ML20034B738
| ML20034B738 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 04/19/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034B737 | List: |
| References | |
| NUDOCS 9004300279 | |
| Download: ML20034B738 (7) | |
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?g UNITED STATES
'g NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 23 TO FACILITY OPERATING LICCNSE NO. NPF-72 AND AMENDMENT NO. 23 TO FACILITY OPERATING LICENSE NO. NPF-77 COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION UNITS 1 AND 2 DOCKET N05. 50-456 AND 50-457 TAC NOS. 75262 AND 75263
1.0 INTRODUCTION
Commonwealth Edison Company (the licensee) submitted a request in References 1 and 2 for Technical Specification (TS) changes to allow refueling and operation of the Braidwood Station Unit 1 Cycle 3 and Unit 2 Cycle 2 cores with the VANTAGE
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5 fuel design.
Currently, both units of Braidwood Station are operating with a Westinghouse 17x17 optimized fuel assembly (0FA) core.
Future core loadings corsist of a mixed core of 0FA and VANTAGE 5 to eventually an all VANTAGE 5 fueled core.
The VANTAGE 5 fuel design has been approved with conditions in the NRC safety evaluation (SE) on Westinghouse topical report WCAP-10444-P-A,
" Reference Core Report VANTAGE 5 Fuel Assembly."
The major design features of VANTAGE 5 fuel relative to the current OFA fuel design include:
integral fuel i
burnable absorbers (IFBA), intermediate flow mixer grids (IFM), reconstitutable top nozzles, extended burnup capability, axial blankets and debris filter bottom nozzle.
The licensee indicated in Reference 1 that the transition core andfullVANTAGE5coresafetyanalyseswereperformedagthermalpowerlevel of 3411 MWt. Other assumptions included a full power F g (hot channel enthalpy rise factor) of 1.65 for the VANTAGE 5 fuel and 1.55 for the OFA fuel, an increase in the maximum Fn (heat flux hot channel factor) to 2.50 and 10 percent steam generator tube plugging for the transient analysis and 15 percent for the LOCA analysis.
The TS changes include (1) use of the Westigghouse WRB-2 DNBR correlation for the VANTAGE 5 fuel, (2) an added maximum F of 1.65 for the VANTAGE 5 fuel, (3) an increased maximum F of 2.50 from 2.$2, and (4) an increased control rod n
drop time from 2.4 to 2.7 Veconds.
During the review of the VANTAGE 5 fuel design in WCAP-10444-P-A, the staff identifier conditions imposed on those licensees using the VANTAGE 5 fuel design.
Our review of the licensee's request for the TS changes, the j
associated supporting analyses and the responses to the staff's review i
questions (Refs. I and 2) will address those conditions listed in the safety evaluation on WCAP-10444-P-A.
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'2.0 EVALUATION 2.1 Statistical Convolution Method In the.SE on WCAP-10444-P-A, the staff required that the. statistical method ~
should not be used in VANTAGE 5 for evaluating the fuel rod shoulder gap. The licensee indicated (Ref. 1) that the statistical convolution, method was not-used for the VANTAGE 5 fuel design and the currently approved method was used.
s for evaluating the fuel rod shoulder gap. Therefore, we consider this acceptable.
2.2 Seismic and LOCA Loads-In the-SE on WCAP-10444-P-A, the staff required that.for each plant application,-
it must be demonstrated that the fuel assembly will maintain its coolable-
.I geometry under combined seismic:and LOCA loads. The licensee performed LOCA J
and seismic load evaluations for-transition cores and~an all VANTAGE 5 core.'
The results-indicate that the fuel assembly'in either case has enough margin tot sustain the combined seismic and LOCA loads.suchLthat the structural integrity.
and-coolable geometry are maintained. Based on tha-licensee's evaluation results, we conclude that the condition of seismic and LOCA-loads is satisfied.
2.3 Irradiation Demonstration Program -
In the SE on WCAP-10444-P-A, the staff reqdired that an irradiation program be performed to confirm the VANTAGE 5 fuel performance.= The licensee indicated that there were numerous demonstration programs involving VANTAGE 5 fuel assemblies. During 1984 through 1988,. four-VANTAGE 5 demonstration assemblies were loaded into the V.C. Summer Unit 1 Cycle 2 and. achieved an average burnup of about 46,000 MWD /MTU.
Individual VANTAGE 5 product features:have been demonstrated at other nuclear plants.
