ML20247M933
| ML20247M933 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 05/24/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247M917 | List: |
| References | |
| NUDOCS 8906050196 | |
| Download: ML20247M933 (7) | |
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UNITED STATES 3 /, g// 3gf[g g
NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NO.153 TO FACILITY OPERATING LICENSE NO. NPF-1 PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY TROJAN NUCLEAR PLANT DOCKET NO. 50-344
1.0 INTRODUCTION
By a [[letter::05000344/LER-1987-030, :on 871021,discovered That Component Cooling Water (CCW) Flow to RHR a HX Was Only 4,460 Gpm.Caused by Procedure Inadequacy.Rhr a HX CCW Outlet Valve Adjusted to Provide 5,100 Gpm|letter dated November 20, 1987]], as supplemented May 27 and August 12, 1988, Portland General Electric Company (PGE) submitted a request for Technical Specification changes that would be needed fo.r I
the use of upgraded fuel assemblies for Trojan Cycle 12 reload core. The features of the upgraded fuel assembly consist of reconstitutable topnozzles,axialfuelblankets,andthecapabilityofachievfnghigh burnups as well as. satisfying higher nuclear peaking factor F and F.
These features were previously reviewed and approved by NRC in Ne q
Westinghouse topical report WCAP-10444-P-A, " Reference Core Report VANTAGE 5 Fuel Assembly."
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The Westinghouse high burnup topical report WCAP-10125-P-A, " Extended Burnup Evaluation of Westinghouse Fuel" was also approved by the NRC staff.
The upgraded fuel assembly based on the approved WCAP-10125-P-A is designed to achieve a burnup of 60,000 mwd /MTV peak rod average.
During the review of VANTAGE 5 fuel design in WCAP-10444-P-A, the staff l
identified several conditions that needed to be resolved by those licensees using VANTAGE 5 fuel design.
Since the upgraded fuel design adopts some features from the VANTAGE 5 fuel design, our review and evaluation will address those conditions listed in the safety evaluation of WCAP-10444-P-A that affect Trojan's upgraded fuel.
2.0 DISCUSSION AND EVALUATION 2.1 Statistical Convolution Method In our Safety Evaluation (SE) of WCAP-10444, we stated that the statistical convolution method should not be used in VANTAGE 5 for evaluating the fuel rod shoulder' gap.
PGE indicated that the statistical convolution method was not used for the upgraded fuel design; instead, the currently approved method was used for evaluating fuel rod shoulder gap, and the method used is therefore acceptable.
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, l 2.2 Irradiation Demonstration Program 1
i In our SE of WCAP-10444, we required that an irradiation program be performed to confirm the VANTAGE 5 fuel performance.
PGE stated that V.C. Summer Unit I has four VANTAGE 5 demonstration assemblies irradiated up to 30,000 mwd /MTU assembly average.
Post-irradiation examinations showed all four assemblies were of good mechanical integrity. These four assemblies have been reinserted for continuing irradiation.
The V. C.
Summer's four VANTAGE 5 fuel assemblies incorporate the upgraded fuel design features; we thus conclude that the upgraced fuel assemblies will perform satisfactorily in Trojan.
2.3 Improved Thermal Design Procedure (ITDP)
In our SE of WCAP-10444, we stated that,those restrictions in approving the use of Westinghouse ITDP should be applied to the VANTAGE 5' fuel design.
PGE indicated that they conformed to these restrictions of ITDP for Trojan.
We therefore conclude that this is acceptable.
2.4 DNBR Limit In our SE of WCAP-10444, we stated that plant-specific analysis should be performedtoshowthatthedeparturefrgmnuclearboiling. ratio (DNBR) limit is not violated with the higher value F PGE examined all the transient analysesrelatedtoDNBRcalculationandNo.und that the DNBR limit was not exceeded for any transient. We therefore consider this condition to be acceptable.
2.5 Positive Moderator Temperature Coefficient (MTC)
In our SE of WCAP-10444, we stated that if a positive MTC is intended, the same positive MTC should be used in the plant specific analysis.
PGE indicated that a positive MTC was considered in the analyses, and did not show'any change from previous Final Safety Analysis Report (FSAR) results.
We thus consider this acceptable.
2.6 Reactor Coolant Pump Shaft Seizure In our SE of WCAP-10444, we stated that the mechanistic approach in determining the fraction of fuel failures during the reactor coolant pump seizure accident was unacceptable, and that the fuel failure criterion should be the 95/95 DNBR limit.
The Trojan FSAR Section 15.3.3 concluded that the coolability was maintained based on a peak clad temperature of 2700 F and that typically 3% of the fuel rods in the core would be below DNBR during the reactor coolant pump shaft seizure accident.
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j In letter dated May 27, 1988, PGE indicated that the design changes from current fuel to upgraded fuel were small and resulted in no signi-ficant alteration to the nuclear and thermohydraulic performance.
- Thus, PGE contended that the FSAR conclusion was still valid and a reanalysis was not necessary for the coolant pump shaft seizure accident.
