ML20034H223
| ML20034H223 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/10/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034H222 | List: |
| References | |
| NUDOCS 9303160195 | |
| Download: ML20034H223 (4) | |
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UNITED STATES j
j-j NUCLEAR REGULATORY COMMISSION
'f WASHINcTON. D.C. 20566-0001 -
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
l RELATING TO THE WOLF CREEK NUCLEAR OPERATING CORPORATION'S REPORT
" RELOAD SAFETY EVALUATION METHODOLOGY
.l FOR WOLF CREEK GENERATING STATION" DOCKET NO. 50-482
1.0 INTRODUCTION
By letter dated November 3,1992, Wolf Creek Nuclear Operating Corporation i
(WCNOC) submitted the report, " Reload Safety Evaluation Methodology for Wolf Creek Generating Station" (Ref. 1). This report and the supplemental information provided by Mr. Forrest Rhodes' letter of February 3,1993 (Ref.
l
- 2) presents the process for the evaluation of reload designs for the Wolf Creek Generating Station. The basic methodology was _ adopted from the NRC l
approved methodology developed by Babcock and Wilcox for the analyses of reload cores for Westinghouse designed plants (Refs. 3 and 4)
The purpose of the reload design evaluation is to confirm that key safety l
parameters for a specific reload cycle are bounded by existing safety.
analysis. The existing safety analysis.is defined as~ the reference analysis which is performed assuming conservative or bounding input parameters. _ The
'I reload safety evaluation confirms the conservatism of the reference analysis compared to specific reload designs or identifies which parameters are not j
bounded by the reference analysis. Those cycle specific parameters which are bounded by the reference analysis are' confirmed to be conservative and require no additional safety determinations. For those parameters which are not bounded by the reference analysis, further analysis will be performed to demonstrate that the predicted value for' a specific reload design will not result in the violation of design or safety criteria.
j Key parameters are identified for each of the transients and accidents analyzed in Chapter 15 of the Updated Safety Analysis Report (USAR). Some parameters are easily compared to specific reload cycle predictions to ensure i
that'the specific reload is' bounded by the reference analysis. 'Several of the USAR Chapter 15 events result in transient core power distributions which are l
not easily compared to a bounding representative core and therefore analyses are performed for each specific reload to demonstrate that~ safety. criteria are satisfied.
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, An important aspect of the nuclear design process is the integration of the various core analyses (transient, thermal-hydraulic, nuclear, and mechanical) to determine the limits for measurable parameters in order to ensure that all steady state and transient safety limits are satisfied. The licensee's topical report and supplemental information described the relationships between the various analyses which had not been addressed in other licensee topical reports (Refs. 5, 6, and 7).
2.0 EVALUATION Changes in cycle length, operating history, and other factors result in cycle specific reload designs varying in terms of both reactivity based parameters, such as the moderator temperature coefficient, and in the distribution of the power generated in the core. The purpose of the reload safety evaluation-process is to demonstrate that a specific reload cycle will not result in the violation of safety limits or other acceptance criteria during steady state or transient conditions.
In general, the reactivity based parameters can be calculated for a specific reload core and compared to the reference or bounding assumptions used in the safety analysis.
The reload can then be shown to be conservative with respect to the reference analysis assumptions related to that parameter or additional analyses can be performed to justify a revised acceptance criteria.
For most transients analyzed, this type of comparison to bounding assumptions can also be utilized to show that power distributions within the core will not result in the violation of safety limits or other acceptance criteria. The licensee provided a list of key parameters which will be included in the reload safety evaluation and the general acceptance criteria to be used in determining that the reference analysis remains bounding (Ref. 2).
Several USAR Chapter 15 transients result in significant perturbations in the core power distribution due to localized additions of reactivity or other asymmetric effects related to control rod positions or temperature changes.
These transients are analyzed each cycle to determine cycle specific power distributions which are in turn input.to thermal-hydraulic analyses to 1
demonstrate that safety limits are not violated.
The nuclear design analysis j
of these transients is performed assuming conservative plant conditions in regard to temperatures, control rod positions, and initial power distributions. The analysis of these transients are combined with the i
evaluation of key parameters in order to ensure that the specific reload cycle will not exceed safety limits or other acceptance criteria during any USAR Chapter 15 transient.
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The nuclear design process includes the simulation of plant maneuvers which result in changes in the core power distributions. The simulated maneuvers are performed using conservative assumptions'related to changes in power, power distribution, and control rod positions.. Combinations of core.
conditions related to variations of power, power distributions, xenon distributions, and control rod positions are compared to maximum allowable-1 peaking criteria that is determined by thermal-hydraulic, fuel mechanical, and loss of coolant accident analyses.
Failure to satisfy the peaking criteria could result in either justification of revised peak.ing criteria or determination of revised limits regarding a parameter such as allowable control rod position. The surveillance methodology for core power distributions is similar to the process approved in the B&W methodology (Ref.
4).
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3.0 CONCLUSION
S The staff has reviewed the WCNOC reload safety evaluation methodology and determined that the process as described in the topical and. supplemental transmittal is similar to the approved process described in References 3 and 4.
Deviations from the approved B&W report which have been introduced by WCNOC for plant specific application of the methodology have been reviewed and determined to be acceptable. Therefore, the WCNOC topical report " Reload Safety Evaluation Methodology for the Wolf Creek Generating Station," as supplemented by the letter dated February 3,1993, has been determined to be acceptable for use by the Wolf Creek Nuclear Operating Corporation for the j
Wolf Creek Generating Station.
4.0 REFERENCES
1.
" Reload Safety Evaluation Methodology for the Wolf Creek Generating Station," Wolf Creek Nuclear Operating Corporation, January 1992.
2.
Letter from Forrest T. Rhodes, Wolf Creek Nuclear Operating Corporation, to NRC dated February 3, 1993.
3.
BAW-10169P-A, "B&W Safety Analysis Methodology for Recirculating Steam l
Generator Plants," B&W Fuel Company, October 1989-1 4.
BAW-10163P-A, " Core Operating Limit Methodology for Westinghouse i
Designed PWRs," B&W Fuel Company, June 1989.
5.
" Qualification of Steady State Core Physics Methodology for Wolf Creek Design and Analysis," Wolf Creek Nuclear Operating Corporation, December 1991.
j 6.
Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to Topical Report TR-90-0025 WOI, Core Thermal-Hydraulic Analysis Methodology for the Wolf Creek Generating Station, October 29,;1992.
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. 7.
" Transient Analysis Methodology for the Wolf Creek Generating Station,"
Wolf Creek Huclear Operating Corporation, January 1991 Principal Contributor: William Reckley Date:
March 10,1993 I
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