
0.00E+00
0.00E+00
6.16E-03
0.00E+00
0.00E+00
0.00E+00
0.00E+00
4.64E-03
0.00E+00
3.25E-03
0.00E+00
Remaining
Absorptions
9.77E-04
9.70E-02
9. 14E-02
4.87E-02
6.55E-04
0.00E+00
4.03E-02
3.64E-02
3.19E-02
2.93E-02
2.16E-02
6.87E-03
1.78E-02
2.35E-02
6.53E-05
2.34E-02
4.69E-05
4.30E-02
4.84E-04
1.39E-04
1.61 E-04
5.90E-05
1.20E-02
7.48E-03
3.79E-11
6.60E-04
8.53E-03
0.00E+00
5.01E-02
4.23E-03
3.19E-03
1.00E-02
9.20E-03
1.54E-03
6.71 E-03
6.33E-03
5.77E-03
4.01 E-03
4.68E-05
3.38E-05
8.13E-04
3.84E-03
Cooled
Nuclide
ND144
IN115
GD154
ND148
1127
RU102
GD156
PDI04
ZR 91
SM148
ZR 96
BA134
MO100
RU104
ND146
BA137
RU100
PD106
XE132
CD110
CE142
CD111
BR 81
ZR 92
ND150
Y 89
CE140
XE134
ND142
SR 90
MO 96
RB 85
SM154
SB121
BA138
XE136
TB159
KR 84
SB123
RB 87
CS137
DY161
KR 82
ZR 94
DY162
CD112
PD110
CD114
TE125
DY160
TE130
0 Years
1 Year
10 Years
% Absorptions [out of 1]
2.61 E-03 3.30E-03 3.41 E-03
1.57E-03 1.82E-03 1.73E-03
2.07E-04 4.40E-04 1.65E-03
1.54E-03 1.70E-03 1.61 E-03
1.37E-03 1.55E-03 1.48E-03
1.41 E-03 1.55E-03 1.47E-03
1.33E-03 1.55E-03 1.47E-03
1.33E-03 1.47E-03 1.39E-03
1.24E-03 1.44E-03 1.37E-03
1.09E-03 1.22E-03 1.16E-03
1.08E-03 1.20E-03 1.14E-03
3.08E-04 5.73E-04 1.07E-03
1.00E-03 1.10E-03 1.05E-03
9.54E-04 1.05E-03 9.99E-04
8.38E-04 9.25E-04 8.78E-04
1.40E-04 2.36E-04 8.53E-04
7.52E-04 8.30E-04 7.89E-04
4.07E-04 5.93E-04 7.00E-04
6.45E-04 7.13E-04 6.77E-04
5.95E-04 6.66E-04 6.37E-04
5.83E-04 6.43E-04 6.11 E-04
5.68E-04 6.36E-04 6.04E-04
5.74E-04 6.33E-04 6.01 E-04
5.62E-04 6.21E-04 5.89E-04
5.39E-04 5.95E-04 5.65E-04
4.61E-04 5.30E-04 5.03E-04
3.81 E-04 4.27E-04 4.05E-04
3.52E-04 3.88E-04 3.69E-04
3.02E-04 3.34E-04 3.17E-04
3.58E-04 3.86E-04 2.96E-04
2.73E-04 3.01E-04 2.86E-04
2.34E-04 2.62E-04 2.72E-04
2.53E-04 2.80E-04 2.65E-04
2.38E-04 2.63E-04 2.50E-04
2.26E-04 2.50E-04 2.37E-04
2.07E-04 2.28E-04 2.16E-04
1.82E-04 2.02E-04 1.92E-04
1.79E-04 1.98E-04 1.88E-04
1.69E-04 1.93E-04 1.84E-04
1.60E-04 1.76E-04 1.68E-04
1.57E-04 1.70E-04 1.31E-04
1.20E-04 1.36E-04 1.30E-04
1.18E-04 1.31E-04 1.25E-04
1.12E-04 1.24E-04 1.17E-04
1.12E-04 1.23E-04 1.17E-04
9.15E-05 1.01 E-04 9.60E-05
9.09E-05 1.00E-04 9.53E-05
8.96E-05 9.89E-05 9.39E-05
2.68E-05 4.47E-05 8.82E-05
