ML19289F463
| ML19289F463 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/02/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19289F455 | List: |
| References | |
| NUDOCS 7906070400 | |
| Download: ML19289F463 (10) | |
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4 UNITED STATES p
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't NUCLEAR REGULATORY COMMISSION WASHINGTON, D. S. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 48 TO FACILITY OPERATING LICENSE NO. DPR-62 CAROLINA POWER AND LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET N0. 50-324 1.0 Introduction By letter dated February 2,1979, as supplemem*ed March 16, 21 and 27, April 13, April 27, and May 1,1979, Carolina ~ 7wer and Light Company (the licensee) requested amendments to Facility Operating License No.
The amendments would modify the Technical Specifications for the Brunswick Stsam Electric Plant, Unit No. 2 (the facility) to establish revised safety and operating limits for operat on 'n Cycle 3 i
with 7x7, 8x8, and 8x8R fuel.
The February 2, 1979 submitt'.1 requested credit for the end-of-cycle recirculation pump trip (E0C RPT) feature which was installed in BSEP Unit No. 2 during the refueling outage.
The April 13, 1979 submittal provided a non-RPT analysis for Unit No. 2 Cycle 3.
This analysis was performed to generate fallback operating MCPR limits if the E0C RPT becomes inoperable during Cycle 3.
As a result of the licensee's proposal and our review we have some reservations about the design of the E0C RPT system.
Therefore, as agreed to with your staff, we have not included cradit for the E0C RPT system in the operating limit minimum critical power ratios. As a result, modifications to the licensee's proposed Technical Specifications were necessary.
These modification.; were discussed with and agreed to by the licensee.
2.0 Discussion The Larolina Power and Light Company has proposed changes to the Technical Specifications of the Brunswick Unit No. 2 Nuclear Power Plant (BSEP 2).
The proposed changes relate to the replacement of 132 foel assemblies constituting refueling of the core for third cycle operation at power levels up to 2436 Mwt (100% power).
79 0 6 0 7 0 doo 2232 095 e
. In support of the reload application, the licensee h.s provided the GE BWR Reload 2 Licensing suomittal for BSEP 2 (References 1, 2),
information on the BSEP 2 Loss-of-Coolan. Accident (LOCA) analysis (References 1 and 3), responses to NRC requests for additional infonnation (Reference 12), and BSEP 2 Physics Startup Tests (Reference 5).
This reload involves loading of General Electric Comoany Retrofit (8x8R) fuel.
The description of the nuclear and mechanical design of the (8x8R) fuel and the (8x8) fuel is contained in GE's licensing topical report for BWR reloads (Reference 6).
Reference 6 also contains a complete set of references to topical reports which describe
- GE's analytical methods for nuclear, thermal-hydraulic, transient and accident calculations, and information regarding the applicability of these methods to cores containing (7x7), (8x8) and (8x8R) fuel.
Values for each plant-specific data such as steady state operating pressure, core flow, safety and safety / relief valve setpoints, rated thermal power, rated steam flow, and other various design parameters are provided in Reference 6.
f.dditional plant and cycle dependent information are provided in the reload application (Reference 1) which closely follows the outline of Appendix A of Reference 6.
Reference 8 describes the staff's review, approval, and conditions of approval for the plant-specific data addressed in Refe'rence 6.
The above mentioned plant-specific data have been used in the transient and accident analysis provided with the reload application.
Our Safety Evaluation (Reference 8) of the GE generic reload licensing topical report concluded that the nuclear and mechanical design of the (8x8R) fuel, and GEs analytical methods' for nuclear, thermal-hydr aulic, and transient and accident calculations as applied to mixed cores containing (7x7), (8x8) and (8x8R) fuel are acceptable. Approval of the nuclear and mechanical design of (8x8) fuel was determined based on infonnation in Reference 7 and ex-pressed i' :.ne staff's status report (Reference 9) on that document.
Because of our review of a large number of generic considerations related to use of (8x8R) fuel in mixed loadings with (8x8) and (7x7) fuel, and on the basis of the evaluations which have been presented in Reference 8, only a limited number of additional areas of review have been included in this safety evaluation report.
For evaluations of areas not specifically addressed in this safety evaluation report, the reader is referred to Reference 8.
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, During the current outage CP&L has modified Unit No. 2 to provide automatic trip of both recirculation pumps after turbina trip or genera?.or load rejection.
