LD-82-003, Forwards Responses to Fuel Performance Analyses Identified in CESSAR SER Confirmatory Item 3.Encl Transmitted for NRC Review of FSAR
| ML20039G467 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 01/11/1982 |
| From: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| LD-82-003, LD-82-3, NUDOCS 8201180279 | |
| Download: ML20039G467 (14) | |
Text
.
C-E Power Systems Tel. 203/688-1911 Combustion Engineenng. Inc.
Tetex 99297 1000 Prospect Hill Road Windsor, Connecticut 06095 POWER H SYSTEMS Teaket I;o.:
ST!;-50 147CF January 11, 1982 LD-82-003 in 4
UCSiVCT) g.
Mr. Darrell G. Eisenhut, Director gkl 61984 'S Division of Licensing D
7 3k g U. S. ?!uclear Regulatory Comnission g
Washington, D. C.
20555 m;c at
Subject:
CESSAR SER Confirmatory Item (3) y Rcrerence: Letter A. E. Scherer to D. G. Eisenhut, dated October 8, 1901, LD-81-069
Dear Mr. Eisenhot:
Transmitted harewith are responses to fuel performance analyses identified as confirmtory item (3) in Section 1.8 of the CESSAR Safety Evaluation Report.
Sevaral resp:nses include revised sections to r,he Section 21. 2 question respom e provided per the Reference.
The enclosure is transmitted for the Staff revicw of the CESSAR FSAR.
Tne enclosed infornation is intended to complete the C-E input that the Staff needs to close out confirmatory iten (3) concerning fuel performcnce analyses.
If I can be of any additional assistance in this matter, please contact me or Mr. G. A. Davis of my staff ct (203)638-1911, Extension 2003 Very truly yours, COMEUSTI0t! E!:GIliEERI!!G, I!!C.
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A. E. Scherer Director
!!uclear Licensing AES:et%
Enclosure ec:
C. I. Grices f
820i180279 020111 PDR ADOCK 05000470 E
CESSAR FUEL
. PERFORMANCE ANALYSES 1.
CEA Fretting Wear (SER Sections 4.2.1.1 (d):
See the attached marked-up sheets of the Section 4.2 re-write provided per the Reference 2.
CEA Axial Growth (SER Sections 4.2.1.1 (g) and 4.2.3.1 (g)):
See the attached marked-up sheets of the Section 4.2 re-write provided per the Reference 3.
Fuel Rod Mechanical Fracturing (SER Sections 4.2.1.2 (g) and 4.2.3.2 (g)):
See the attached marked-up sheets of the Section 4.2 re-write provided per the Reference 4.
Fragmentation of Embrittled Cladding - Non LOCA Coolability Criteria (SER Sections 4.2.1.3 (a) and 4.2.3.3 (a)):
See the attached marked-up sheets of the Section 4.2 re-write provided per the Reference 5.
Fuel System Design Stress (SER Section 4.2.3.1 (a)):
The calculated stress: results for CESSAR fuel assenblies,.f ael. rods, bornabis poison rods, upper end fitting springs, and CEA, will be verified to meet the.~ design stress reqdirements prior totinitial core criticality.
l 6.
Fuel System Design Strain (SER Section 4.2.3.1 (b)):
The calculated strain results for CESSAR fuel assemblies, fuel rods, burnable poison rods, upper and fitting springs, and CEAs will be verified to meet the design strain requirements prior to initial core criticality.
7.
Fuel System Strain Fatigue (SER Section 4.2.3.1 (c)):
The calculated strain fatigue results for CESSAR fuel assemblies, fuel rods, burnable poison rods, upper end fitting springs, and CEAs will be verified to meet the design strain fatigue requirements prior to initial core criticality.
8.
Fuel Assembly Lift-off (SER Section 4.2.3.1 (1)):
See the attached marked-up sheets of the Section 4.2 re-write provided per 4
the Reference
Reference:
Letter A. E. Scherer to D. G. Eisenhut, dated October 8, 1981, LD-81-069.
Control' Element Assembly The net unrecoverable circumferential strain shall not exceed 1% of the CEA clad (Inconel Alloy 625) considering the effects of pellet swelling and clad creep.
C.
Strain Fatigue Cumulative strain cycling usage, defined as the sum of the ratios of the number of cycles in a given effective strain range (ac) to the permitted number (N) at-that range, as taken from Figure 4.2-2, will not exceed 0.8.
The cyclic strain limit design curve applicable to Zircaloy cladding is shownonFigure4.2-2andg)basedupontheMethodofUniversalSlopes developed by S. S. Manson and has been adjusted to provide a strain cycle margin for the effects of uncertainty and irradiation. The resulting curve has been compared with known data on the cyclic loading of Zircaloy and has been shown to be conserv Specifically, it encompasses all the data ofO'DonnellandLanger.ge.
