THEMATIC NETWORK FOR A PHEBUS FPT-1
INTERNATIONAL STANDARD PROBLEM
(THENPHEBISP)
CO-ORDINATOR
B. CLEMENT
IRSN/DPAM, Cadarache
127, Rue des Martyrs ( B. P. 85 )
F - 38054 Grenoble Cedex 9
FRANCE
Tel.: + 33 7688 3244
Fax: + 33 7688 5177
LIST OF PARTNERS
1) IRSN, Cadarache (Fr)
2) JRC, Petten (Nl)
3) AEA Technology (UK)
4) AEKI, Budapest (Hu)
5) NRI, Rez (Cz)
6) ENEA, Bologna (It)
7) UPI, Uv. Pisa (It)
8) UPM, Uv. Madrid (Sp)
9) FZK, Karlsruhe (Ge)
10) GRS, Cologne (Ge)
11) PSI, Villigen (CH)
12) SCK, Mol (Be)
13) EDF, Clamart (Fr)
14) JSI, Ljubljana (Sl)
15) ENPROCO (Bu)
CONTRACT N°:
EU contribution
Starting Date:
Duration:
FIKS-CT-2001-20151
Euro 240899
December 2001
24 months
CONTENTS
LIST OF ABBREVIATIONS AND SYMBOLS
EXECUTIVE SUMMARY
A. OBJECTIVES AND SCOPE
B. WORK PROGRAMME
B.1
Preparation
B.2
Intermediate Comparison Workshop
B.3
Comparison and Assessment
C. WORK PERFORMED AND RESULTS
C.1
Specification Report
C.2
Intermediate Comparison Workshop
C.3
Comparison and Assessment
C.3.1
Representation of the facility
C.3.2
Analysis of the results
C.3.3
Assessment of Codes and Models
C.3.4
Computing Assessment
C.3.5
Integral Aspects
C.3.6
Implications for Plant Studies
CONCLUSION
REFERENCES
FIGURES
LIST OF ABBREVIATIONS AND SYMBOLS
CSNI
Committee for the Safety of Nuclear Insatllations
FP
Fission Products
FPT-i
Phebus FP test n°I
IRSN
Institut de radioprotection et de Sûreté Nucléaire
ISP
International Standard Problem
OECD
Organisation for Economic Development and Cooperation
PBF-SFD
Power Burst facility-Severe Fuel Damage
RCS
Reactor Coolant System
EXECUTIVE SUMMARY
The THENPHEBISP 2-year thematic network started in December 2001, and
was concerned with OECD/CSNI International Standard Problem 46, itself based on
the Phebus FPT1 core degradation/source term experiment. The aim was to assess the
capability of computer codes to model in an integrated way the physical processes
taking place during a severe accident in a pressurised water reactor, from the initial
stages of core degradation, the fission product transport through the primary circuit
and the behaviour of the released fission products in the containment. ISP-46,
coordinated by IRSN Cadarache, attracted 33 participating organisations, from 23
countries and international bodies, who submitted 47 base-case calculations and 21
best-estimate calculations, using 15 different codes.
The thermal behaviour of the fuel bundle and the hydrogen production were
generally well captured, and good agreement for the core final state could be obtained
with a suitable choice of bulk fuel relocation temperature, however this is unlikely to
be representative of all plant studies so sensitivity calculations are needed with the
modelling in its current state. Total volatile fission product release was simulated, but
its kinetics, and the overall modelling of semi-volatile, low-volatile and structural
material release (Ag/In/Cd, Sn) needs improvement. Overall retention in the circuit is
well predicted, but calculations underestimate deposits in the upper plenum and
overestimate those in the steam generator, also the volatility of some elements could
be better predicted. Containment thermal hydraulics and depletion rate of aerosols are
well calculated, but with difficulties related to partition amongst the deposition
mechanisms. Calculation of iodine chemistry in the containment turned out to be more
difficult. Its quality strongly depends of the calculation of release and transport in the
integral codes. The major difficulties are related to the existence of gaseous iodine in
the primary circuit and to the prediction of the amount of organic iodine in the gas
phase.