IFBA demonstration fuel rods have been irradiated in Turkey Point Units 3-and'4 for'two reactor cycles and the IFM grid feature has been irradiated at McGuire Unit 1 for threel reactor cycles, i
The. satisfactory performance of these demonstration assemblies: resulted in the VANTAGE 5-fuel reload in many Westinghouse reactors. Thus, we conclude that VANTAGE 5 fuel will perform satisfactorily in the Braidwood Station.
2.4 Improved Thermal Design Procedure ITDP)
I In the SE on WCAP-10444-P-A, the staff: required that those-restrictions in approving the use of the NRC approved Westinghouse improved thermal design, procedures, ITDP (Ref. 3),)should be applied to the VANTAGE 5-fuel design.
The licensee indicated-(Ref. 1 that they complied with the restrictions of ITDP for Braidwood. We therefore conclude that.this is acceptable.
2.5 Positive Moderator Temperature Coefficient i
In the SE on WCAP-10444-P-A, the staff required that if a positive moderator temaeraturecoefficient(MTC)isintended,thesamepositiveMTCconsistent.
wit 1 the p(lant TS should be used in the plant-specific analysis. The licensee i
indicated Ref.1)thattheMTCwillnotbepositive,andtheplantspecific
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MTC consistent with the TS was used for the analyzed cycles. Thus, we conclude that this restriction is satisfactorily met.
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In the SE on WCAP-10444-P-A, the staff required that plant-specific analysis be 4
performed to-Aow that the appropriate. safety criteria are not-violated with
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the higher valm of F and use of the' VANTAGE 5' fuel.- The licensee evaluated all the trantia r analyses for Braidwood Units 1 and-g upgraded to VANTAGE 5 -
fuel and plant Lperation with an increased maximum F
-(from 1,55 to 1.65), an increased maximum Fn.(from 2.32 to 2.50) and a.increa$ed contr61 rod drop; time-(from-2.4to2.7-seVonds).
The licensee also assumed the. steam generator' tube
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plugging to a level of 10 percent-in theirl evaluation! The licensee determined-11 the-events affected significantly by the; fuel design updates and operating; condition changes and reanalyzed those events.
In References 1 and 2 the '
licensee presented the reanalyzed results for the transients to support the reload application and Technical Specification changes.
o The reanalyzed events can be summarized into.three categories-i (1) DNBR transients affected by-increase of F.H The events are partial. loss, offlow,'completelossofflow,RCPshaft-$r.eak and RCP locked rotor with loss of offsite power.
(2) The transients affected by increase of F.. The transients are RCP locked rotor and rod ejection.
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.l (3) The transients affected by increase of the control rod: drop time..The events are RCP locked rotor and rod ejection.
1 The licensee determined that for this application, the minimum required DNBR values for the OFA fuel analysis are 1.32 for thimble cold wall. cells (three q
J fuel rods and a thimble tube) and 1.34 for a typical cell (four fuel rods).
The design DNBR values for the VANTAGE 5 fuel are 1.32 and l'.33 for thimble 1
and typical cells, respectively. However, in order to demonstrate that. the design DNBR values have enough margin to accommodate fuel: rod bow penalty and i
effect of the mixed cores, the licensee determined that the-minimum operating DNBR limits are 1.47 for thimble and l'.49 for typical cells for 0FA fuel, and 1.65 and 1.67 for thimble and typical cells respectively for VANTAGE 5.
Since the licensee used NRC approved methods to show that-all applicable 4
transient analysis acceptance criteria will not be violated for the proposed j
cycles, we find the results to be acceptable.
2.7 Reactor Coolant Pump Shaft Seizure j
In the SE on WCAP-10444-P-A, the staff required that the mechanistic approach in determining the fraction of the fuel failures during the reactor pum) seizure accident was unacceptable and the fuel failure criteria should ;>e 95/95 a
DNBR limit. The licensee reanalyzed the reactor coolant pump shaft seizure (locked rotor) accident based on a failure criterion of the peak clad.
temperature of 2700*F.
The licensee concluded that there is no fuel failure and the coolability was maintained since the calculated peak clad temperature-(1853*F) remained much less than 2700*F and the amount of Zirconium-water reaction was small. As indicated above, we disapproved of.the use of a mechanistic approach based on 2700*F peak clad temperature in determining the, fuel failure.
In response (Ref.1), the licensee indicated.that' this event was
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' analyzed by using the previously a) proved methods and showed that no rod was predicted to be below the 95/95 DN3R limit.
Since the acceptable fuel failure criterion of. 95/95 DNBR limit is used_ for DNBR-analysis, we conclude that the.
reactor coolant pump shaft seizure _ accident is-satisfactorily' addressed for VANTAGE 5 fuel.
