SCWR Fuel Rod Design Requirements
Design Limits Input for Performance Evaluations
H. Garkisch, Westinghouse Electric Co.
VG .1
SCWR Fuel Rod Design Requirements
- Overview • The primary function of the fuel rod is to generate and
transfer heat to the reactor coolant.
• A second function of the fuel is to contain the fuel and
the fission products and provide a barrier against
coolant contamination with fission products. For this the
structural integrity of the fuel rod must be maintained in
compliance with applicable requirements.
• The SCWR operating conditions exceeds the current
experience with LWRs and LMFBRs, thus specific
criteria must be developed.
VG .2
SCWR Fuel Rod Design Requirements
- Overview • Generic Considerations for SCWR developed during 1st year
of NERI program:
– Compile NRC SRP criteria for nuclear reactors
– Identify main areas of difference between SCWR and standard
criteria
– Identify simplified requirements for preliminary analysis
• Proposal for specific design criteria/fuel rod failure modes
developed and submitted to INEEL during 2nd year
– Develop detailed review of NRC SRP Criteria
– Develop specific fuel failure modes and design criteria for
SCWR
VG .3
Design Criteria: LWR experience
• Fuel Design requirements for LWRs are defined in the NRC Standard
Review Plan to satisfy 10CFR50 GDC10
“The objectives of the fuel system safety review are to provide assurance
that
(a) the fuel system is not damaged as a result of normal operation and
anticipated operational occurrences,
(b) fuel system damage is never so severe as to prevent control rod
insertion when it is required,
(c) the number of fuel rod failures is not underestimated for postulated
accidents, and
(d) coolability is always maintained.”
VG .4
Design Criteria: LWR experience
ASME Code Sec. III
(Article NB-3112-3)
Normal:
Assumed Probability
of occurrence
(per Reactor Year)
Acceptance Criteria
1
no significant loss of effective
lifetime
1 to 3 x10-2
no significant loss of effective
lifetime
3x10-2 to ~10-4
Accumulated cladding stress and
strain below rupture limit
System startup/shutdown,
design range operation, hot
standby.
Upset:
Abnormal incident not
causing a forced outage or
causing a forced outage for
which the corrective action
does not include any repair of
mechanical damage.
Emergency:
Infrequent incident requiring
shutdown or correction of the
condition or repair of damage
of the system; no loss of
structural integrity.
Faulted:
Postulated events and
consequence where integrity
and operability may be
impaired to the extent that
consideration of public health
and safety are involved.
(Integrity limit, above this limit
fuel cladding can fail)
~10-4 to ~10-7
No loss of coolable geometry
caused by the fuel rod.
 No clad melting and
relocation
VG .5
Design Criteria for SCWRs
Table 4 Design Basis acceptance criteria for Fuel System Damage
Paragraph
(a)
Acceptance Criteria Description (from SRP section 4.2-II-A-1)
Stress, strain, or loading limits for spacer grids, guide tubes, thimbles, fuel
rods, control rods, channel boxes, and other fuel system structural members
should be provided.
Stress limits that are obtained by methods similar to those given in Section III
of the ASME Code are acceptable. Other proposed limits must be justified.
(b)
The cumulative number of strain fatigue cycles on the structural members
mentioned in (a) above should be significantly less than the design fatigue
lifetime, which is based on appropriate data and includes a safety factor of 2
on stress amplitude or a safety factor of 20 on the number of cycles.
Other proposed limits must be justified.
(c)
Fretting wear at contact points on the structural members mentioned in
paragraph (a) above should be limited. The allowable fretting wear should be
stated in the Safety Analysis Report and the stress and fatigue limits in
paragraphs (a) and (b) above should presume the existence of this wear.
(d)
Oxidation, hydriding, and the buildup of corrosion products (crud) should
be limited. Allowable oxidation, hydriding, and crud levels should be
discussed in the Safety Analysis Report and shown to be acceptable.
These levels should be presumed to exist in paragraphs (a) and (b) above.