Beyond the assessment of codes and models, as usually done in International
Standard Problems, conclusions were made with respect to plant sequences
calculations looking at overall signatures such as the degraded core final state and the
fission product source term. A number of recommendations for model development
and various implications for plant studies have been identified.
A.
OBJECTIVES AND SCOPE
The THENPHEBISP 2-year thematic network started in December 2001, and
was concerned with OECD/CSNI International Standard Problem 46, itself based on
the Phebus FPT1 core degradation/source term experiment. The aim was to assess the
capability of computer codes to model in an integrated way the physical processes
taking place during a severe accident in a pressurised water reactor, from the initial
stages of core degradation, the fission product transport through the primary circuit
and the behaviour of the released fission products in the containment. The ISP-46
provided the first opportunity to assess the capability of integrated severe accident
analysis codes in four different areas, corresponding to the phases of the Phebus FP
experiments [1], namely: (a) fuel degradation, hydrogen production, release of fission
products and of structural materials; (b) fission product and aerosol transport in the
circuit; (c) containment thermal hydraulics and aerosol physics; and (d) iodine
chemistry in the containment. Participants were encouraged to submit integrated
calculations, but detailed calculations for individual phases were also welcome.
Choice of noding schemes and model parameters corresponding to plant calculations
(base-case) was strongly encouraged, while more detailed optional (best-estimate)
studies were also accepted. This formalism made it easier to draw conclusions
regarding plant calculations, and to identify ‘user effects’.
B.
WORK PROGRAMME
The work programme involved three main phases.
1) Preparation, resulting in the Specification Report for the ISP [2], during this phase
a draft was circulated so that feedback from participants on data requirements and
modelling needs could be obtained;
2) Calculation, including the organisation of an Intermediate Comparison Workshop,
where participants presented their contributions and the coordinators gave their
first impressions, at this stage a plan for the detailed analysis was agreed and a
schedule arranged for revised contributions as needed;
3) Comparison and Assessment, which included detailed analysis by the
coordinators, presentation of the draft Comparison Report at the Final Workshop,
and production of the final version of the Comparison Report taking into account
all comments received. Feedback from code developers was vigorously
encouraged, and valuable information was gained from them.
B.1 Preparation
The main task of this work package was to issue the specification report for the
ISP 46. An initial draft version was distributed to the partners and discussed at the
Preliminary Workshop in November 2001.
B.2 Intermediate Comparison Workshop
During the Intermediate Comparison Workshop, in October 2002, the
participants presented their contributions and the coordinators gave their first
impressions, at this stage a plan for the detailed analysis was agreed and a schedule
arranged for revised contributions as needed.
B.3 Comparison and Assessment
This phase corresponds to the detailed analysis by the co-ordinator of the results
submitted by the participants. This analysis was discussed at the final comparison
workshop in March 2003. An application workshop [3] was held in June2003 in order
to discuss the implications of the work performed on plant studies.
C.
WORK PERFORMED AND RESULTS
C.1 Specification Report
An initial draft version of the specification report was distributed to the partners
and discussed at the Preliminary Workshop in November 2001. The report covers the
objectives of the exercise, indicates the timescale, summarises the Phebus facility and
the test itself, indicates the boundary conditions and material data required, provides
references to the available results of the experiment, details what results are to be
provided by the participants, and invites participants to draw their own conclusions
with regard to needed model improvements and, importantly, to accident sequence
analysis in commercial power plants. Finally, concluding remarks are given.
Appendices provide additional data not present in the FPT1 Final Report, recommend
source terms at bundle exit (required for calculations considering the circuit only) as
well as circuit exit (required for calculations considering the containment only), and a
list of participants to the ISP.
