ML20072L100
| ML20072L100 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/28/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072L093 | List: |
| References | |
| NUDOCS 8307130025 | |
| Download: ML20072L100 (6) | |
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UNITED STATES j*j.mflgq NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 56 TO FACILITY LICENSE NO. DPR-71 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNIT l DOCKET N0. 50-325 1.0 Introduction
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By letter dated May 2,1983 (Ref.1) Carolina Power and Light Company (the licensee) proposed to change the Technical Specifications for the Brunswick Steam Electric Station Unit 1 to permit its operation for fuel Cycle 4 In support of the application the licensee submitted a Supplemental Reload licensing document (Ref. 2) and a commitment to add certain technical Specifications with regard to the turbine bypass and high reactor water level trip systems (Ref. 9). We have reviewed the licensee's application and prepared the following evaluation.s 2.0 Evaluation 2.1 Fuel System Design The licensee's analysis of the safety considerations involved in the proposed fourth cycle of operation at Brunswick Unit 1 are described in the Supple-mental Reload Licensing Submittal (Ref. 2).
In all fuel-design-related areas, the reload submittal relies on the General Electric generic report, Generic Reioad Fuel Application (Ref. 3).
Reference 3 has been reviewed and approved by the staff (Ref.10). With the exception of the revised Maximum Average Planar Linear Heat Generaticn Rate (MAPLHGR) limits, we conclude that additional staff review of those portions of the generic application con-cerning the standard fuel designs is unnecessary for the Cycle 4 application.
The licensee's submittal provided MAPLHGR limits for the standard 8x8, 8x8R and P8x8R fuel assemblies in the Cycle 4 core. Although~ the methodology used in Reference 3 is generically applicable for the MAPLHGR limit deter-mination, the staff believes that the effects of enhanced fission gas release at high burnup (i.e., greater than 20 mwd /kg/U) were not adequately conside' red in the fuel performance model.
In response to this concern, General Electric (GE) requested (Refs. 5 and 6) that credit for approved, but unapplied, ECCS evaluation model changes and calculated. peak cladding temperature margin be used to avoid MAPLHGR penalties at higher burnups.
This proposal was found 9
8307130025 830628 DR ADOCK 05000325 PDR
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acceptable provided that certain plant-specific conditions were met (Ref. 7).
The licensee has stated (Ref. 8) that the GE proposal is applicable to the Brunswick Unit 1 safety analysis.
On this basis we conclude that the proposed MAPLHGR limits are appropriate for Cycle 4 operation.
2.2 Nuclear Design The nuclear design and analysis of the proposed reload has been performed by thc methods described in Reference 3.
This report has been approved for use in the design and analysis of reloads in BWR reactors and its use is acceptable for this reload.
The results of the nuclear design analysis are consistent with those for similar reloads and are acceptable, 2.3 Thermal Hydraulic Design The obje'c'tive of the review is to confirm that the thermal-hydraulic design of the reload core has been accomplished using acceptable methods, and provides an acceptable margin of safety from conditions which could lead to fuel damage during normal operation and anticipated transients, and the core is not susceptable to thermal-hydraulic instability.
The review includes the following areas:
(1) safety limit minimum critical power ratio (MCPR),
(2) operating limit MCPR and (3) thermal-hydraulic stability.
The licensee has submitted the reload analysis report for Cycle 4 operation (Ref. 2), which is based on the approved GE report (Ref. 3).
i Discussion of the review concerning the thermal-hydraulic design for Cycle 4 operation follows:
Safety Limit MCPR A safety limit MCPR has been established to assure that 99.9 percent of the l
fuel rods in the core are not expected to experience boiling transition I
during normal operation and anticipated transients. The approved safety limit i
MCPR of 1.07 as sated in Ref. 3 was used for the Cycle 4 analyses.
I Operating Limit MCPR (OLMCPR)
Various transients could reduce MCPR below the intended safety limit MCPR l
during Cycle 4 operation.
The most limiting events have been analyzed by l
the licensee to determine which event could potentially induce the largest reduction in the critical power ratio ( ACPR).