Hydriding is not applicable to non_Zirconium alloy cladding
Dimensional changes such as rod bowing or irradiation growth of fuel rods,
control rods, and guide tubes need not be limited to set values (i.e., damage
limits), but they must be included in the design analysis to establish
operational tolerances.
(e)
(f)
Fuel and burnable poison rod internal gas pressures should remain below
the nominal system pressure during normal operation unless otherwise
justified.
(g)
Worst-case hydraulic loads for normal operation should not exceed the
holddown capability of the fuel assembly (either gravity or holddown springs).
A fuel assembly design issue
Control rod reactivity must be maintained. This may require the control
Rods to remain watertight if water-soluble or leachable materials (e.g., B4C)
are used.
A fuel assembly design issue, not a fuel rod issue
(h)
VG .6
Fuel Rod Failure Modes
 Fracture induced by rod pressure and fuel cladding mechanical interaction,
assisted by irradiation assisted stress corrosion cracking and stress corrosion
induced embrittlement of the cladding.
 Creep rupture burst due to over-pressure or sustained stress induced by fuelcladding differential thermal expansion.
 Cladding failure induced by fuel cladding mechanical interaction during steady
state and then transient operation at high burn-up.
 Cladding fatigue failure, an unlikely failure mode for a reactor in base load
operation.
 Excessive cladding growth and swelling exceeding the functional constraints of
the fuel assembly, an unlikely failure mode in a thermal flux operating environment.
 Cladding corrosion, which thins the cladding and increases cladding temperatures,
is unknown in a supercritical steam environment. A thick corrosion layer on the
cladding increases cladding and fuel temperatures and can induce failure. Due to the
high system operating pressure cladding collapse under external differential
pressure.
 Cladding collapse under external differential pressure
VG .7
Fuel Rod Design Criteria Basis
 NUREG-0800, STANDARD REVIEW PLAN, NUCLEAR REGULATORY
COMMISSION
 Criteria Summarized in Three Tables (Fuel Rod Damage, Fuel Rod Failure,
Core Coolability).
 ASME B&PV Code Section III Article NH-3000
 The fuel rod cladding is not covered by the code.
 However, by applying criteria based on the code approach, a design finds
acceptance without further justification.
VG .8
SCWR Criteria Issues
• At super critical steam temperature (> 800 0F) ASME
high temperature design rules apply;
• Creep, irradiation and thermal creep must be
checked or considered;
• The high system pressure requires check of
buckling and stress at depressurization.
VG .9
ROD INTERNAL PRESSURE CRITERION
The rod internal pressure of the lead rod in the reactor shall not
exceed the pressure that could:
• Cause the diametral gap between the fuel and the cladding to
increase due to steady state operation, cause ballooning and
affect the coolant flow;
• Exceed the rupture pressure of the cladding (if known);
• Local overheating of the cladding.
VG .10
CLADDING STRESS CRITERION
Criterion, Basis, Implementation
Cladding Stress Criterion: Where creep is significant, the ASME B&PV Code
Section III Article NH-3000 specifies that the strain limiting criteria, rather than
stress limiting criteria are applied. However, simplified methods can be used to
establish conservative limits for stress.
yield stress
Sm = min. of
stress rupture
St = min. of
|
|
|
|
2/3 Sy at ambient (room) temperature
2/3 Sy at service temperature
1/3 Sult at ambient (room) temperature
1/3 Sult at service temperature
| 100 % of the stress to cause 1% strain.