The areas covered by the experiment, and therefore by the Standard Problem, are
fourfold:
1. Fuel degradation, hydrogen production, release of fission products, fuel, and
structural materials ('bundle' part of the ISP);
2. Fission product and aerosol transport in the circuit ('circuit' part of the ISP);
3. Thermal hydraulics and aerosol physics in the containment ('containment' part
of the ISP);
4. Iodine chemistry in the containment ('chemistry' part of the ISP).
Participants were encouraged to perform integral calculations covering all four
aspects of the exercise. However, the ISP was so organised that it was also possible
for participants to calculate any of the above phases in a stand-alone manner, using
detailed-level mechanistic codes that treat for example core degradation or
containment thermal hydraulics and aerosol physics on their own. To the latter end,
recommendations were made regarding the noding to be used in the analysis, for two
cases; a base case with discretisations similar to those that could be used in a reactor
study, and for an optional, more detailed, 'best estimate' case more typical of those
used in experimental interpretation. A number of participants have been able to
perform two sets of calculations for each code they chose to employ, so the effect of
fineness of noding could be examined. The submission for each code (not limited to
one per organisation) could consist of a base case and a best-estimate case for each of
which numerical data would be provided, accompanied by such sensitivity studies to
illuminate the results as the participant thinks fit. However the first set of calculations
was deemed more important
The sources of information for the ISP were, in descending order of priority:
 the Specification Report;
 the FPT1 Final Report;
 the FPT1 Data Book.
In addition, detailed data were provided in numerical form in supplementary files
in electronic form, for example source term data and experimental measurements such
as temperatures in the circuit which are not all provided in the Final Report.
C.2 Intermediate Comparison Workshop
The period for calculations lasted six months, during this period a
supplementary Workshop was held to clarify points arising. The Comparison
Workshop was held about one year after the start of the project, with participants
having had the opportunity to submit revised calculations in the meantime. The ISP
was well supported, with participation from 33 institutes, companies etc. in 23
countries and international organisations. The latter comprised EC-JRC, Austria,
Belgium, Bulgaria, Canada, Croatia, Czech Republic. France, Germany, Greece,
Hungary, Italy, Japan, Korea, Mexico, Russia, Slovenia, Spain, Sweden, Switzerland,
Turkey, UK and USA. The participating organisations included utilities, regulators
and their technical support organisations, research institutes and private engineering
consultancy companies, thus providing a good range of backgrounds to the technical
work. Fifteen different codes were used: ASTEC, ATHLET-CD, COCOSYS,
CONTAIN, ECART, FEAST, IMPACT/SAMPSON, ICARE/CATHARE, IMPAIR,
INSPECT, MAAP4, MELCOR, SCDAP/RELAP5, SCDAPSIM and SOPHAEROS,
of these 4 are integral codes (ASTEC, IMPACT/SAMPSON, MAAP4 and
MELCOR). For the base case, 47 calculations were received, with 21 for the optional
best-estimate version. Of the base case calculations, 14 were integral (at least 3
phases calculated)
During the Comparison Workshop, the participants have presented their results
and the co-ordinators their first impressions.
C.3 Comparison and Assessment
C.3.1
Representation of the facility
For the base case, a noding scheme was recommended in the specification report
The bundle is divided into 11 axial nodes and typically 3-5 radial rings, with normally
1 or 2 thermal hydraulic flow channels. The circuit is divided into 11 nodes, this
being the minimum considered necessary for an adequate calculation of deposition.
The containment model is simple, with 1 node for the main volume and 1 for the
sump, taking advantage of the well-mixed conditions. A typical nodding scheme is
given in figure1. For best-estimate calculations, noding density was increased by
typically a factor 2 or more, at the choice of the user.
C.3.2
Analysis of the results
The results were analysed in detail, comparing the results amongst each other and
with the FPT1 data. There was considerable scatter amongst the results obtained from
each code by different users, the ‘user effect’. To minimise this effect, representative
cases were selected where necessary, taking into account the quality of key output
variables, completeness and accuracy of the technical reports, and including code
developers where possible. This analysis led to an assessment of the main models in
each of the four areas considered. These are grouped below, in order of their
perceived adequacy. There was on the whole little significant difference between the
base and best-estimate cases, with at most a small improvement only in the results of
the latter cases, so conclusions could be drawn on the basis of the former.