5.13E-05 7.84E-05 7.51 E-05
7.10E-05 7.83E-05 7.44E-05
% Removal
Our Forecast
0.75
0.99
0.8
0.8
0.8
0.8
1
0.8
0.8
1
0.8
1
0.99
0.99
0.99
0.8
0.8
0.99
0.99
Removed
Absorptions
0.00E+00
1.18E-03
0.00E+00
0.00E+00
1.36E-03
1.12E-03
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
8.00E-04
7.63E-04
0.00E+00
0.00E+00
6.02E-04
0.00E+00
6.45E-04
4.76E-04
0.00E+00
4.54E-04
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
3.52E-04
0.00E+00
0.00E+00
2.18E-04
0.00E+00
0.00E+00
0.00E+00
0.00E+00
2.07E-04
0.00E+00
1.77E-04
0.00E+00
0.00E+00
1.56E-04
0.00E+00
1.17E-04
0.00E+00
0.00E+00
7.32E-05
0.00E+00
7.17E-05
2.65E-05
0.00E+00
7.02E-05
Remaining
Absorptions
2.61 E-03
3.93E-04
2.07E-04
1.54E-03
1.37E-05
2.81 E-04
1.33E-03
1.33E-03
1.24E-03
1.09E-03
1.08E-03
3.08E-04
2.00E-04
1.91 E-04
8.38E-04
1.40E-04
1.50E-04
4.07E-04
0.00E+00
1.19E-04
5.83E-04
1.14E-04
5.74E-04
5.62E-04
5.39E-04
4.61E-04
3.81 E-04
0.00E+00
3.02E-04
3.58E-04
5.45E-05
2.34E-04
2.53E-04
2.38E-04
2.26E-04
0.00E+00
1.82E-04
1.79E-06
1.69E-04
1.60E-04
1.57E-06
1.20E-04
1.18E-06
1.12E-04
1.12E-04
1.83E-05
9.09E-05
1.79E-05
2.68E-07
5.13E-05
7.10E-07
1 Year
10 Years
Cooled
0 Years
Nuclide
TE128
DY163
XE130
SE 77
SN117
ZR 90
SN116
SN124
KR 85
SE 79
SN118
KR 86
H0165
XE128
EU152
BA136
SN119
TE122
SE 80
BA135
TE123
IN113
SN115
SE 78
SN126
CD116
SB125
AS 75
SR 88
SN120
SE 82
SR 86
TE124
GD152
SN122
TE126
GD160
ER167
XE129
GE 73
LI 6
ER166
RU 99
SE 76
LA138
Y 90
GE 76
SR 87
RU106
GE 74
% Removal
Our Forecast
% Absorptions [out of 1]
0.99
7.09E-05 7.82E-05 7.43E-05
6.42E-05 7.09E-05 6.73E-05
1
5.67E-05 6.26E-05 5.95E-05
0.8
3.87E-05 4.28E-05 4.06E-05
3.43E-05 3.78E-05 3.59E-05
5.56E-06 8.93E-06 2.97E-05
2.60E-05 2.87E-05 2.72E-05
2.56E-05 2.83E-05 2.69E-05
0.99
3.52E-05 3.65E-05 1.94E-05
0.8
1.78E-05 1.97E-05 1.87E-05
1.78E-05 1.96E-05 1.87E-05
0.99
1.64E-05 1.81E-05 1.71E-05
1.59E-05 1.76E-05 1.67E-05
1
1.54E-05 1.70E-05 1.62E-05
2.45E-05 2.57E-05 1.54E-05
1.31E-05 1.49E-05 1.41 E-05
1.31E-05 1.45E-05 1.38E-05
0.99
1.30E-05 1.45E-05 1.38E-05
0.8
1.19E-05 1.31E-05 1.24E-05
1.10E-05 1.22E-05 1.16E-05
0.99
8.69E-06 1.18E-05 1.15E-05
0.75
1.08E-06 2.15E-06 8.52E-06
7.85E-06
7.46E-06 8.27E-06
0.8
6.68E-06 7.37E-06 7.00E-06
