ML20057B982
| ML20057B982 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, Zion File:ZionSolutions icon.png |
| Issue date: | 09/16/1993 |
| From: | Chrzanowski D COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| GL-93-04, GL-93-4, NUDOCS 9309240242 | |
| Download: ML20057B982 (5) | |
Text
f Commonwealth Edison O
- 1400 Opus Place O
Downers Grove, Illinois 60515 September 16,1993 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention: Document Control Desk
Subject:
Supplemental Response to Generic Letter 93-04, " Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies."
Byron Station Units 1 and 2, (NRC Docket Numbers 50-454 and 50-455)
Braidwood Station Units 1 and 2, (NRC Docket Numbers 50-456 and 50-457)
Zion Station Units 1 and 2, (NRC Dockets 50-295 and 50-304)
Reference:
Letter from D.J. Chrzanowski to U.S. NRC dated August 4, 1993.
The purpose of this letter is to provide the 90 day, supplemental response to Generic Letter 93-04. In the attachment to this letter is the assessment that discusses the Byron, Braidwood and Zion compliance to GDC 25 as well as the Commonwealth Edison (CECO) commitment to review the existing current order surveillances that are in place at the CECO PWRs and, if necessary, enhance the current order acceptance criteria. Also, CECO will evaluate the Westinghouse Owners Group (WOG) proposed current order changes after successful demonstration of the adjustments at an operating plant.
The referenced letter provided the initial,45 day response to the Generic Letter. That 45 day response discussed the compensatory actions taken at Byron, Braidwood and Zion Stations in response to the Salem Rod Control event and also provided a discussion of the results of the generic safety analysis program conducted by the WOG and its applicability to Byron, Braidwood and Zion Stations.
CECO will notify the NRC of the results of the CECO evaluation of the proposed current order changes for Braidwood, Byron and Zion Stations when these changes have been successfully implemented at an operating plant.
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S N.S. Nuclear Regulatory Commission September 16,1993' To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects these statements are not based on my personal knowledge, but on information furnished by other CECO employees, contractor employees, and/or consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.
If there are any questions or comments, please contact me at (708) 663-7292 Sincerely, a
U.
David. Chrzanowski Generic Issues Administrator Nuclear Regulatory Services Attachments:
Braidwood, Byron and Zion Station Response to NRC Generic Letter 93-04 cc:
J. Martin, Regional Admimetrator-RIII J. Hickman, Byron Project Manager-NRR/PDIII-2 R. Assa, Braidwood Project Manager-NRR/PDIII-2 C. Shiraki, Zion Project Manager-NRR/PDIII-2 H. Peterson, Senior Resident Inspector (Byron)
S. DuPont, Senior Resident Inspector (Braidwood)
J. D. Smith, Senior Resident Inspector (Zion)
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ATTACHMENT BRAIDWOOD, BYRON AND ZION RESPONSE TO NRC GENERIC LETTER 93-04 Assessment of Licensine Basis Compliance The purpose of this response is to provide an assessment of whether or not the licensing basis for Braidwood, Byron and Zion Stations is still satisfied with regard to the requirements for system response to a single failure in the rod control system and to provide supporting discussion for this assessment in light of the information generated as a result of the Salem event (Required Response 1.(a)).
CECO has reviewed the following WOG initiatives to support the response to NRC Generic Letter 93-04: conducting Rod Control System testing in the Salem training center, examining the existing Rod Control System Failure Modes and Effects Analysis (FMEA), analyzing the worst-case asymmetric RCCA withdrawal combinations with three-dimensional analytical methods, and performing an equipment survey of Westinghouse plants to determine the frequency and significance of control system circuit card failures.
After this extensive investigation, CECO agrees with the WOG conclusions that GDC 25 continues to be met, but also recognizes that there are questions as to the interpretation of not only the intent of GDC 25 but also the appropriate definition of the specified acceptable fuel design limit as well.
Based on previous communications, the NRC has conservatively interpreted the GDC 25 fuel design limit to be the DNB design basis. CECO agrees with the WOG belief that this is a conservative definition if applied to all events. The equipment survey conducted by the WOG demonstrated that the failure rate of card failures that could result in the movement ofless than a whole group is on the order of 4 E-8/ critical reactor hours. This would indicate that the likelihood of a Salem-type event is extremely remote. With this in mind, CECO agrees with the WOG proposal that a Condition III (or IV) specified acceptable fuel design limit would be applicable.
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3 CECO concurs with the WOG's understanding of GDC 25, the purpose of this criterion is to ensure that the appropriate limits (commensurate with the probability of occurrence) are not violated for a " worst-case" stand-alone single failure. The test program conducted at the Salem training center demonstrated that all the rods within a given group would receive the same signals. The corrupted current orders generated by the logic cabinet failures at Salem were transmitted identically to all 8 RCCAs in Shutdown Bank A. The fact that only one RCCA withdrew in the plant was due to a second unrelated effect. Had all the rods in SBA responded, as predicted in the existing FMEA, all the rods would have withdrawn uniformly and have been enveloped by existing FSAR accident analyses. In addition, existing rod motion surveillance requirements would detect the type of rod motion failure observed at Salem. Thus, the requirement that one single failure not result in a specified acceptable fuel design limit being exceeded, in this case the DNB design basis, would remain satisfied.
Assessment of the Safety Sienificance of Potential Asymmetric Rod Motion in the Rod Control System Westinghouse has also performed a safety analysis using three-dimensional safety analysis techniques to assist the WOG in its determination of the safety significance of an uncontrolled asymmetric rod withdrawal. WCAP-13803, Revision 1 documented the safety analysis program and concluded that the generic analysis and their plant-specific application demonstrate that DNB does not occur for a worst-case asymmetric rod withdrawal for all affected Westinghouse plants.
As such, the analysis program concluded that there is no safety significance for affected Westinghouse plant for a Salem-type rod withdrawal.
Lone-term Enhancements While the assessment indicates that the licensing basis is currently satisfied, CECO and the WOG believe that there are measures that can be taken by utilities to make compliance with GDC 25 more clear. Those recommended modifications include a combination of Rod Control System logic cabinet changes (current order timing adjustments) and a plant surveillance.
CECO will take the following actions at Braidwood, Byron and Zion. Each station will evaluate existing slave cycler timing surveillances and, if necessary, enhance the current order acceptance criterin. The current order modification will be l
evaluated to ensure that this change can be 'made safely and without i
compromising system integrity. Based on this evaluation, CECO will implement j
the modification or otherwise notify the Commission ofits intent.
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a The CECO schedule for the proposed long-term actions is based on the successful-demonstration of the timing adjustments at an operating plant and receipt of the official technical bulletin from Westinghouse. The basis for allowing this time period is that the existing rod motion surveillance tests provide assurance that the failures scenarios of an uncontrolled asymmetric rod withdrawal will be detected and the conclusions of the analysis program performed and documented in WCAP-13803, Revision 1, determined that there was no safety significance for affected Westinghouse plants from a Salem-type rod withdrawal.
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