| 80% of the stress to initiate tertiary creep,
| 67% of the minimum stress to cause rupture
VG .11
CLADDING STRESS CRITERION (Condition I)
Criterion, Basis, Implementation
Cladding Stress Criteria: The time independent stress limits for the load categories are as
•
•
follows:
With: Pm, primary membrane stress (dP across the cladding and PCI); Pl primary local stress
(stress raiser due to pellet cracking and bambooing); Pb, primary bending stress (bowing or PCI
gradients); and Q, secondary stress (thermal stresses)
• Sin (Pm)
< 1.0 Sm
• Sin (Pm + Pl)
< 1.5 Sm
• Sin (Pm + Pl + Pb)
< 1.5 Sm
• Sin (Pm + Pl + Pb+ Q)
< 3.0 Sm
The stress adder Q is included to assure that the transient thermal stresses do not exceed
stresses which could exhaust the deformation capability of materials
Typical fuel performance codes (like FRAPCON) calculate Pm and Q, not Pl and Pb.
 t l  <D


T
D_al

l 1 
q
The time dependent stress limit is as follows:
Basis:
•
•
•

Table 3, Paragraphs a; SRP II, A-1-a
ASME B&PV Code, Section III, Division 1, Article NH-3000
ASME B&PV Code, Section III, Article III-2000, III-2100
VG .12
CLADDING STRAIN CRITERION (Condition I and II)
• The total permanent uniform strain shall not exceed:
– 1 % membrane strain (limiting)
– 2 % bending strain
– 5 % local strain
•
•
•
The intent of this requirement is to limit cladding damage due to slow rate strain
accumulation at which the stress does not reach the stress limit (yield stress). The
clad loading mechanism is the rod internal differential pressure with the system
pressure and clad straining by the pellet expansion and PCI.
A bending strain and local strain are not calculated by FRAPCON and the limits are
not applied at this time.
Basis:
– Table 3, Paragraph a, and SRP II, A-1-a
– ASME B&PV Code Section III Article NH-3000
VG .13
FUEL TEMPERATURE CRITERION
• During Condition I and Condition II events the peak kw/ft fuel rods
shall not exceed the UO2 melting temperature with 95%/95%
probability and confidence level.
• The un-irradiated fuel melting temperature is 2805 0C. It reduces
by 58 0C for every 10000 MWd/MTU burnup. For rods with
Gadolinia the melting temperature is reduced 3.75 0C for each w/0
Gadolinia oxide.
– For preliminary design purposes of high burnup fuel limit the
maximum fuel temperature to < 2600 0C
VG .14
CORROSION AND FATIGUE CRITERIA
• Corrosion:
– Cladding corrosion reduces the effective thickness of the
cladding, decreases the effective thermal conductivity of the
cladding and thus increases the cladding and fuel
temperatures.
– In absence of any data a conservative linear increase of
the corrosion layer of 0.1 mm (4 mil) shall be assumed.
• Cladding Fatigue:
– Cumulative number of strain cycles shall be less than the
design fatigue lifetime with appropriate margins.
» The cumulative number of strain cycles shall be less than the design
fatigue lifetime, with a safety factor of 2 on stress amplitude and a
safety factor of 20 on the number of cycles
VG .15
CLAD COLLAPSE AND ROG GROWTH
• Clad Collapse:
– The clad shall be free standing at BOL (before densification)
– No clad collapse in the gas plenum region
– No clad collapse into gaps between pellets
– At high temperatures elastic, plastic and potential creep deformation must be
considered as well as the tube ovality.
• Rod Growth:
– Fuel rod length changes due to irradiation effects and differential
thermal expansion shall not cause interference with the fuel
Assembly Structure
– From the fuel cladding expansion, thermal swelling and creep the total length
change of the cladding can be estimated with the axial temperature and axial growth
profile as input. Similar calculations are required to estimate the growth of the
assembly structure. The differential between both must show a gap between the rod
length and the assembly structure.This evaluation is a critical design input because it
determines the assembly length.
VG .16
OTHER CRITERIA
• End Plug Weld Stress Criterion
• Fuel Rod Length Change Criterion
VG .17
MA956 is not isotropic
VG .18
Design Criteria and Evaluation
• Available material properties for MA956, other ODS
and high Nickel materials are spotty and not
consistent
• Irradiated properties are largely missing
• Guesses of some properties and limits are required
• A systematic compilation of material properties
required
VG .19