D.
Fretting Wear j A pn1 Y Fretting wear of any component within the Fuel System Design would be evaluated y
on the basis of the stress and strain criteria listed previously in this section.
E.
Oxidation and Crud Buildup During normal operating and upset conditions (Conditions I and II) oxidation and crud buildup have not been observed as a problem.
Therefore, no specific criteria has been defined.
F.
Rod Bowing Experience has proven that any specific criterion on allowable deflections (bowing), with respect to the effects which such deflections might have on thermal-hydraulic perfonnance, is not necessary beyond the initial fuel rod positioning requirements required of the grids.
This variation in spacing is accounted for in thermal-hydraulic analysis through the introduction of hot channel factors in calculating the maximum enthalpy rise in calculating DNBR.
This adjustment is called the Pitch, Bowing, and Clad Diameter Enthalpy Rise Factor, which is conservatively applied to simulate a reduced flow area along the entire channel length.
The value of this factor is given in Table 4.4-1 and its application is discussed in Section 4.4.
The subject of fuel rod bowing is discussed in Reference 53.
Further infonnation is provided on rod bowing in the response to CESSAR Round 1 question 492.3.
3 There is no specific limit on lateral fuel rod deflection for structural integrity considerations except which is brought about through application of cladding stress criteria.
The absence of a specific limit on rod deflection is justified because it is the fuel assembly structure, and not the individual fuel rod, that is the limiting factor for fuel assembly lateral deflection, t
1 Insert A or any type of wear i
e L_
-+.
G.
Axial Growth j
The fuel assembly is designed on the basis of maintaining an adequate shoulder gap between the fuel (and poison) rods and the upper end fitting, and sufficient clearance between the fuel assembly and the upper guide structure throughout the expected life (burnup) of the fuel assembly.
Spacer grids are designed to allow axial rod growth withoutintroducing significant restraints that would potentially enhance fuel rod bowing.
In general, there is no criterion for axial growth per SE; however, adequate clearances are maintained on-fuel. assenbly structual components, fuel and poison rods, control element assemblies, neutron sources and in-core instruments to insure functionability for their respective lifetimes.
'44 resert B -*
HT Fuel and Poison Rod Pressure Fuel and poison rod internal pressure increases with increasing burnup and toward end-of-life the total internal pressure, due to the combined effects of the initial helium fill gas and the released fission gas, can approach values i
comparable to the external. coolant-pressure.
The maximum predicted fuel and poison rod internal pressures will be consistent with the following criteria.
1.
The primary stress in the cladding resulting from differential pressure will not exceed the stress limits specified earlier in this section.
l 2.
The internal pressure.will not cause the clad to creep outward from the pellet surface while operating at the design peak linear heat rate for i
nonnal operation.
In detennining compliance with this criterion for fuel rods, internal pressure is calculated for the peak power rod in l
'the reactor, including accounting for the maximum computed fission gas release.
In addition, the pellet swelling rate (to which the calculated clad creep rate is compared) is based on the observed swelling rate of " restrained" pellets (i.e., pellets in contact with clad), rather than on the greater observed swelling behavior of pellets which are free to expand.
The criteria discussed above do not limit fuel rod internal pressure to values less than the primary coolant pressure, and the occurrence of positive differential pressures would not adversely affect normal operation if appropriate criteria for cladding stress, strain, and strain rate were satisfied.
I.
Fuel Assembly Liftoff The combination of the fuel assembly wet weight andholddown spring force must maintain a net downward force on the fuel assembly during all nonnal and anticipated transient flow and temperature conditions.
J.
Control Material Leaching A specific criteria for control material leaching has not been specified since leaching has not been observed as a problem.
In addition, adherance to previously specified stress and strain limits provide assurance that cladding integrity is maintained, therefore exposure of the control material to the coolant is avoided.
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Insert B co >. J~
Adequate clearance is evaluated on the basis that under adverse design conditions, a clearance of greater than zero shall be maintained.
l l
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(Regulatory Cuide 1.77, May 1974). The threshold for fuel melting is assumed to be 4940 F (250 calories per_ gram at the fuel centerline) for all fuel rod burnups. This temperature is obtained by reducing the melting point suggested in Regulatory Guide 1.77,
50*F) to account for a maximum expected burnup of about 35~,000 mwd /T.
E.
Pellet / Cladding Interaction (PCI)
Damage criteria previously specified for cladding strain (1%) and fuel-pellet overheating minimize the probability of failure due to pellet-cladding interaction.