6.11E-06 6.74E-06 6.40E-06
0.8
4.96E-06 5.48E-06 5.20E-06
5.18E-05 4.48E-05 4.48E-06
3.92E-06 4.33E-06 4.11E-06
3.54E-06 3.90E-06 3.70E-06
2.69E-06 2.97E-06 2.82E-06
0.8
2.65E-06 2.93E-06 2.78E-06
2.50E-06 2.89E-06 2.74E-06
0.99
2.10E-06 2.73E-06 2.59E-06
2.29E-06 2.56E-06 2.59E-06
2.06E-06 2.28E-06 2.16E-06
0.99
2.03E-06 2.27E-06 2.16E-06
1.98E-06 2.18E-06 2.07E-06
1.60E-06 1.76E-06 1.68E-06
1
1.35E-06 1.49E-06 1.42E-06
1.29E-06 1.42E-06 1.35E-06
1.22E-06 1.35E-06 1.28E-06
1.03E-06 1.15E-06 1.09E-06
0.8
1.37E-07 2.34E-07 9.27E-07
0.8
6.65E-07
6.33E-07 7.01E-07
6.06E-07 6.69E-07 6.35E-07
6.47E-07 6.59E-07 5.05E-07
4.30E-07 4.75E-07 4.51 E-07
1.65E-07 1.83E-07 1.73E-07
0.8
8.26E-05 4.58E-05 8.93E-08
6.05E-08 6.68E-08 6.34E-08
132
Removed
Absorptions
7.02E-05
0.00E+00
5.67E-05
3.09E-05
0.00E+00
0.00E+00
0.00E+00
0.00E+00
3.49E-05
1.43E-05
0.00E+00
1.62E-05
0.00E+00
1.54E-05
0.00E+00
0.00E+00
0.00E+00
1.29E-05
9.49E-06
0.00E+00
8.60E-06
8.13E-07
0.00E+00
5.34E-06
0.00E+00
3.97E-06
0.00E+00
0.00E+00
0.00E+00
0.00E+00
2.12E-06
0.00E+00
2.08E-06
0.00E+00
0.00E+00
2.00E-06
0.00E+00
0.00E+00
1.35E-06
0.00E+00
0.00E+00
0.00E+00
1.10E-07
5.06E-07
0.00E+00
0.00E+00
0.00E+00
0.00E+00
6.61 E-05
0.00E+00
Remaining
Absorptions
7.09E-07
6.42E-05
0.00E+00
7.74E-06
3.43E-05
5.56E-06
2.60E-05
2.56E-05
3.52E-07
3.56E-06
1.78E-05
1.64E-07
1.59E-05
0.00E+00
2.45E-05
1.31E-05
1.31E-05
1.30E-07
2.37E-06
1.10E-05
8.69E-08
2.71 E-07
7.46E-06
1.34E-06
6.11 E-06
9.92E-07
5.18E-05
3.92E-06
3.54E-06
2.69E-06
5.31E-07
2.50E-06
2.10E-08
2.29E-06
2.06E-06
2.03E-08
1.98E-06
1.60E-06
0.00E+00
1.29E-06
1.22E-06
1.03E-06
2.74E-08
1.27E-07
6.06E-07
6.47E-07
4.30E-07
1.65E-07
1.65E-05
6.05E-08
Cooled
0 Years
1 Year
10 Years
% Removal
Nuclide
BR 79
CE144
ER168
GE 72
NB 94
AG107
TM169
KR 80
NB 93
AG110M
ZN 68
YB170
CD108
SN114
ZN 70
BE 9
YB171
GA 69
GA 71
SB126
LI 7
H3
CD109
BE 10
TM171
YB172
ER170
ZN 66
ZN 67
TE127M
SN123
C 14
AG110
TM 170
TB160
NB 95
ZR 95
Y 91
SB124
SR 89
CD115M
RU103
CE141
TE129M
AG111
BA139
Our Forecast
% Absorptions [out of 1]
6.13E-09 1.04E-08 4.05E-08
2.04E-04 9.22E-05 2.89E-08
2.61 E-08 2.88E-08 2.74E-08
2.27E-08 2.51E-08 2.39E-08
2.21 E-08 2.43E-08 2.31E-08
1.22E-09 2.40E-09 1.13E-08