- However, j
since no plant-specific analysis had been performed for Trojan, the staff requested that the Trojan core be analyzed to demonstrate the validity of
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the assumption that only 3 of the fuel rods below the DNBR limit would have N
a higher value of F during the pump shaft seizure accident.
i AH By. letters dated August 12 and December 15, 1988, PGE committed to perform and submit a plant-specific analysis of the reactor coolant pump shaft seizure accident for the Trojan core containing the upgraded fuel, based on approved methods and fuel failure criteria.
The plant-specific extended analysis was submitted by letter dated February 15,.1989.
This extended analysis supplements the locked reactor analysis given in Section 15 of the Trojan FSAR, but does not replace it.
The extended analysis maximinizes the number of fuel pins in the DNB condition, and conservatively assumes that all of them fail and release their fission products to the reactor coolant.
The resulting doses at the site boundary, at the low population zone boundary, and in the control room, are well within the regulatory limits.
It should be noted that the extended analysis results are not relied upon for the staff conclusion regarding the acceptability of the proposed upgrade fuel assembly use.
Since the upgraded fuel assemblies currently reside in the non-limiting core position, the analysis supporting the Trojan operating Cycle 12 is a bounding one, and the acceptability of the proposed use of the upgraded fuel does not depend on the results of the February 1989 extended analysis.
The extended analysis shows only that the radiological consequences of a locked rotor coolant pump, even using the most conservative 4
assumptions regarding failed fuel from the event, do not produce unacceptable
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doses either to the general public or to plant workers.
- 2. 7 LOCA Analysis In our SE of WCAP-10444, we stated that plant specific LOCA analyses should be performed to show that the requirements of 10 CFR 50.46 and Appendix K to 10 CFR Part 50 are met.
PGE analyzed the cases of large break LOCA and small break LOCA using the approved Westinghouse LOCA Evaluation Model.
The results showed that all the requirements in 10 CFR 50.46 and Appendix K to 10 CFR Part 50 were met for Trojan.
We therefore conclude that thisconditionissatisfiedfortheupgradedfuelinTrojan.
2.8 Design Basis Accident Analysis Relative To Extended fuel Burnup PGE's amendment application implicitly requested authorization to allow feel burnup up to 60,000 megawatt days per metric ton (MWD /MT).
PGE evaluated the potential impact of this change un the radiological assessment of design basis accidents (DBA) which were previously analyzed in the licensing of the Trojan Nuclear Plant.
In their submittal of
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November 24, Wi7, PGE concluded that the design basis accidents previously analyzed in tryir FSAR bound any potential radiological consequences of DBA that cvh result with the extended burnup fuel.
During tw review of a similar request for extended fuel burnup by the licersee sur the Shearon Harris Nuclear Power Plant, the staff reviewed.a pub 7irdion which was prepared for the NRC entitled, " Assessment of the Use cf Extended Burnup Fuel in Light Water Reactors," NUREG/CR 5009, Fenrgny 1988.
The NRC contractor, the Pacific Northwest Laboratory (P E)'of Battelle Memorial Institute, examined the changes that could rM d?t in the NRC DBA assumptions described in the various appropriate handard Review Plan sections and/or Regulatory Guides, that could result from the use of extended burnup fuel (up to 60,000 MWD /MT).. The staff Egreed that the only DBA that could be affected by the use of. extended burnup fuel, even in a minor way, was the potential thyroid doses that could result from a fuel handling accident.
PNL estimated that I-131 fuel gap activity in the peak fuel rod with 60,000 MWD /MT burnup could be as high as 12%.
This value is approximately 20% higher than the value normally used by the staff in evaluating fuel handling accidents (Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the' Fuel Handling and Storage Facilities for Boiling and Pressurized Water Reactors").
The staff, therefore, reevaluated the postulated fuel hand' ling accident for Trojan with an increase in iodine gap activity in the fuel damaged in a fuel handling accident.
Table 1 presents the estimated fuel handling accident thyroid doses presented in the operating licensing Safety Evaluation Report dated October 1974, and the increased thyroid doses (by 20%) resulting from extended burnup fuel.
The staff concludes that the only potential increased doses resulting from DBA with extended fuel burnup to,60,000 MWD /MT is the thyroid dose resulting from fuel handling accidents, that these doses remain well within the 300 Rem thyroid exposure guideline values set forth in 10 CFR 100 and that this small calculated increase is not significant,'and therefore accept-able.
Table 1 Thyroid Doses as a Consequence of DBA Fuel Handling Accident'.
Exclusion Area low Population Zone Thyroid Dose (Rem)
Thyroid Dose (Rem)
A*
B**
A*
B**
30.0 36.0 4.0 e.8
- A
- SER Dose l
- B
- Extended Fuel Burnup Dose l
r-F 3.0 TECHNICAL SPECIFICATION CHANGES The preceeding discussions have shown the acceptability of the use of upgraded fuel design for Trojan.
