ML20011E561
| ML20011E561 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 01/31/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20011E562 | List: |
| References | |
| NUDOCS 9002160092 | |
| Download: ML20011E561 (7) | |
Text
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UNITED STATES n
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NUCLEAR REGULATORY COMMISSION
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$AFETY EVALUATION.BY.TNE OFFICE-OF. NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 36 TO FACILITY OPERATING. LICENSE NO. NPF-37 AND AMENDMENT NO. 36"TO FACILITY OPERATING LICENSE NO. NPF-66 COMMONWEALTH EDISON COMPANY BYRON STATION.. UNITS 1 AND 2 i
DOCKET NOS. 50-454 AND 50-455 TAC NOS. 74166 AND 74167
1.0 INTRODUCTION
By letter dated July 31, Technical Specification (TS) changes to allow refueling i 1989(Reference 1),CommonwealthEdison(thelicensee) submitted a request for and operation of the Byron Station Unit 1 Cycle 4 and Unit 2 Cycle 3 cores with the VANTAGE 5 fuel design. Currently, both units of Byron Station are operating i
with a Westinghouse 17x17 optimized fuel assembly (OFA) core.
Future core loadings consist of a mixed core of 505-70% OFA and 30-501 VANTAGE 5 to even-l tually an all VANTAGE 5 fueled core. The VANTAGE 5 fuel design has been approved with conditions in the NRC Safety Evaluation on Westinghouse To Report WCAP-10444-P A, " Reference Core Report VANTAGE 5 Fuel Assembly." pical The major design features of VANTAGE 5 fuel relative to the current OFA fuel design 1
include:
integral fuel 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 2 that the transition core and full YANTAGE 5 core safety analyses were per-
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formedatatgermalpowerlevelof3411MWt. Other assumptions included a of 1.65 for the VANTAGE 5 fuel and 1.55 for the OFA fuel, an
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full power F increaseintNemaximumFnto2.50and10%steamgeneratortubepluggingfor j
the transient analysis anB 151 for the LOCA analysis.
l The TS changes include: (1) use of Westfnghouse WRB-2 DNBR correlation for the VANTAGE 5 fuel,(2)anaddedmaximumF (3)an time from 2.4 to 2.9 secon.50 from 2.32,H of 1.65 for the VANT increased maximum F of 2 and(4)anincreasedcontrolroddrop ds.
During the review of the VANTAGE 5 fuel design in WCAP-10444-P-A, we identified i
conditions imposed on those licensees using the VANTAGE 5 fuel design. Our l
review of the licensee's request for the TS changes, the associated supporting analyses and the responses to the staff's review questions (Refs. 1 and 2) will 1
address those conditions listed in the Safety Evaluation (SE) on WCAP-10444-P-A.
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2-2.0 EVALUATION
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2.1 Statistical Convolution Method In the SE on WCAP-10444-P-A, we stated that the statistical method should not be used in VANTAGE 5 for evaluating the fuel rod shoulder gap.
The licensee indicated (Ref. 2) that the statistical convolution method was not used for the VANTAGE 5 fuel design and the currently approved method was used for evaluating the fuel rod shoulder gap. Therefore, we consider this acceptable.
2.2 Seismic end LOCA Loads In the SE on WCAP-10444-P-A, we stated that for each plant application,-it must be demonstrated that the fuel assembly will maintain its coolable geometry under combined seismic and LOCA loads. The licensee performed LOCA and seismic load evaluations for transition cores and an all VANTAGE 5 core.
The results indicate t % t the fuel assembly in either case has enough margin to sustain the combined seismic and LOCA loads such that the structural integrity and coolable geometry are maintained. Based on the 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, we required 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 I for three reactor cycles.
The satisfactory performance of these demonstration assemblies resulted in the VANTAGE 5 fuel reinad in many Westinghouse reactors. Thus, we conclude that VANTAGE 5 fuel wn serform satisfactorily in the Byron Station.
2.4 Improved Thermat Design Procedure (ITDP)
In the SE on WCAP-10444-P-A, we stated that those restrictions in approving the use of the NRC approved Westinghouse improved thermal design procedures, ITDP(Ref.3)f.shouldbeappliedtotheVANTAGE5fueldesign.
The licensee indicated (Re
- 2) that they complied with the restrictions of ITDP for Byron.
We therefore conclude that this is acceptable.
2.5 Positive Moderator Temperature Coefficient In the SE on WCAP-10444-P-A, we stated that if a positive moderator temperature coefficient (MTC) is intended, the same MTC should be used in the plant-i specific analysis. The licensee indicated (Ref. 2) that the MTC will not be positive for the anblyted cycles. Thus, we conclude that this restriction is satisfactcrily met.