9.60E-09 1.12E-08 1.07E-08
6.76E-09 7.46E-09 7.08E-09
0.99
3.09E-1 0 5.69E-1 0 4.70E-09
8.14E-05 3.26E-05 3.40E-09
2.00E-09 2.20E-09 2.09E-09
3.47E-1 0 6.48E-1 0 6.56E-1 0
6.04E-1 0 6.67E-1 0 6.33E-10
0.8
4.07E-1 0 4.82E-1 0 4.58E-1 0
3.49E-1 0 3.86E-10 3.66E-1 0
1.15E-10 1.27E-10 1.21E-10
3.32E-11 5.23E-11 8.25E-11
1.68E-11 1.86E-11 1.77E-11
1.56E-11 1.72E-11 1.64E-11
1.19E-07 1.43E-11 1.35E-11
8.98E-12 9.91E-12 9.41E-12
1
6.68E-1 2 6.97E-12 3.99E-12
0.8
1.87E-10 1.19E-10 8.35E-1 3
7.96E-1 3 8.78E-1 3 8.34E-1 3
1.25E-11 9.61E-12 3.54E-1 3
1.75E-13 1.94E-13 1.85E-13
6.20E-14 1.14E-13 1.15E-13
3.33E-14 3.68E-14 3.49E-14
1.65E-14 1.83E-14 1.73E-14
1.26E-05 1.41E-06 1.12E-15
0.99
2.05E-07 3.18E-08 6.59E-16
1
1.15E-16 1.27E-16 1.20E-16
1.68E-09 2.47E-1 3 2.57E-17
6.40E-10 9.87E-1 1 1.89E-18
2.77E-05 9.23E-07 1.81 E-20
1.79E-04 8.08E-06 2.68E-21
8.75E-05 1.85E-06 5.99E-22
3.70E-05 5.43E-07 6.29E-24
6.29E-07 1.04E-08 3.58E-25
7.69E-06 5.64E-08 1.36E-27
0.8
2.61E-06 9.84E-09 6.03E-31
0.8
6.36E-04 1.12E-06 6.81 E-32
5.45E-04 2.51E-07 8.71 E-38
0.99
1.77E-06 1.05E-09 3.51 E-39
8.04E-06 1.55E-20 0.00E+00
1.87E-07 0.00E+00 0.00E+00
Removed
Absorptions
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
6.69E-09
0.00E+00
0.00E+00
0.00E+00
0.00E+00
4.83E-10
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
6.68E-12
1.49E-10
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
1.25E-05
0.00E+00
1.15E-16
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
2.08E-06
5.09E-04
0.00E+00
1.75E-06
0.00E+00
0.00E+00
Remaining
Absorptions
6.13E-09
2.04E-04
2.61 E-08
2.27E-08
2.21E-08
1.22E-09
9.60E-09
6.76E-11
3.09E-10
8.14E-05
2.00E-09
3.47E-10
1.21E-10
4.07E-10
3.49E-10
1.15E-10
3.32E-11
1.68E-11
1.56E-11
1.19E-07
8.98E-12
0.00E+00
3.73E-11
7.96E-13
1.25E-11
1.75E-13
6.20E-14
3.33E-14
1.65E-14
1.26E-07
2.05E-07
0.00E+00
1.68E-09
6.40E-10
2.77E-05
1.79E-04
8.75E-05
3.70E-05
6.29E-07
7.69E-06
5.21 E-07
1.27E-04
5.45E-04
1.77E-08
8.04E-06
1.87E-07
Cooled
Nuclide
BA140
CD118
CE143
CS134M
CS136
CS141
CU 66
DY165
ER171
EU156
EU157
GD161
GE 75
1130
1131
IN117
IN117M
IN119
IN119M
IN120
IN120M
KR 87
LA140
MO 99
ND147
NI 66
PM151
PR142
PR143
RB 86
RB 88
RH104
RH104M
RU105
SM153
SN125
TE132
TE134
XE133