The purpose of this trip is to reduce the reactor pressure and peak heat flux resulting from these transients coincidant with a failure of the bypass system.
Our safety evaluation (Reference 8) did not include an evaluation of the prompt recirculation trip (RPT) proposed by BSEP 2.
Because several issues remain unresolved regarding the implementation of the proposed E0C RPT system at BSEP 2, CP&L requested approval with no credit for the ECC RPT thermal margin improvements (Reference 12).
Our review and approval of the BSEP 2 operating limits are discussed in Section 3.2.2.
3.0 Evaluation 3.1 Nuclear Characteristics For Cycle 3 operation of BSEP2, 64 (8x8R) fuel bundles of type 8DR B 265H and 68 (8x8R) bundles of type 80R B 283 will be loaded into the core (Reference 1).
The remainder of the 560 fuel bundles in the core will be fuel used during the previous cycle.
The fresh fuel will be loaded in a core pattern as shown in Figure 1 of Reference 1, which is acceptable.
Based on the data presented in sections 4 and 5 of Reference 1, both i
the control rod system and the standby liquid control system will have acceptable shutdown capability during Cycle 3.
3.2 Thermal Hydraulics 3.2.1 Fuel Claddina Inteority Safety Limit As stated in Reference 6, the minimum critical power ratio (MCPR) which may be allowed to result from core-wide or localized transients i s 1.07.
This limit has been imposed to assure that during transients 99.9% of the fuel rods will avoid transition boiling.
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. raised to 1.07 because the The safety limit MCPR for BSEP2 is being(8x8R) fuel bundles is flatter distribution of fuel rod power with the than that of the (8x8) fuel. The reas for the flatter power distribution is the presence of two ratner than one water rods in (8x8R) fuel.
The issue has been addressed in Reference 8 and the 1.07 limit has be~en found acceptable for BURS with uncertainties in flux monitoring and operational parameters no greater than those listed in Table 5-1 of Reference 6, for which the CPR distribution is within the bounds of Figures 5.2 and 5.2a of Reference 6.
It has been shawn in Reference 1 that these conditions are met for BSEP2 Cycle 3.
In addition to the 1.07 MCPR safety limit discussed above, the (8x8) and (8x8R) fuel must be maintained within the 17.5 KW/ft exposure-dependent Linear Heat Generation Rate (LHGR) safety limit.
Maximum LHGR conditions can occur during abncrmal operational conditions which af fect the fuel locally, e.g., Rod Withd wal Error and the Fuel Loading Error.
In this regard, the staf r r,ie J that the calculated maximum transient thGR for the 8x8 and 8x8R fuel be As stated augmented by a fuel densification power spike allowance.
in Reference 11 since implementation of this requirement for CSEP2 meets the exposure-dependent safety limit for the 8x8 and 8x8R fuel, the staff finds it acceptable that the 8x8 and 8x8R fuel densifi-cation power spike penalty be deleted from the BSEP2 Technical Specifications.
Because the (7x7) fuel was designed before fuel densification and its effects were known, the newly implemented and revised GE analytical procedures to mechanistically account for densification power spiking do not apply to the (7x7) fuel.
Therefore, the power spiking penalty, as included in the Technical Specifications, shall continue to be used for the (7x7) fuel.
3.2.2 Operating Limit MCPR Various transients could reduce the CPR below the intended operating limit MCPR during Cycle 3 operation. The most limiting of these operational transients and also the potential fuel loading errors have been analyzed by the licensee to determine which event could induce the largest reduction in the critical power rat.o (ACPR).
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. The transients evaluated were the generator load rejection without bypass, turbine trip without bypass, feedwater controller failure at maximum demand, imdvertent HPCI pump startup, and the control rod withdrawal error.
!qitial conditions and transient input parameters as specified in Tables f and 7 of References 1 and 2 were assumed.
In the analyses of these reactor transients, the licensee submitted results based on transients that include the prompt RPT feature, and transients which take no credit for the RPT being operable.
As stated in Section 1.0, several issues remain unresolved relative to the implementation of the proposed E0C RPT system.
Therefore, we cannot give credit for the reductions in operating limits (MCPR) afforded by the E0C RPI feature.
Since the turbine trip and generator load rejection transients without bypass represent the limiting transients only near the E0C, with all control rods withdrawn, we can approve the earlier part of Cycle 3 (B0C to E0C - 2000 MWD /t) based on other transients which are most limiting over this interval.