C.3.3
Assessment of Codes and Models
The following phenomena/parameters are in general well simulated by the codes:
 Bundle – thermal response (figure 2, given adjustment of input nuclear
power and shroud thermal properties within experimental uncertainties),
hydrogen production (figure 3, including oxidation of relocated melt),
bundle final state material distribution (figure 4, given suitable reduction of
the bulk fuel relocation temperature from the ceramic value, in the longer
term a more mechanistic model is desirable), total release of volatile fission
products (figure 5);
 Circuit – total retention of fission products and structural materials (figure 6,
but after cancellation of errors);
 Containment – thermal hydraulic behaviour (as exemplified by average gas
temperature, pressure, relative humidity and condensation rate), depletion
rates (figure 7);
 Chemistry – models of the Ag/I reaction in the liquid phase are adequate for
FPT1 (this cannot be extended to other cases where the Ag is not so much in
excess with respect to I; due to the large excess of silver, in the experiment,
radiolytic production of gaseous iodine and dissociation of silver iodide did
not play an important role in the overall iodine behaviour).
The following phenomena/parameters were reasonably well simulated, but some
modelling improvement is desirable:
 Bundle – outlet coolant temperatures (overprediction), time dependence of
volatile FP release (figure 5, generally too fast a release at low temperatures,
e.g. for CORSOR-type approaches);
 Circuit – distribution of deposition in the circuit (underestimation in the
upper plenum where vapour condensation and thermophoresis are the
dominant mechanisms, overestimation in the steam generator hot leg where
the mechanisms are thermophoresis for all elements + vapour condensation
for I and Cd), noting that too coarse a noding leads to underestimation of
deposition;


Containment – relative importance of the two main depletion processes
(diffusiophoresis and gravitational settling), but it is hard to make firm
conclusions owing to the variability in the results;
Chemistry – no items identified.
The following phenomena/parameters were not well simulated and substantial
model development is necessary:
 Bundle – release of medium and low volatiles (e.g. tendency to calculate low
for Mo –figure 8, very high for Ba, reasonable order of magnitude for Ru
and U but considerable scatter), and of structural materials (Ag/In/Cd, figure
9, from the control rod where the basic process of evaporation from a molten
AIC pool is not captured, tin from the Zircaloy cladding);
 Circuit – iodine speciation and physical form;
 Containment – no items identified;
 Chemistry – gas phase reactions (figure 10), organic iodine reactions (figure
11), including production and destruction through radiolytic processes
(definition of optimum parameters for the modelling codes such as
adsorption velocity and desorption rate on/from painted surfaces, and the
facility to input the gaseous iodine fraction at containment entrance, are
recommended).
The good prediction of hydrogen production, generally near the upper bound of
the experimental uncertainty range, (+10%), is an important safety-relevant
conclusion. The good prediction of the bundle material distribution in the final states
requires a suitable reduction of the bulk fuel relocation temperature from the ceramic
value. Implications, in the short term and, in the longer term, the need for a more
mechanistic model will be discussed later on as integral aspects. Although the
structural materials do not themselves have radiological significance, they potentially
react with fission products, and their source terms are therefore needed for accurate
calculation of chemistry and transport in the circuit. A particular need is to saturate
the iodine reaction. The semi-volatile fission products are also of importance, either
because of their radio-toxicity and influence on the residual power, or by their
propensity to react with other fission products. The structural materials also form the
bulk of the aerosol mass, affecting the aerosol concentration and the agglomeration
processes.
Concerning the circuit, the overestimation of bundle outlet temperature cannot
fully explain the upper plenum results; its main effect is to displace the zone where
vapours nucleate. For some elements, part of the discrepancy in the deposition pattern
is due to the wrong prediction of the chemical form, and thus of its volatility; Cs is
generally calculated as a vapour at 700°C, whereas it was condensed in the
experiment. However, this is also not enough to explain the underestimation in the
upper plenum and overestimation in the steam generator rising line. Finding
explanations is presently part of the work performed in the frame of the interpretation
of Phebus-FP tests.
Care is needed in extrapolating the rather good results for the containment
directly to the reactor case, as the Phebus containment thermal hydraulics are
relatively simple, and the role of gravitational settling is overscaled, with a shorter
residence time of aerosols in the atmosphere and probably less effect of agglomeration
than for real plant.