We reviewed each of the Technical Specifications changes which are the results of higher nuclear peaking factors, F and F and that they appropriately reflect the analyses, and therefoN are a0ceptable. (See Attached Table 2, List of Changes)
We have reviewed the licensee submittal of upgraded fuel design and Technical Specification changes for Trojan Cycle 12 reload core.
Based on the approved generic topical reports, WCAP-10444 and WCAP-10125, and plant-specific analyses (including a commitment for reanalysis of the reactor coolant pump shaft seizure accident based on approved methods and fuel failure criteria), we approve the upgradeo fuel design and Technical Specification changes for Cycle 12.
4.0 CONTACT WITH STATE OFFICIAL The NRC staff has notified the Oregon Department of Energy of the proposed issuance of this amendment along with the proposed determination of no significant' hazards consideration.
No comments were received.
5.0 ENVIRONMEN1AL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published (53 FR 22386)) in the Federal Register on May 23, 1989.
Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.
- 6. 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
PRINCIPAL CONTRIBUTORS:
S. L. Wu T Chan R. Bevan Dated: May 24, 1989 I
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. l., _.e Table 2 List of Changes 1
DESCRIPTION OF CHANGE A description of the proposed changes to the existing Trojan Technical pi
-Specifications-(TTS) is as follows.
Pane Section Description of Ch=ade 2-2 2.1 Safety Figure 2.1-1 is revised.to' change the four-loop-Limits safety limit on the combination of thermal power, pressurizer pressure and the highest operating loop-Tayg. This is due to reanalysis with a revised-expression for the Enthalpy Rise Hot Channel Factor F{H. See the change description for TTS page B 2-2, below.
2-5, 2.2 Limiting Notes 3 and 4 are changed to revise the margin for 2-9
. Safety the maxianas trip points'(Allowable Values) for System Overtemperature and Overpower AT.
Settings In the Overtemperature AT trip setpoint function:
2-7,
- T' and T" are revised for consistency with 2-8 current analysis, 2-7
- The definition of AT in Note 1 is corrected o
by deleting the parenthetical phrase. Also, the K1 setpoint is revised from 1.32 to 1.28 for consistency with the revised core limits.
2-8
- The ft (AI) function is modified'for consistency with the revised axial offset limits, and e
2-8
- A typographical error is corrected in Note 2.
B 2-2 Bases 2.1.1 Theenthalpyrisehotchannelfactor,F{H,is Reactor Core revised from 1.49 to 1.56.
B 2-6 Bases 2.2.1 The minimum limit of 1.73 for the Departure from Reactor Trip Nucleate Boiling Ratio (DNBR) is replaced by the System Instru-phrase "the safety analysis DNBR limit", such that mentation future analysis changes will not require a change Setpoints to the Technical Specification Bases.
L 3/4 2-5, 3.2.2 The constant values in the expressions for the 3/4 2-7 Heat Flux Hot limits on Fq(2) are revised to 2.50 and 5.00 for Channel Factor greater than, and less than or equal to 50 percent Fg(Z) power, respectively. The K(Z) function in Figure 3.2-2 is also revised, as established by a Loss-of-Coolant Accident (LOCA) reanalysis.
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Table 2 (continued)
Paae Section Description of Chanae 3/4 2-6a An error is corrected in Paragraph 4.2.2.2.e.
3/4 2-8 3.2.3, 4.2.3.2 Therevisedvalueof1.56isinsertedforF[H-3/4 2-9 RCS Flow Rate and Fg 3/4 2-9a 4.2.3.2 Therevisedvalueof1.56isinsertedforF{H RCS Flow Rate in the expression for the limit on F
- R and FR B 3/4 2-1 3/4.2.1 The value for Fg(2) is revised from 2.32 to Axial Flux to 2.50.
Difference (AFD)
B 3/4 2-4, Bases 3/4.2.2 The revised value of 1.56 is inserted in the B 3/4 2-5 and 3/4.2.3 expression for maximum F H, and revised values Heat Flux Hot for DKBR margins are incorporated based on incorpo-Channel Factor, ration of a new rod bow penalty. Also, the 2.5 per-etc.
cent DNBR penalty due to core bypass flow is accounted for in the new safety limit DNBR value.
B 3/4 2-6 Bases 3/4.2.5 The minimum limit of 1.73 for DKBR is replaced by DNB parameters the phrase " greater than or equal to the safety analysis DNBR limit" puch that future analysis changes will not require a change to the Technical Specification Bases.
B 3/4 4-1 Bases 3/4.4.1 The minimum limit of 1.73 for DNBR is replaced by Reactor Cool-the phrase "the safety analysis DKBR limit" such ant Loops and that future analysis changes will not require a coolant change to the Technical Specification Bases.
Circulation B 3/4 5-2 3/4.5.5 A third basis is added to indicate that RWST minimum Refueling volume and boron concentration sust meet suberiti-Water Storage cality requirements for cold post-large break LOCA Tank (RWST) conditions.
5-4 5.3.1 This section is revised to provide for loading of Fuel substitute rods for fuel rods as required for fuel Assemblies reconstitution.
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