For the later part of Cycle 3, (E0C - 2000 MWD /t to E0C), the operating limit MCPR will be based on the most limiting of these two transients without benefit of the E0C RPT reductions in thermal margin (ACPR).
However, even though we cannot give credit at this time for the E0C RPT installation at BSEP 2, we do believe it prudent that CP&L perform functional testing of the E0C RPT operational aspects, e.g., flow coastdown rate and time response measurements during their startup test program.
These tests should give the necessary information to provide assurance that the E0C RPT system will perform within the bounds of the analysis.
In addition, even though we have not g uen credit for this feature in this safety analysis for the reasons previously stated, we recognize the potential benefits afforded by the immediate reduction in core flow with increased core voiding and the resultant negative reactivity.
Therefore, until we can approve the implementation of the E0C RPT system. at BSEP 2, operation of the "as built" E0C RPT should provide an extra margin of conservatism in BSEP 2 operating limits.
We thus have no objection to the use of this system during Cycle 3,5 provided the licensee perfcrms the appropriate 10 CFR 50.59 evaluation.
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. As shown below, addition of the highest ACPR resulting from the most severe transient during the specified exposure interval to the safety limit (1.07) gives the appropriate operating limit MCPR for each fuel type.
This sum will assure that the safety limit is not violated during Cycle 3 operation at BSEP 2.
Limiting Exposure ACPR (7x7)/
MCPR Operating Transient Interval (8x8)/(8x8R)
Limit w/o RPT Rod with-BOC to EOC-2.13/**/.19 1.20*/**/1.26 drawal error GWD/t Inadvertent BOC to EOC-2 **/.14/**
- /1.21/**
HPCI Pump GWD/t start Generator Load EOC-2 GWD/t
.14/.20/.20 1.21*/1.27/1.27 rejection w/o to E0C bypass
- For the 7x7 fuel, the 102.5% core flow Kr curve is nonconservative (Reference 6) with operating limits <l.23.
Therefore, at reduced flow conditions, the Kf factor for the 7x7 fuel assemblies will be based on the 112% flow curve of Figure 3.2.2-1 of the Technical Specifications rather than the actual setpoint of 102.5%.
3.3 Overpressure Analaysis The overpressure analysis for the MSIV closure with high flux scram, which is the limiting overpressure event, has been performed in accordance with the requirements of Reference 8.
As specified in Reference 8, the sensitivity of peak vessel pressure to failure of one safety valve has also been evaluated.
We agree that there is sufficient margin between the peak calculated vessel pressure and the overpressure design linit (1375 psi) to allow for the failure of at least one valve.
Therefore the limiting overpressure event as analyzed by the licensee is acceptable.
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. 3.4 Thermal Hydraulic Stability The results of the themal hydraulic stability analysis (Reference 1) show that the channel hydrodynamic and reactor core decay ratios at the Natural Circulation - 105% Rod Line intersection (which is the least stable physically attainable point of operation) are below the 1.0 stability limit.
Because operation in the natural circulation mode is restricted by Technical Specifications, there will be added margin to the stability limit. We find this is acceptable.
3.5 Accident Analysis 3.5.1 ECCS Appendix K Analysis On December 27, 1974, the Atomic Energy Commission issued an Order for Modification of License, implementing the requirements of 10 CFR 50.46, " Acceptance Criteria and Emergency Core Cooling Systems for Light Water Nuclear Power Reactors." One of the require-ments of the Order was that prior to any license amendment authorizing any core reloading..."the licensee shall submit a reevaluation of ECCS performance calculated in accordance with an acceptable evaluation model which confoms to the provisions of 10 CFR Part 50.46."
The Order also required that the evaluation shall be accompanied by such proposed changes in Technical Specifications or license amend-ments as may be necessary to implement the evaluation assumptions.
The licensee has reevaluated the adequacy of ECCS performance in connectior with the new reload fuel design, using methods previously approved by the sta ff.
The results of these analyses are given in References 1, 2, and 3.
We have reviewed the information submitted by the licensee and con-clude that all requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 will be met when the reactor is operated in accordance with the MAPLHGR versus Average Planar Exposure values given in Figures 3.2.1-1, 2, 3, 4, 5, 6, and 7 of the Technical Speci fications.
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. 3.5.2 Control Rod Drop Accident For BSEP2, Cycle 3, the accident reactivity insertion curves satisfy the requirements for the bounding analyses described in Reference S.