Concerning the chemistry, the reaction of iodine with silver, forming non-soluble
silver iodide, dominates the phenomenology in the liquid phase. In the FPT1
conditions with a large excess of silver, the models behave sufficiently well, provided
enough silver is injected into the sump water. Gas phase chemistry is dominated by
early injection of gaseous iodine from the primary circuit, that will be discussed later
on as an integral aspect, and by inorganic iodine adsorption on the atmospheric paints
followed by organic desorption, together with destruction mechanisms. The results
were particularly contrasted, with a large scatter on the total gaseous iodine
concentration, and a fraction of organic iodine ranging from less than 10% to nearly
100%. Overall, they range from unreliable to very good (after tuning).
C.3.4
Computing Assessment
Key output variables for code assessment, such as those requested in the ISP,
were not always accessible to the user; these should be available as standard code
output. Graphics dumps to enable post-processing of results should be a standard
feature to aid in code assessment, to aid in detailed analysis. Computer (CPU) time
and timestep information should be available for plotting, to help optimise code use,
while temporal (timestep) and spatial (noding) convergence studies should always be
performed.
Platform dependence (both concerning computer hardware, and
sensitivity to compiler options) should be eliminated as far as possible.
C.3.5
Integral Aspects
This part considered the results with respect to the ‘key signatures’ of plant
sequence calculations, namely the core final state (relevant to in-vessel retention) and
the fission product source term. Good agreement for the bundle final state could be
obtained with suitable reduction of bulk fuel relocation temperature, but this is
unlikely to be representative for similar tests such as Phebus FPT2 and PBF SFD1.4
which show evidence for a higher temperature. Therefore, default values should not
be reduced on the evidence of FPT1 alone, and similar studies on other experiments
such as these are encouraged. In the longer term, a more mechanistic treatment of
bulk fuel relocation is desirable, and it seems unlikely that a simple temperature
criterion will suffice. More detailed model development, possibly needing new
separate-effects data, is therefore indicated, as the mechanisms involved are not clear
(effect of irradiated fuel, presence of Fe in the melt … ). In the meantime, plant
studies need sensitivity calculations on relocation temperature with the modelling in
its current state. Further studies are recommended on control rod failure (influence of
control rod materials) and the fuel rod oxide shell breach criterion (first movement of
U/Zr/O melt). A general integral point for the bundle is the need to take into more
account the interaction between bundle state and fission product/structural material
release, especially for the latter.
Concerning the source term, the accuracy of containment calculations in integral
treatments is sensitive, often highly, to results of previous stages (propagation of
uncertainties). Key features are the calculation of FP release from the bundle, and of
the structural materials Ag, In, Cd and Sn (the kinetics of release of these and of FPs
are as important as the final amounts); the temperatures at the entrance to the circuit,
which strongly influence the deposition pattern; while for those codes which calculate
the chemistry, the speciation is influenced by the calculated release. The release of
structural materials was often undercalculated or not calculated at all, leading to
undercalculation of total mass of aerosols, but this had only a weak impact on overall
retention in the reactor coolant system (RCS) and depletion in the containment.
Iodine speciation and physical form in the circuit was poorly predicted - no code
reproduced the observed gaseous iodine fraction in the RCS.
Given these limitations, it is hard for an integral calculation to predict well the
containment chemistry, however detailed the modelling for its phenomena – the
uncertainty on iodine release from fuel, aerosol transport in the RCS and behaviour in
containment is overwhelmed by uncertainties in chemistry. This has implications on
conduct of plant assessments, for example it may be better for the chemistry
calculations to be carried out in a stand-alone manner, using a range of sensitivity
studies, rather than as part of an integral calculation.
Finally, in determining the priorities for code improvement, attention should be
paid to finding the weakest link(s) in the chain of calculation which contribute most to
uncertainty in the assessment of risk - a ‘cost-benefit’ approach - is it a model itself or
the input to it?
C.3.6
Implications for Plant Studies
A strong user effect is visible in ISP-46, as in previous ones, therefore the user
effect in plant studies cannot be ruled out. A major objective must be to limit its
consequences on the quality of the study. It is recommended that this could be
achieved by: checking that previous training has been efficient; using adequate
procedures are used for controlling the results and peer reviewing, involving
experienced specialists in the field; and by checking that enough support is provided
by developers when necessary.