2.8 LOCA Analysis in the SE on WCAP-10444-P-A, the staff required that plant-specific analyses should be performed to _show that the requirements of 10 CFR 50;46 are met., The' licensee analyzed large and-small break LOCAs to support'the reload licensing ?
application. -In the licensee's large: break LOCA analysis (Ref.:1)~ only double end cold leg guillotine.(DECLG) breaks:were_ analyzed since they wer,e' identified
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previously as limiting cases that result ~1n_the highest peak clad temperature._
gi The'DECLG break analysis:was performed with a total peaking factor of 2.5,102
- i percent of the. core power of-3411 MWt. RCS hot leg temperatures between-'600 and 619.3*F, and RCS cold leg temperatures batween 535.6 and 556.7,1and an assumed loss of offsite power at the beginning of,the accidente An assumption of 15 percent steam generator. tube plugging was made forEthe analysis. A sensitivity:
J, study of'DECLG break sizes on:the effect of the peak clad temperature was'-
performed by use of discharge coefficients of: 0.8 0.6,; and.0.4. The results showed that the DECLG break with-a discharge coefficient of 0.6 with the RCS operating-at a nominal _ hot leg! temperature of 619.3*F is the worst large break-
'1 case resulting in a 3eak clad: temperature of 1883.1'F.. Analysis: performed:
assuming the RCS to >e operating with a reduced hot leg; temperature of:600*F.
i was found to be_less limiting than the result obtained when the RCS was assumed 3'
to be operating with a hot leg temperature of 619.3*F. The licensee evaluated k
the effect _of transition core cycles on: the calculated PCT and_ determined that j
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the maximum increase in PCT is 50'F which yields.a transition-core PCT of?
l 1933.1'F.
i The analysis of a large break LOCA transient is divided-into three phases: _(1)-
j blowdown.(2) refill,and(3)reflood. The: licensee used SATAN-V1 code'(Ref. 4) for the transient thermal hydraulic calculation during blowdown period;-the WREFLOOD (Ref. 5) and BASH codes'(Ref 6) for the thermal: hydraulic calculation ofrefillandrefloodtransientperiods;theLOCBARTcode-(Ref.'7)for calculation of peak clad temperature and: the C0C0 code (Ref.'8) for the calculation of containment pressure transient.
As a result of our review, we find that the approved analytical models.and computer codes were used and'results showed that the peak clad tem)erature of i
1933.1*F, total metal-water reaction of less than 0.3 percent of; tie. fuel clad and local clad oxidation of-less than 3.26 percent are within the'10 CFR 50.46-acceptable criteria which are 2200'F, 1 percent and 17 percent, respectively.
i Therefore, we conclude that the'large break LOCA analysis is acceptable.
In the licensec's small break LOCA analysis, we find that the licensee used the approved NOTRUMP code (Refs. 9 and 10) for the calculation of transient depressurization of the reactor coolant system and core power and the LOCTA code (Ref. 7) for the calculation of the peak clad temperature. Only one core flow channel is modeled in NOTRUMP since the core flow during a small break is relatively slow,(providing enough time to maintain flow equilibrium between fuel assemblies i.e., no crossflow) in mixed cores. Hydraulic resistance
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- mismatch is not at facter for small-breaks.
Therefore, the licensee referenced Lthe small break LOCA fon the complete core of-the VANTAGE 5 fuel design as the bounding case for all trusition cycles. The analysis was done with assumptions of 102 percent of the core power of:3411 MWt= and a total peaking factor of 2.5.
Analyses for these break. sizes were performed to show that the worst break ~ size is a 3-incb diameter break which results in the highest geak clad temperature.of 1453.1'Fs well below the acceptable criterion or 2200 F.
Since the approved methods)ece used to show the' analytical results to be 4
within the acceptance criteria imposed in 10 CFR 50.46, we therefore conclude that the small areak LOCA' analysis-is acceptable.
3.0' TECHNICAL SPECIFICATION ChylGES The-proposed (TS) changes reflect the fuel design change and! assumptions used.
in the safety analysis to supportshe reload application. These are discussed-below.
~ (1) LNew DNBR Correlation and Operating DNBR Limits - pp B2-1, B3/4-2-1, B3/4-2-4 A new DNBR correlation (WRB-2) is referenced and the cycle specific operating DNBR limits with incluskn of the rod bow penalty. factor and-4 effect of the mixed core are added.?.o the TS..Since the. changes are 1
consistent with the assumptions used 'in_ the transient-analysis, they are.
acceptable.