Therefore, the peak fuel enthalpy for this event would be less than 280 calories / gram, which is acceptable.
3.5.3 Fuel Loading Error Potential fuel loading errors involving misoriented bundles have been explicitly included in the calculation of the operating limit MCPR.
Potential errors involving bundles loaded into incorrect positions have also been analyzed by a method which consids.rs the initial MCPR of each bundle in the core, and the resultant MCPR was shown to be greater than 1.07.
The GE method for analysis of misoriented and misloaded bundles has been reviewed and approved by the staff (Reference 10).
The analyses which have been performed for potential fuel loading errors for BSEP2, Cycle 3, are acceptable for assuring that CPRs will not be below the safety limit MCPR of 1.07.
4.0 Physics Startup Testing The safety analysis for the upcoming cycle is based upon a specifi-cally designed core configuration. We have assumed that, af ter re-loading, the actual core configuration will conform to the designed configurr ion. A startup test program can provide the assurance that the core conforms to the design. We requi e that a startup test program be performed and the minimum recommended tests are:
1.
Visual inspection of the core using a photographic or videotape r' cord.
2.
A check of core power symmetry by checking for mismatches be-tween synmetric detectors.
3.
Withdrawal and insertion of each control rod to check for criticality and mobility.
4 Comparison of predicted and measured critical insequence rod pa ttern for nonvoided conditions.
We find the startup test program, (Reference 5), submitted by CP&L acceptable for Cycle 3 operation.
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i o
. In the future, as a result of our ongoing generic review of BWR startup test, we anticipate requiring a description of each test sufficient to show how it provides assurance that the core conforms to the design.
The description is anticipated to include both the' acceptance criteria and the actions to be taken in case the accept-ance criteria are not obtained.
In addition to the requirements, above, we request that a brief written report of the startup tests be submitted to the NRC within 45 days of the completion of the tests.
Environmental Considerations We have determined that this amendment does not autnorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that this amendment involves an action.which is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR !i51.5(d)(4) that an environmental impact statement, or negative declaration and environ-rental impact appraisal need not be prepared in connection with the issuance of this amenJment.
Conclusion We have concluded, based on the consideratione discussed above, that:
(1) because the amendment does not involve a significant increase in t!.e probability or consequences of accidents previous 1y l
considered and does not involve a significnat decrease in a safety margin, the change does not involve a significant hazards considertion, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities wd be conducted in compliance with the Commission's regulations and t.he issuance of this amendment will not be inimical to the common de'ense and security or to the health and safety of the public.
2232 ie Dated: May 2,1979
. References 1.
"Eupplemental Reload Licensing Submittal" for Brunswick Unit 2 Reload 2, NED0-24179-1, March 1979.
2.
" Supplemental Reload Licensing Submittal NED0-24182, March 1979.
3.
James A. FitzPatrick Nuclear Power Plant, July 1977 (NED0-21662).
4.
Letter, E. E. Utley (CPL) to T. A. I.,polito, (NRC), dated March 27, 1979 requesting deletion of Power.giking Penalty.
5.
Letter, E. E. Utley (CPL) to T. A. Ippolito, (NRC), dated March 16, 1979 transmitting Physics Startup Test Program.
6.
General Electirc Boiling Water Reactor Generic Reload Fuel Application, NED0-240ll-P, May 1977.
7.
General Electric Boiling Water Reactor Generic Reload Application for 8x8 Fuel, NEDO-20360, Rev.1, Supplement 4, April 1,1976.
8.
Safety Evaluation of the GE Generic Reload Fuel Application (NEDE-240ll-P), April 1978.
9.
Status Report on the Licensing Topical Report " General, Electric Boiling Water Generic Reload Application for 8x8 Fuel," NED0-20360, Revision 1 and Supplement 1 by Division of Technical Review, Office of Nuclear Reactor Regulation, United States Nuclear Regulatory Commission, April 1975.
10.
Safety Evaluation of New GE Fuel Loading Error Methods, April 1978.
11.
Letter, D. G. Eisenhut (NRC), to R. Gridley (GE), dated June 9,1978, transmitting: Safety Evaluation of the General Electric Methods for the Consideration of Power Spiking due to Densification Effects in BWR 8x8 Fuel Design and Performance.
12.
Letter, E. E. Utley (CP&L) to T. A. Ippolito (NRC), dated May 1,1979, requesting operating limits with no credit for the EOC. recirculation pump trip.
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