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h 2.6 Transient Analysis In the SE on WCAP-10444-P-A, we required that plant-specific analysis be performed to show that the appropriate safety criteria are not violated with g
the higher value of F and use of the VANTAGE 5 fuel.
The licensee evaluated all the transient analyses for Byron Units 1 and 2 upgraded to VANTAGE 5 fuel N
n(from2.32to2.50)andanYn(from1.55to1.65),an and plant operation with an increased maximum F creased control rod drop time increased maximum F (from 2.4 to 2.7 seconds). The licensee also assumed the steam generator tube plugging to a level of 10% in their evaluation. The licensee determined the events affected significantly by the fuel design updates and operating condition changes and reanalyzed those events.
In Reference 1, the licensee presented the reanalyzed results for the transients to support the reload application and technical specification changes.
The reanalyzed events can be suninarized into three categories:
N (1) DNBR transients affected by increase of F The events are partial loss offlow,completelossofflow,RCPshaftUr.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.
0 (3) The transients affected by increase of the control rod drop time. The events are RCP Tocked rotor and rod ejection.
The licensee determined that for this application, the minimum required DNBR values for the OFA fuel analysis ce 1.32 for thimble cold wall cells (three 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 1.33 for thimble and typical cells, respectively. However, in order to demonstrate that the design DNBR values have enough margin to accommodate fuel rod bow penalty and effect of the mixed cores, the licensee determined that the minimum operating DilBR limits are 1,47 for thimble and 1.49 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 the NRC approved methods to show that all applicable transient analysis acceptance criteria will not be violated for the proposed cycles, we approve the transient analyses.
2.7 Reactor Coolant Pump Shaft Seizure In the SE on WCAP-10444-P-A we stated that the mechanistic approach in deterneining the fraction of,the fuel failures during the reactor pump seizure accident was unacceptable and the fuel failure criteria should be 95/95 DNBR limit. The licensee reanalyzed the reactor coolant pump shaft seizure (locked rotor) accident based on a failure criterion of the pea k clad temperature of 2700'F. The licensee concluded that there is no fuel failure and the ecclability 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 disapprove of the use of a mechanistic approach
,- 3.., > - e based on 2700'F peak clad temperature in determining the fuel failure.
In response (Ref. 2), the licensee indicated that this event was analyzed by using the 3reviously approved methods and showed that no rod was predicted to be below tie 95/95 DNBR limit.
Since the acceptable fuel failure criterion of 95/95 DNBR limit is used for DNBR analysis, we conclude that the reactor coolant pump shaf t seizure acci'ent is satisfactorily addressed for VANTAGE 5 fuel.
2.8 LOCA Analysis In the SE on WCAP-10444-P-A, we stated that the plant specific analysis should be performed to show that the requirements of 10 CFR 50.46 are met. The licensee analyzed large and small break
.s to support the reload licensing application.
In the licensee's large bre n LOCA analysis (Ref. 1), only double end cold leg guillotine (DECLG) breaks were analyzed since they were identified previously as limiting cases that result in the highest peak clad temperature.
The DECLG break analysis was performed with a total peaking factor of 2.5,102%
of the core power of 3411 MWt, temperatures between 600 to 619.3'F in the RCS hot legs and 535.6 to 556.7 in the RCS cold legs respectively, and an assumed loss of offsite power at the beginning of the accident.
An assumption of 15%
steam generator tube plugging was made for the analysis.
A sensitivity 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 case r
resulting in a peak clad temperature of 1883.1*F. Analysis performed assuming the RCS to be operating with a reduced hot leg temperature of 600'F was found to be less limiting than the result obtained when the RCS was assumed to be
. operating with a hot leg temperature of 619.3'F. The licensee evaluated the effect of transition core cycles on the calculated PCT and determined that the maximum increase in PCT is 50'F which yields a transition core PCT of 1933.1*F.
The analysis of a large break LOCA transient is divided into three phases:
(1) blowdown, (2) refill, and (3) reflood. The licensee used SATAN-V1 code (Ref. 4) for the transient thermal hydraulic calculation during blowdown period;theWREFLOOD(Ref.5)andBASHcodes(Ref.6)forthethermal i
hydraulic calculation of refill and reflood transient periods; the LOCBART code (Ref. 7) for calculation of peak clad temperature and the C0CO 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 temperature of 1933.1'F, to'.a1 metal-water reaction of less than 0.3% of the fuel clad and local clad oxidation of less than 3.26% are within the 10 CFR 50.46 acceptable criteria which are 2200*F, 1% and 17%, respectively. Therefore, we conclude that the large break LOCA analysis is acceptable.