The ACPR values given in Section 9 of Ref. 2 are plant specific values calculated by using the methods given in Ref. 3 including ODYN methods.
The calculated ACPRs were adjusted to reflect "A" Option or "B" Option ACPRs by employing the conversion ;
method described in Ref 4.
The cycle MCPR values are determined by adding the ACPRs to the safety limit MCPR.
Section 11 of Ref. 2 presents both the cycle MCPR values for the pressurization and non-pressurization events.
The maximum cycle MCPR values (for "A" Option and "B" Option) in Section 11 are specified as the operating limit MCPRs and are incorporated into the Technical Specifications.
3 The OLMCPR must be >1.21 for 8x8 fuel and >l.25 for 8x8R and P8x8R fuel types at fuel exposures from Begining of Cycle (B0C) to End of Cycle (E0C) minus
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2000 MWD /t.
For fuel exposures from E0C minus 2000 MWD /t to E0C, the OLMCPR must be >l.26 for 8x8 f.uel,>l.27.for 8x8R fuel and >l.29 for P8x8R fuel.
Thermal-Hydraulic Stability The results of thermal-hydraulic analyses (Ref. 2) show that the maximum core stability decay ratio is 0.72 for Cycle 4 as compared to 0.74 for Cycle 3 We find that (1) the calculated decay ratio for Cycle 4 is less than that for the approved Cycle 3 operation and (2) operation for more that 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.s.
in the natural circulation mode will be prohibited by Technical Specification 3/4.4.1. We therefore conclude that the thermal-hydraulic stability results are acceptable for Cycle 4 operation.
2.4 Transient and Accident Analyses The rcd withdrawal error, fuel misorientation event and rod drop accident have been analyzed for this cycle.
The cycle specific rod drop accident analysis was necessary because certain parameters (accident reactivity shape function and scram shape function in the cold startup mode) were not bounded by the generic analysis. The results of the cycle specific analysis (220 calories per gram peak enthalpy) meets o.ur acceptance criterion (280 calories per gram) for this event and is therefore accepta.b' e.
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l. The' fuel, mis The rod withdrawal. orientation..
event is not limiting at any time in the cycle.
event is the limiting event with respect to 0LMCPR during the portion of the cycle prior to 2000 MWD /t before End-of-Cycle if the ODYN Code OPTION B Transient Analysis is used.
Since approved methods have been used to perform the analyses and to obtain input parameters for them, we conclude that the transient analyses of the three cited events is acceptable.
For discussion of core-wide transient analyses, see OLMCPR above.
2.5 Technical Specification Changes The control rod average scram insertion times were changed to correct errors that were introduced when the designation of control rod position was changed from percent of rod insertion to rod notch position.
The designation was changed in order to conform to the BWR Standard Technical Specifications but the conversion of the positions and times was not accurate.
The new values agree with the original bases for insertion times and are therefore, acceptable.
Changes were made in Technical Specifications 3/4.2.2 and 3/4.2.4 in order to establish the operating limits on the flow-biased APRM scram set point based on new values of the Design Total Peaking Factor.
These new values are consistent with the previously established linear heat generation rate (13.4 KW/
Ft) and are therefore, acceptable.
In addition, the editorial changes are correct.
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Section 3/4.2.3 and Table 3.2.3.2-1 of the Technical Specifications have been modified to include the operating limit MCPR for Cycle 4 operation.
We find that for the limiting event for "B" Option, feedwater controller failure that causes maximum demand, credit is assumed for operation of the high water level (L8) trip and turbine bypass system. Accordingly, we require that I
Technical Specifications be included to ensure the operability of these systems.
To assure an acceptable level of performance, our position is that this equipment (the turbine bypass system and the level 8 high water level trip) should be identified in the Technical Specifications with limiting conditions for operation, surveillance requirements and with adequate degree of powep-reduction, in case of inoperability.
The licensee has agreed to submit.,
within 30 days of startup of Cycle 4 operation, the proposed changes to the Technical Specifications that will incorporate the Limiting Conditions for Operation and surveillance requirements for the turbine bypass and high water level trip systems (Ref. 9).