(2)' Increased Control Rod Drop Time - p 3/4-1-19 i
The control rod drop time is revised to 2.7 seconds from 2.4 seconds due to the use of the VANTAGE 5 fuel design. The licensee has taken into account the effect of the increased control rod drop time in all related-y safety analysis. Thus, we conclude' that this change is acceptable.
(3) Increased peaking Factor - F (ppb 2-2;3/4-2-8,B3/4-2-5)'
Fq (pp 3/4-2-4, 3/4-2-5, B3/4-2-1)
The maximum F-and F are. increased from 1.55 to 1.65, and 2.32 to 2.50,
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Sincekhechanges'areconsistentwiththeassumptionsused respectively.
l in the' analyses to support the reload application, they are acceptable.
j (4) VANTAGE 5 Design - pp 2-B, 3/4-2 The VANTAGE 5 fuel design is added to the TS. Since VANTAGE 5 is acceptable for use in the Braidwood cores, we conclude that the changes are acceptable.
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(5) SurveillanceRequirementChanges-(pp 3/4-1-4,3/4-1-5,B3/4-1-2)
Surveillance 4.1.1.3.a is modified to compare BOL MTC with the predicted MTCs at various burnup conditions and to' develop rod withdrawal limits in order to-keep MTC negative. The-changet are supported by the analytical assumption that no positive MTC is used through the cycle, and are '
therefore acceptable.
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. 4.0
SUMMARY
We have reviewed the licensee submittal of Technical Specification changes and related analytical results to support the request to allow the operation'of Cycles 3 and 2 of Braidwood Units 1 and 2 cores, respectively.- Based on' the, approved generic topical report,.WCAP-10444-P-A, and plant-specific analyses.
(Refs.1 and 2), we approve the use of VANTAGE 5 fuel design and Technical Specification changes, and changes to the associated Bases, for Braidwood Station Unit 1 Cycle 3 and Unit 2 Cycle 2' reload cores, i
5.0 ENVIRONMENTAL CONSIDERATION
1 The amendments involve changes to a requirement with respect to the insta11ation' or use of a facility component located within the restricted area as defined in a
The staff has determined.that the amendments involve no j
1 significant increase in'the amounts, and no significant change in the types,,
of any effluents that may be release offsite, and that there is no significant -
increase in individual or cumulative occupational radiation exposures. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration-and there'has been no public comment on such finding. Accordingly the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be-prepared in connection with the issuance of these amendments.
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6.0 CONCLUSION
1 The staff has further concluded, based on the considerations discussed above,-.
that:
(1) there is reasonable assurance that the health and safet public will-not be endangered by operation =in the proposed manner,y of the-and(2) such activities will be conducted in compliance with the Commission's
- regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
1.
Letter from S. Hunsader (Commonwealth Edison) to T. Murley (NRC), dated October 19, 1989.
2.
Letter from S. Hunsader (Connonwealth Edison) to T. Murley (NRC), dated i
February 16, 1990.
3.
WCAP-8567-P-A, " Improved Thermal Design Procedure," February 1989.
4.
WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary)f Loss of' C
" SATAN-VI Programs:
Comprehensive Spacetime Dependent Analysis o June 1974, 5.
WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), " Calculated Model i
for Code Reflood After a Loss of Coolant Accident (WREFLOOD)," June-1974 i
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WCAP-10266-P-A. Revision 2 with Addenda (Proprietary), "Th'e 19f.) Version of the Westinghouse ECCS Evaluation Model.Using the. BASH Code,* August.
-1986.-
7.
_ WCAP-8301 (Proprietary) an'd WCAP-8305' (Non-Proprietary), "LOCTA-IV Program:l Loss of Coolant Transient Analysis,"~ June 1974.
WCAP-8327 (Proprietary)(C0CO)," June 1974.'andWCAP-8326(Non-Propr 8.
4 Pressure Analysis Code 9.
WCAP-10079-P-A (Proprietary) and'WCAP-10080-P-A (Non-Proprietary),-
"NOTRUMP, A. Nodal Transient Small Break and General Network-Code " August 1985.
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10.. WCAP-10054-P-A (Proprietary) Land WCAP-10081-P-A (Non-Proprietary),
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~ " Westinghouse Small Break ECCS Evaluation Model' Using. the-NOTRUMP Code "
August 1985.'
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Principal Contributor:
S. Sun, S.' Sands Dated: April 19, 1990' 1
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