The quality of the models must also be taken into account. A number of necessary
improvements in codes and models have been identified above, the main ones being: a
better estimation of structural material release, especially for control rod elements and
tin from Zircaloy cladding, and of semi/low-volatile release; the possibility to take
into account the presence of gaseous iodine in the RCS; and the definition of optimum
parameters for iodine chemistry codes. As not all the necessary improvements can be
achieved in a short term, users have to be well aware of the validation status of codes
and must take into account their limitations when performing plant studies.
Severe accident codes are difficult to handle, and their validation is not complete.
They should not be used as “black boxes”, i.e. their results have to be interpreted,
according to the goal of the study for which they are used. Extensive training of new
users should be mandatory, and efficient quality assurance procedures for reactor
studies have to be used, involving review of the results by experienced experts not
directly involved in the work.
Finally, users should not trust automatically the results of their calculation, but
make a critical analysis! Do the results seem consistent and reasonable (“reality
check”)?
CONCLUSION
This integral ISP has enjoyed a wide and varied participation, with almost fifty
submissions. There was strong support for the bundle and circuit phases, moderate
for the containment aerosol phase, and least for the containment chemistry phase.
Follow-on studies may be proposed to continue the exercise, focussing on areas where
the greatest uncertainties remain.
Recommendations have been made on model development needs, which have
been agreed after discussions with the development teams of the major integral codes,
and progress is already being made in addressing some of the issues raised. Various
implications for plant studies have been identified, and the need for user experience
and training is emphasised [3]. Effective review procedures for reports are seen as
essential in ensuring the quality of such applications
1. Conclusions about the strengths and weaknesses of models and calculation codes
have been extensively discussed during the final workshop and agreed upon.
Recommendations have also been made by the co-ordinators and the code
developers about the desirable improvements to codes and models. The discussion
at the application workshop was more focussed on the recommendations for
optimum use of the codes in their present state, as not all the planned
improvements will be available in the near future. The conclusions from the
discussion are presented below.
2. Concerning the fuel degradation and the hydrogen release, the results are generally
good. The remaining weak point is that the early fuel relocation during the heat-up
phase (late phase) can only be reproduced by adjusting bulk fuel relocation
temperature to a suitable value. It was recognised that it is unlikely that such a
temperature would be valid for all the plant applications and situations. It was
agreed that the only solution was to proceed to sensitivity calculations, in order to
appreciate the effects. From the existing knowledge, a range of bulk fuel
relocation temperatures should be defined.
3. Concerning the releases from the core, the main difficulties are encountered for
semi-volatile fission products and structural material behaviour. It is hoped that
modelling improvements can be achieved in a reasonable timeframe for some of
the elements. In the meantime, the observed overestimation or underestimation
should be taken into account in the interpretation of the results. They should be
provided as a band more than as a precise value.
4. Concerning iodine chemistry, sensitivity studies should be performed as an usual
good practise. In addition, an important lesson from the integral ISP-46 is that
iodine calculations may suffer from propagation of errors coming from other
modules in integral calculations, as iodine chemistry in the containment is at the
end of the calculation chain. It was recommended to complement integral
calculations by iodine chemistry stand-alone calculations, as necessary. These
calculations should use boundary conditions, such as the silver amount or gaseous
iodine from the RCS, differing from the ones calculated by integral codes.
More generally speaking, awareness of the codes' strengths and weaknesses is
essential in interpreting their results and defining and running the necessary sensitivity
and additional calculations. Guidance should be provided by the code developers by
indicating, for instance, a range of variation for uncertain parameters, as well as bestestimate values
REFERENCES
[1]
. M. Schwarz, B. Clement A.V. Jones: Applicability of Phebus FP results to
severe accident safety evaluations and management measures, Nuclear Engineering
and Design 209 (2001) 173–181
[2]
Haste, T. et al., “Specification of International Standard Problem ISP-46
(Phebus FPT1)”, IRSN Note Technique SEMAR 02/05, 2002 and 03/05, 2003.