0 Years
1 Year
10 Years
% Removal
Our Forecast
% Absorptions [out of 1]
4.16E-05 1.16E-13 0.00E+00
0.8
1.83E-10 0.00E+00 0.00E+00
1.27E-05 0.00E+00 0.00E+00
7.55E-09 0.00E+00 0.00E+00
0.99
0.99
7.17E-06 3.21E-14 0.00E+00
1.65E-12 0.00E+00 0.00E+00
0.99
1.08E-17 0.00E+00 0.00E+00
2.43E-07 0.00E+00 0.00E+00
2.31E- 17 0.00E+00 0.00E+00
1.27E-03 8.08E-11 0.00E+00
3.81 E-06 0.OOE+00 0.00E+00
3.42E-08 0.00E+00 0.00E+00
4.41 E-1 1 0.00E+00 0.00E+00
7.58E-07 0.00E+00 0.00E+00
0.99
9.58E-06 2.31E-19 0.00E+00
0.99
0.75
9.76E-10 0.00E+00 0.00E+00
0.75
3.27E-09 0.00E+00 0.00E+00
2.62E-12 0.00E+00 0.00E+00
0.75
4.92E-11 0.00E+00 0.00E+00
0.75
0.75
2.15E-14 0.00E+00 0.00E+00
0.75
1.49E-15 0.00E+00 0.00E+00
4.17E-06 0.00E+00 0.00E+00
0.99
2.43E-05 7.41E-14 0.00E+00
0.8
1.92E-05 2.80E-45 0.00E+00
5.17E-04 6.54E-14 0.00E+00
0.00E+00 0.00E+00 0.00E+00
9.08E-05 0.00E+00 0.00E+00
2.20E-06 0.00E+00 0.00E+00
8.35E-04 8.05E-12 0.00E+00
2.49E-07 3.51E-1 3 0.00E+00
2.32E-09 0.00E+00 0.00E+00
8.07E-09 0.00E+00 0.00E+00
6.45E-08 0.00E+00 0.00E+00
2.43E-07 0.00E+00 0.00E+00
0.8
4.08E-04 0.00E+00 0.00E+00
2.83E-07 1.23E-18 0.00E+00
7.98E-09 1.58E-42 0.00E+00
0.99
0.99
1.42E-09 0.00E+00 0.00E+00
1
8.97E-04 1.33E-24 0.00E+00
134
Removed
Absorptions
0.00E+00
1.47E-10
0.00E+00
7.47E-09
7.10E-06
1.64E-12
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
7.50E-07
9.48E-06
7.32E-1 0
2.45E-09
1.97E-12
3.69E-11
1.61E-14
1.12E-15
4.13E-06
0.00E+00
1.53E-05
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
0.00E+00
1.94E-07
0.00E+00
0.00E+00
7.90E-09
1.40E-09
8.97E-04
Remaining
Absorptions
4.16E-05
3.67E-11
1.27E-05
7.55E-11
7.17E-08
1.65E-14
1.08E-17
2.43E-07
2.31E-17
1.27E-03
3.81E-06
3.42E-08
4.41 E-11
7.58E-09
9.58E-08
2.44E-10
8.17E-10
6.55E-13
1.23E-11
5.38E-15
3.74E-16
4.17E-08
2.43E-05
3.83E-06
5.17E-04
0.00E+00
9.08E-05
2.20E-06
8.35E-04
2.49E-07
2.32E-09
8.07E-09
6.45E-08
4.86E-08
4.08E-04
2.83E-07
7.98E-11
1.42E-11
0.00E+00
APPENDIX G
Sample Determination of (Ak/AX)i for U-235
Table G-1 shows a sample spreadsheet for determining the concentration-dependent
reactivity worth, (Ak/AX)i, for U-235. Note that all 'D's in Table G-1 represent the symbol 'A'
as used throughout the rest of this study.