Based on our review of the significance of these trip systems with regard to the limiting transients during this fuel cycle, we find this commitment acceptable.
3.0 Summary of Evaluation The following statements summarize the staff conclusions concerning the Cycle 4 reload for Brunswick Unit 1.
Based on our previous generic approval of the fuel-related design features of BWR reloads provided by General Electric, we find these features to be acceptable for Brunswick Unit 1 Cycle 4.
On the basis that certain plant specific conditions relating to extended fuel burnup have been met, we conclude that extension of the MAPLHGR limits to higher burnup values;is acceptable for Cycle 4.
In the course of our review, we found that approved methods were used and that the results are consistent with those for similar reloads.
We there-fore, conclude that the nuclear and thermal-hydraulic design of the reload is acceptable. On the same basis we conclude that the transient and accident analyses are acceptable and that the results support the operation of Brunswick Unit 1 for Cycle 4.
We further conclude that the proposed changes in the plant Technical Specifi-cations are acceptable.
In summary, we. conclude that Brunswick Unit 1 may be operated for Cycle 4 without undue hazard to the public health and safety.
3.1 Environmental Consideration We have determined that the amendment does not authorize 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 the amendment involves an action which is in-significant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmental impact statement, or negative de-claration and environmental impact appraisal need not be prepared in connection with the issuance. of the amendment.
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3.2 Final Determination of No Significant Hazards Consideration On May 19, 1983 the Commission published a notice in the Federal Register (48 FR 22658) seeking public comment on its proposed determination that No public this amendment involves no significant hazards consideration.
comments were received.
The state of North Carolina was consulted on this matter and had no comments on the proposed determination.
The Commission has provided examples of amendments that are considered not likely to involve significant hazards considerations (48 CFR 14870).
One such amendment is a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the Commission for a previous core at the facility in question are involved.
In reviewing the application for this amendment, we found that this refueling of the Brunswick Steam Electric Plant, Unit 1 involves no fuel assemblies significantly different from those used in a previous refueling and found acceptable to the Commission.
Furthermore, we found that there are no significant changes to the acceptance criteria for the technical specifica-tions.
As discussed above, the analytical n athods used to establish the changes in the technical specifications and Jemonstrate conformance with the regulations have been found previously acceptable to the Commission.
None of the char.ges set forth in this amendment involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of a new or different kind of accident from any decident previously evaluated; or involve a significant reduction in a margin of safety.
Therefore, the Commission has made a final determination that this amendment does not involve a significant hazards consideration.
4.0 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 Commission's regulations and the issuance of this amendment will not be inimical to the ccmmon defense and security or to the health and safety of the public.
Dated: June 28,1983 Principal Contributors: J. Voglewede W. Brooks S. Sun
.GL References 1.
Letter from P. W. Howe (CP&L) to D. Vassallo (NRC), Request for Revision to Technical Specifications (Fuel Cycle No. 4-Reload Licensing), May 2,1983.
2.
Y1003J01 A53, Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plar. Unit 1, Reload 3 (Without Recirculation Pump Trip), January 1983, submitted by letter dated May 6,1983 from S. R. Zimmerman (CP&L)' to DJ B. Vassallo (NRC).
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3.
NEDE-24011-P-A-4, General Electric Boiling Water Reactor Generic Reload Fuel Applications, January 1982.
4.
Letter from R. Bucholz (GE) to P. Check (NRC), " Response to NRC Request for Information on ODYN Computer Model", September 5,1980.
5.
R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May 6,1981.
6.
R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May 28, 1981.
7.
L. S. Rubenstein (NRC) memorandum for T. M. Novak (NRC) on " Extension of General Electric Emergency Core Cooling Systen Performance Limits" dated June 25, 1981.
8.
P. W. Howe (CP&L) letter to D. B. Vassallo (NRC) dated June 7,1982.
9.
P. W. Howe '(CP&L) letter to D. B. Vassallo (NRC) dated June 20, 1983.
- 10. D. G. Eisenhut (NRC) letter to R. Gridley (GE) May 12, 1978
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