[3]
B. Clément, T. Haste, "Thematic Network for a Phebus FPT-1 International
Standard Problem - Minutes of the Application Workshop", CR SEMAR 2003/037,
2003
FIGURES
+5.000
40010
+1.7015
35401
35402
+1.215
11
356
353
10
35602
40008
+0.682
+1.0915
356
12
353
40009
+1.7015
354
11
354
+1.5905
35301
+1.295
35601
357
35701
40006
+0.147
35201
352
352
-1.2965
35101
-0.240
357
9
-1.205
12
358
358
351
-1.440
12
351 8
359
365
-3.5155
25004
CV300
-3.605
CV250
25002
6
-3.5445
250
25003
5
30001
-3.50955
-2.2825
305
CV303
301
6
-3.55045
30301
30302
CV305
303
-3.545
30303
30502
20302
20301
-5.880
CV203
4
-2.419
36502
G
-2.624
40011
-6.8735
-6.905
CV201
CVnnn x
3
150
15015
15014
15013
15012
15011
15010
15009
15008
15007
15006
15005
15004
15003
15002
nnnnn
CV150 2
-6.987
-7.037
-7.137
-7.237
-7.337
-7.437
-7.537
-7.637
-7.737
-7.837
-7.937
-7.987
-8.041
R2
100R1
CV100
-8.161
1
nnn
+x.xxxx
Control Volume No.nnn, CVTYPE x
Heat Structure No.nnnnn
Flow Path No.nnn
Altitude in [m]
-3.157
40003
40002
-6.773
10002
36501
201
-6.705
20105
20104
20103
20102
20101
CV365
360
30501
C
203
-2.413
36503
14
36002
25001
-5.806
-2.389
CV360
7
-2.154
-2.154
13
-3.515
15
CV400
-2.378
-3.515
20
600
40004
35901
35902
36001359
-2.740
CV600
35801
-3.605
-3.755
40001
60001
COR Component: Fuel + Cladding
COR Component: Top & Bottom Ends of Fuel Rod
COR Component: SS Structures (Spacer Grids, Core Support Plate)
COR Component: NS Structures (Control Rod Cladding & Guide Tube, Stiffeners)
COR Component: NS Structures (Control Rod Poison)
PHEBUS FPT-1 Nodalization Prepared by UJV Řež for ISP-46
MELCOR 1.8.5 dsp fpt-1v12.mg5 July 2, 2002
-8.361
10001
Figure 1: Typical nodding scheme for a MELCOR calculation
Figure 2 : Calculated and measured fuel Figure 3 : Calculated and measured
temperature (measurement in bold dashed hydrogen production (measurement in
line)
bold dashed line)
-4.500
-5.000
Figure 4 : Calculated and measured axial Figure 5: Calculated and measured release
material distribution (measurement in
of iodine (measurement in black bold
black)
dashed)
0.70
Cesium fractional retention
0.60
0.50
0.40
0.30
0.20
0.10
ea
n
1
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t
0.00
Figure 6 : calculated and measured cesium retention in the circuit for different users of
the same code (experiment on the left, mean calculated value on the right)
Figure 7 : Calculated and measured aerosol depletion rate in the containment (bestestimate calculations)
Figure 8 : calculated and measured release Figure 9 : calculated and measured release
of molybdenum (fission product)
of silver (control rod material)
( m o le s )
a tm o s p h e re
in
io d in e
a m o u n t o f g a s e o u s
1 .E -0 4
1 .E -0 5
1 .E -0 6
A E 1 IT O T
E C 3 IT O T
E C 3 B E IT O T
E F 1 IT O T
N P 1 IT O T
IP 3 IT O T
I2 + O I E X P
1 .E -0 7
1 .E -0 8
0
50000
100000
150000
200000
250000
300000
350000
tim e (s )
Figure 10: Calculated and measured total gaseous iodine concentration in the
containment’s atmosphere (measurement: red squares)
1
0.9
fraction of organic iodine
0.8
0.7
AE1 AMOIATM/ITOT
EC3 OI/ITOT
EC3 BE OI/ITOT
EF1 OI/ITOT
NP1 OI/ITOT
IP3 OI/ITOT
OI/ITOT EXP
GR4 OI/ITOT
GR3 OI/ITOT
0.6
0.5
0.4
0.3
0.2
0.1
0
0
50000
100000
150000
200000
250000
300000
350000
time (s)
Figure 11: calculated and measured organic iodine fraction in the containment’s
atmosphere (measurement: red squares)