F.
Cladding Rupture In ECCS analysis, an empirical model is used to predict the occurrence of cladding rupture. The failure temperature is expressed as a function of differential pressure across the cladding wall.
Predictions of cladding rupture are used in ECCS analysis to show that.the core geometry remains amenable to cooling.
There-fore, there are no specific design limits associated with cladding rupture.
The rupture model is a portion of the ECCS evaluation model and is discussed in Section 6.3.3.1 of the FSAR.
4.2.1.3~
Fuel Coolability Criteria Fuel coolability is maintained such that' continued removal of decay heat is ensured for all anticipated operational occurrences and accidents.
Except as described below, the need for specific criteria for core coolability is precluded by the design basis damage criteria discussed above.
The maintenance of core coolability is discussed'further in Section 4.2.3.3.
(a)
Fragmentation of Embrittled Cladding For ECCS analysis, limits of 2200*F on peak cladding temperature and 17% on maximum cladding oxidation are used (see Section 6.3.3.1).
m s.,se<+ n -r (b) Violent Expulsion of Fuel Material For the CEA ejection event, the radially averaged energy deposition at the hottest. axial location is limited to a value less than 280 cal /gm (Regulatory Guide 1.77, May 1974) to prevent fuel rod dispersal due to the rapid reactivity insertion.
(c) Cladding Ballooning and Flow Blockage For ECCS analysis, maintenance of a coolable core geometry is insured by consideration of cladding swelling and rupture as an integral part of the ECCS performance analysis.
C.
Structural Damage from External Forces The fuel assembly damage criteria (Section 4.2.1.1) are used in the analysis of LOCA and seismic conditions to insure that structural integrity and function are maintained.
AMs,cutc "
Insert C Fracture stress limits for fuel rods shall be in accordance with page 9-6 of CENPD-178-P, Rev. 1.
Based on. previous experience, theilimiting clad stress conditions from externally applied forces results from seismic and/or LOCA conditions (See Section 4.2.3.la).
For the core-plate motion on the San Onofre 16 X 16 fuel system design, it was shown that the resultant clad stresses were within design limits.
ll d
I e
i f
Insert D The Standard Review Plan, Section 4.2, includes NRC review guidance to assure that core coolability is always maintained. This implies that the fuel rods should not defom to.such an extent that major flow blockages or redistributions That is, they should retain their rod-like configuration and their occur.
position in the core.
It is not considered credible that a rod ~ could lose its coolable geometry without losing its hermeticity. Therefore, if a rod does not lose its hematicity, it t
is considered to remain coolable. The Specified Acceptable Fuel Design Limits (SAFDL) are detemined to assure that'the hermeticity of the fuel will be maintained as long as SAFDLS are not violated. All anticipated operating occurrences remain within the SAFDLs so there will be no loss of hermeticity and thus no concern j
about coolability for these events.
l Among the non-LOCA accidents, the CEA ejection event has a criterion to limit the total fuel enthalpy to 280 cal /gm. Regulatory Guide 1.77 ascribes this limit to the prevention of " prompt pin rupture" which "could econceivably disarrange the reactor core". This requirement was met by a significant margin as described in section 15.4.8.
Fort.all non-LOCA accidents loss of hermeticity is assumed to occur only if-departure from nucleate boiling (DNB) is predicted. Although the onset of DNB is con-servatively assumed to result in fuel failure through loss of hemeticity, there is a large body of evidence (see for example References 74 and 75) that demonstrates that such failures do not occur until fuel rods operate in the post-DNB regime for an extended period of time. Moreover, the failed rods in these tests maintained coolable geometry.
Reference 74, for example, provides the time-temperature boundary between survivors and failures as observed in the PBF (Power-Burst-Facility) experiments. The comparison of this boundary with the expected clad time-at-temperature calculations for 1imiting SAR events given in Table 4.2-4 shows that there is a very large margin between the fail-survive boundary and the calculated values. The time-at-temperature values in the table are based on the assumption that the maximum temperature reached in the event lasts for the period that the hot pin remains in DNB which is clearly an upper bound on the actual value.
On this basis, these limiting events remain far from the conditions that might produce a loss of fuel pin henneticity. Therefore, core coolability is main-tained by an even greater ergin.
I TABLE 4.2-4 Maximum Maximum Event Time in Clad DNB Temperature CEA 15 sec 1500 0F Ejection Locked Rotor 25 sec 2000 0F with Loss of Offsite Power Steam Line Break 30 sec. 2000 0F
\\
l l
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1
I l
63.
" TORC Code:
A Computer Code for Determining the Thermal Margin of a Reactor Core," Combustion Engineering, Inc., CENPD-161-P, (Propri-etary) July 1, 1975.