I In the licensee'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 (Refs. 7) for the calculation of the peak clad temperature. Only one l
core flow channel is modeled in NOTRUMP since the core flow during a small
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break is relatively slow (i.e., no crossflow) in mixed cores.providing enough time to main between fuel assemblies Hydraulic resistance mismatch is not a factor for small break. Therefore, the licensee referenced the small break LOCA for the complete core of the VANTAGE 5 fuel design as the bounding case for all transition cycles. The analysis was done with assumptions of 102% 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-inch diameter break which results in the highest peak clad temperature of 1453.1'F, well below the acceptable criterion of 2200'F.
Since the approved methods were used to show the analytical results to be within the acceptance criteria imposed in 10 CFR 50.46, we therefore conclude that the small break LOCA analysis is acceptable.
3.0 TECHNICAL SPECIFICATION CHANGES The proposed technical specification (TS) changes reflect impact of the fuel design change and assumptions used in the safety analysis to support the reload application. We discuss the TS changes as follows.
(1) New DNBR Correlation and Operating DNBR Limits - pp B 2-1, B 3/4 2-1, B 3/4 2-4.
A new DNBR correlation of WRB-2 is added and the cycle specific operating DNBR limits with inclusion of the rod bow penalty factor and effect of the mixed core are added to the TS. Since the changes are consistent with the assumptions used in the trensient analysis, we approve the changes.
(2) Increased Control Rod Drop Time - p 3/4 1-19.
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 safety analysis. Thus, we conclude that this change is acceptable.
U (pp B 2-2, 3/4 2-8 (3) Increased Peaking Factor - F q(pp$/42-4,B3/42-1),B3/42-5)
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The maximum F and F are increased from 1.55 to 1.65, and 2.32 to 2.50, l.
respectively. Sincekhechangesareconsistentwiththeassumptions used in the analyses support the reload application, we approve the changes.
(4) VANTAGE 5 Design - pp 2-8, 3/4 2-7.
l The VANTAGE 5 fuel design is added to the TS.
Since VANTAGE 5 is I
acceptable for use in the Byron cores, we conclude that the changes are acceptable.
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(5) Surveillance Requirement Changes - (pp 3/41-4,3/41-5,B3/41-2).
BOL is deleted from MTC LCO. Surveillance 4.1.1.3.a is modified to compare BOL MTC with burnup and rod withdrawal limits are developed to keep MTC negative. Since the changes are supported by the analytical
. assumption that no positive MTC is used through the cycle, we approve the
' changes.
We have reviewed the licensee submittal of technical specification changes and related analytical results to support the request to allow the operation of Cycles 4 and 3 of Byron Units 1 and 2 cores, respectively. Based on the approved generic topical report WCAP-10444-P-A, and plant specific analyses (Ref. 1), we approve the use of VANTAGE 5 fuel design and technical specifica-tion changes for the Byron Station Unit 1 Cycle 3 and Unit 2 Cycle 4 reload Cores.
4.0 FINDING OF NO SIGNIFICANT IMPACT Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact has been prepared and published (55 FR 3123) in the Federal Register on January 30, 1990. Accordingly, based upon the environmental assessment, the Comission has determined that the issuance of this amendment will not have a significant effect on the quality of the.
human environment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by-operation in the proposed manner, and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the comon defense and security or to the health and safety of the public.
6.0 REFERENCES
1.
Letter from R. Chrzanowski (Comonwealth Edison) to T. Morley (NRC), dated July 3?, 1989.
2.
Letter from R. Chrzanowski (Comonwealth Edison) to T. Murley (NRC), dated October 19, 1989.
3.
WCAP-8567-P-A, Improved Thermal Design Procedure. February 1989.
4.
WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), SATAN-VI Program:
Comprehensive Spacetime Dependent Analysis of Loss of Coolant, June 1974.
4 5.
WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), Calculated Model for Code Reflood After a Loss of Coolant Accident (WREFLOOD), June 1974.
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-7 6.
WCAP-10266-P-A, Revision 2 with Addenda (Proprietary), the 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code, August 1986.
7.
WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), LOCTA-IV Program:
' Loss of Coolant Transient Analysis, June 1974.-
4 8.
WCAP-8327 (Proprietary)(and WCAP-8326 (Non-Proprietary), Containment Pressure Analysis Code C0CO), June 1974.
9.
WCAP-100/9-P-A (Proprietary) and WCAP-10080-P-A (Non-Proprietary),
I NOTRUMP, A Nodal Transient Snell Break and General Network Code, August 1985.
- 10. WCAP-10054-P-A (Proprietary) and WCAP-10081-P-A (Non-Proprietary).
Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985.
Principal Contributor:
S. Sun Dated: January 31, 1990
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