Table G-1: Sample Determination of U-235 (Ak/AX)i
Burnup
[MWD/kg]
0
1
2
3
4
5
6
7
8
9
10
11
12
13
14
Reference Case
K-inf
92235
[w/o]
1.2128
0.778
1.1654
0.735
1.1439
0.694
1.1217
0.652
1.0996
0.611
1.0776
0.571
1.0559
0.532
1.0346
0.494
1.0138
0.457
0.9935
0.421
0.9739
0.386
0.9549
0.353
0.9367
0.322
0.9193
0.292
0.9028
0.264
92235
[w/o]
1.778
1.711
1.646
1.581
1.516
1.452
1.389
1.327
1.265
1.204
1.145
1.086
1.029
0.973
0.918
Test Case 1
Dk
K-inf
Simp Dk
92235
Test Case 2
Dk
K-inf
Simp Dk
[w/o]
1.3268
1.2770
1.2631
1.2487
1.2340
1.2192
1.2042
1.1892
1.1740
1.1588
1.1434
1.1280
1.1125
1.0970
1.0814
0.1140
0.1116
0.1192
0.1269
0.1344
0.1416
0.1483
0.1545
0.1602
0.1653
0.1696
0.1731
0.1758
0.1777
0.1787
0.1140
0.4462
0.2385
0.5077
0.2688
0.5662
0.2966
0.6182
0.3204
0.6610
0.3391
0.6925
0.3517
0.7108
0.1787
2.278
2.204
2.131
2.059
1.987
1.916
1.846
1.776
1.707
1.638
1.571
1.504
1.439
1.374
1.311
1.3663
1.3154
1.3035
1.2912
1.2787
1.2660
1.2532
1.2403
1.2272
1.2140
1.2007
1.1873
1.1737
1.1600
1.1462
0.1536
0.1500
0.1597
0.1695
0.1791
0.1884
0.1973
0.2056
0.2134
0.2205
0.2269
0.2324
0.2370
0.2407
0.2435
0.1536
0.5998
0.3193
0.6778
0.3582
0.7536
0.3946
0.8226
0.4268
0.8820
0.4537
0.9295
0.4740
0.9629
0.2435
Summary of Results
Case 1 Case 2
Dx
1
1.5
Average Dk
Simpson's Dk
DkDx
0.1501
0.1502
0.1502
0.2012
0.2012
0.1342
For both cases of each isotope, the concentration-dependent isotopic reactivity worth,
given as "Dk/Dx" in Table G-1, was determined at each burnup step as follows:
(1)
Calculate the isotopic change in K-Infinite, Dk, by subtracting Test Case K-Infinite from
Reference Case K-Infinite. (The reference DUPIC CANDU calculation is discussed in
Chapter 7.)
(2)
Integrate Dk using Simpson's Rule.
(3)
Calculate Dx by subtracting Test Case U-235 from Reference Case U-235.
135
(4)
Obtain Dk/Dx by dividing integrated Dk by Dx.