64.
Joon, K., " Primary Hydride Failure of Zircaloy Clad Fuel Rods,"
ANS Transactions, Vol 15, No.1.
65.
" Application of Zircaloy Inadiation Growth Correlations for the Calculation of Fuel Assembly (and Fuel Rod Growth Allowances" Supplement 1 to CENPD-198P, Proprietary), December 1977.
- 66. Marlowe, Mo.
0., "High Temperature Isothennal Elastic Moduli of UO " Journal of Nuclear Materials, Vol. 33 (1969), pages 242-244.
2 67.
Pickman, D.
0., " Properties of Zircaloy Cladding," Nuclear Engineer-ing and Design, Vol 21, No. 2 (1972).
68.
" Standard Specification for Sintered Uranium Dioxide Pellets,"
ASTM Standard C776-76, Part 45 (1977).
69.
" Data Transmitted for Review of SCE Fuel Structual Integrity Under Faulted Conditions," Combustion Engineering, Inc., CEN-151(S)-P, March,1981.
70.
"CE Thenno-Structual Fuel Evaluation Method, Combustion Engineering, Inc., CENPD-179, April 1976.
71.
" Introduction to Ceramics," W. D. Kingery, John Wiley & Sons,1960 pp 469-471.
72.
Simnad, M.
T., Meyer, R. A.; "Be0 Review of Properties for Nuclear Reactor Applications," Proceedings of the Conference on Nuclear Appli-cations of Nonfissionable Ceramics, May 9-11, 1966, pp. 193-206.
73.
Thorne, R.
P., Howard, V. C.; " Changes in Polycrystalline Alumina by Fast Neutron Irradiation," pp. 441-445, Proceedings of the British Ceramic Society, No. 7, February 1967.
R. VanHouten, " Fuel Rod Failure as a Consequence of Departure fbm 74.
Nucleate Boiling or Dryout", USNRC, NUREG-0562, June,1979.
75.
S. Levine to E. G. Case and RT B. Minogue, "Research Information Lett-
- 17 PBF Single Rod PCM Test Results", NRC Memorandum, May 5,1978.
I l
E
r G.
Axial Growth The fuel assembly is designed on the basis of maintaining an adequate shoulder gap between the fuel (and poison) rods and the upper end fitting, and sufficient clearance between the fuel assembly and the upper guide structure throughout the expected life (burnup) of the fuel assembly.
Spacer grids are designed to allow axial rod growth withoutintroducing significant restraints that would potentially enhance fuel rod bowing.
In general, there is no criterion for axial growth per SE; however, adequate clearances are maintained on fuel assembly structual components, fuel and poison j
rods, control element assemblies, neutron sources and in-core instruments to insure functionability for their respective lifetimes.
H.
Fuel and Poison Rod Pressure Fuel and poison rod internal pressure increases with increasing burnup and toward end-of-life the total internal pressure, due to the combined effects of the initial helium fill gas and the released fission gas, can approach values comparable to the external. coolant pressure.
The maximum predicted fuel'and poison rod internal pressures will be consistent with the following criteria.
1.
The primary stress in the cladding resulting from differential pressure will not exceed the stress limits specified earlier in this section.
2.
The internal pressure.will not cause the clad to creep outward from the pellet surface while operating at the design peak linear heat rate for nomal operation.
In determining compliance with this criterion for fuel rods, internal pressure is calculated for the peak power rod in the reactor, including accounting for the maximum computed fission gas release.
In addition, the pellet swelling rate (to which the calculated clad creep rate is compared) is based on the observed swelling rate of " restrained" pellets (i.e., pellets in contact with clad), rather than on the greater observed swelling behavior of pellets which are free to expand.
The criteria discussed above do not limit fuel rod internal pressure to values less than the primary coolant pressure, and the occurrence of positive differential pressures would not adversely affect normal i
operation if appropriate criteria for cladding stress, strain, and I
strain rate were satisfied.
I.
Fuel Assembly Liftoff
(
The combination of the fuel assembly wet weight andholddown spring force i
must maintain a net downward force on the fuel assembly during all normal c
and anticipated transient flow and temperature conditions.
Add w & E
- J.
Control Material Leaching A specific criteria for control material leaching has not been specified since leaching has not been observed as a problem.
In addition, adherance to previously specified stress and strain limits provide assurance that cladding integrity is maintained, therefo m exposure of the control material to the coolant is avoided.
i
l Insert E CESSAR Fuel assemblys will be verified to meet the design criteria for no net upward force on the fuel assembly during all normal and anticipated transient conditions.
s