The values from Steps (2) through (4) for each case are summarized in the "Summary of
Results" subtable in Table G-1. The average value of Dk, "Average Dk", is included to allow
comparison of the results from two methods of determining Dk as well as to serve as a check of
the Simpson's Rule integration.
Only the Dk/Dx values from Case 1 are used in the integrated model. The Case 2 Dk/Dx
value for U-235 differs significantly from the Case 1 value, as do the Case 2 values for most of
the other isotopes, as shown in Table G-2.
Table G-2: Comparison of Case 1 and Case 2 Isotopic Concentration-Dependent Reactivity Worths
Isotope
DkIDx
Case 1
Case 2
U-234
U-235
U-236
U-238
Pu-238
Pu-239
Pu-240
Pu-241
-0.0314
0.1502
-0.0109
-0.0008
-0.0762
0.1939
-0.0440
0.2755
-0.0304
0.1342
-0.0107
-0.0008
-0.0707
0.1780
Invalid (1)
0.2573
Pu-242
Notes:
-0.0312
-0.0271
(1) An invalid data set was obtained
for this case due to CASMO input
error.
The large differences in Dk/Dx between Case 1 and Case 2 show that there is a non-linear
relationship between change in isotopic concentration and the resultant change in K-Infinite.
This, in turn, suggests that the appropriateness of a set of Dk/Dx values for use in calculations
depends on the range of isotopic perturbations that are being examined. For this study, where
isotopic concentration variability from test case conditions varied with the isotope being
considered (maximum Dx of-4 w/o for U-238 and minimum Dx of 0.009 for U-234), the use of
one blanket set of perturbations did not provide the maximum accuracy. It is suggested that
136
future work in this area should include recalculation of the isotopic concentration-dependent
reactivity worths using base-case perturbations that are more appropriate to the actual isotopic
concentration variability.
137
APPENDIX H
Benefits from Use of Slightly Enriched Uranium in CANDUs
Instead of using spent PWR or MOX fuel in CANDUs, equivalent gains in burnup can be
achieved using slightly enriched uranium (SEU) without the significant increases in fuel
handling concerns that occur with use of MOX or DUPIC fuel. A SEU CANDU can realize the
burnups given in Table H-i estimated using Eq. (H-1). The results of the correlation given in
Eq. (H-I) are compared to "Reference Bd" values, given in Table H-I, obtained from AECL.
(Ref. 31)
Bd
[MWD/kg]= 50 X -0.11-31.25
(H -1)
Table H-1: Approximate Discharge Burnup of SEU-Fueled CANDU
Xp
Correlation Bd
Reference Bd
[w/o U-235]
0.711
0.9
1.0
[MWD/kg]
7.51
13.19
15.9
[MWD/kg]
8
14
NA
1.2
1.5
20.95
27.70
21
28
By following the same procedure as applied in Ref (4) and in Chapter 8, maximum
uranium utilization occurs at approximately Xp = 1.3 w/o U-235, at which point Bd is
approximately tripled compared to a NU CANDU (Ref 4). Spent fuel arisings are thus reduced
by a factor of three. The uranium utilization of a Parallel cycle using a SEU-fueled CANDU can
be given by Eq. (H-2) where the subscript 'SEU' denotes the variable value corresponding to a
SEU-fueled CANDU (Ref. 4).
138
rB,
jmsEv
+ rlSEUBdSEU
mpM
U,[MWDe/kgUnat] =
(H- 2)
It would be interesting to evaluate the DUPIC cycle relative to a Parallel cycle using an
SEU-fueled CANDU in addition to a Parallel cycle using an NU CANDU. The advantage of the
DUPIC cycle in terms of uranium utilization and spent fuel efficiency would be greatly
decreased relative to an SEU Parallel cycle. It is also likely that a CANDU operator would first
operate on an SEU cycle prior to burning MOX fuel in order to gain experience operating with
more enriched, higher-reactivity fuels.
139