ML20057C115

Submits Response to GL 93-04, Rod Control Sys Failure & Withdrawal of Rod Control Cluster Assemblies

ML20057C115
Person / Time
Site:	Prairie Island
Issue date:	09/20/1993
From:	Richard Anderson NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-93-04, GL-93-4, NUDOCS 9309270084
Download: ML20057C115 (5)

Text

l Northem States Power Company 414 Nicollet MaH Minneapolis, Minnesota 55401-1927 Telephone (612) 330-5500 September 20, 1993 Generic Letter 93-04 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Response to NRC Generic Letter 93-04 Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies Pursuant to the requirements of 10 CFR 50.54(f), the NRC issued Generic Letter 93-04, " Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies," on Monday, June 21, 1993, addressed to all licensees with the Westinghouse Rod Control System (except Haddam Neck) for action and to all other licensees for information.

The generic letter requires that, within 45 days from the date of the generic letter, each-addressee provide an assessment of whether or not the licensing basis for each facility is still satisfied with regard to the requirements for system response to a single failure in the Rod Control System (GDC 25 or equivalent). If the assessment (Required Response 1.(a)) indicates the licensing basis is not satisfied, then the licensee must describe compensatory short-term actions consistent with the guidelines contained in the generic letter, and within 90 days, provide a plan and schedule for long-term resolution. Subsequent correspondence between the Westinghouse Owners Group and the NRC resulted in schedular relief for Required Response 1.(a)(NRC Letter to Mr. Roger Newton dated July 26, 1993).

, Our letter to NRC dated August 5, 1993 provided our 45-day response. The response summarized the compensatory actions taken in response to the Salem rod control syste a failure event (the second part of Required Response 1.(b)).

It also provided a summary of the results of the generic safety analysis program conducted by the Westinghouse Owners Group and its applicability to Prairie Island Units 1 and 2.

We hereby submit our 90-day response to the Generic Letter as it applies to Prairie Island. The attached response concludes that the licensing basis is satisfied for GDC 25 (or equivalent) (Required Response 1.(a)) and also provides additional information for long-term clarification of this issue.

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t Northem States Power Company USNRC ,

Sep tembe.r 20, 1993 Page 2 The safety assessment that was provided in the 45-day response confirmed that there is no safety significance for any asymmetric RCCA withdrawal by using three-dimensional safety analysis.

In this response we are making the following new NRC commitments: ,

i We will continue to perform the current order surveillance that is recommended by WOG. ,

f We will modify the rod control systems current order timing to prevent any uncontrolled asymmetric rod withdrawal in the event of the failure identified at Salem. The modifications will be made at the next scheduled refueling shutdown of each unit following receipt of the official technical bulletin from Westinghouse.

Please contact Jack Leveille (612-388-1121, Ent 4662) if you have any questions related to our response.

! , /^ l h 17[d bd~c 6' h Ro[ger0 Anderson Director ,

Licensing and Management Issues c: Regional Administrator - Region III. NRC Senior Resident Inspector, NRC NRR Project Manager, NRC J E Silberg Attachments >

1. Affidavit

2. Generic Letter 93-04 Response sem .n l

9 UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PIANT DOCKET NO. 50-282 50-306 ROD CONTROL SYSTEM FAILURE AND WITHDRAWAL OF ROD CONTROL CLUSTER ASSEMBLIES Northern States Power Company, a Minnesota corporation, with this letter is submitting information requested by NRC Generic Letter 93-04. -

i This letter contains no restricted or other defense information.

1 NORTHERN STATES POWER COMPANY By '

/ 4/[. / 44M5r '

Jt6ger O Anderson Director Licensing and Management Issues b

Onthis[d"dayof _ ] N before me, a notary public in and for said County, personall appeared Roger 0 Anderson, Director, Licensing and Management Issues; and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interpose- for .1 = r r - -

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e RESPONSE TO NRC GL 93-04 Assessment of L'icensine Basis Compliance The purpose of this response is to provide an assessment of whether or not the licensing basis for Prairie Island Units 1 and 2 is still satisfied with regard to the requirements for system response to a single failure in the rod contro1' 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)).

The Westinghouse Owners Group (WOG) has undertaken the following 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, the WOG has concluded 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. The WOG believes 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 of less than a whole group 11s on the order of 4E-8/ critical reactor hours. This would indicate that the-likelihood of a Salem-type event is extremely remote. With this in mind, the WOG would propose that a Condition III (or IV) specified acceptable fuel design limit would be applicable.

Based on 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 uithdrew 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 the 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 fue]

design-limit being exceeded, in this case the DNB design basis, would remain aww

Anachnum 2 Septender30, 1993

• Pp2d2 satisfied.

Assessment of the Safety Sinnificance of Potential Asymmetric Rod Motion in  !

the Rod Control System 4

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 plants for a Salem-type rod withdrawal.  :

Our letter to NRC dated August 5, 1993 provided our 45-day response. The  !

response provided a summary of the results of the generic safety analysis ,

program conducted by the Westinghouse Owners Group and its applicability to  !

Prairie Island Units 1 and 2.  ?

Long-term Enhancements

'!h?Ie the assessment indicates that the licensing basis'is currently ,

satisfied, we believe that there are measures that can be taken to make compliance with GDC 25 more clear. WOG recommendations include.a combination of Rod Control System logic cabinet changes (current order timing adjustments) and an additional plant surveillance, or FSAR safety analyses analyzing asymmetric rod withdrawal and an additional plant surveillance, i The current order surveillance recommended by WOG has been in place at Prairie i Island since original plant startup and we will continue to perform this surveillance.

We will modify the rod control systems current order timing to prevent any uncontrolled asymmetric rod withdrawal in the event of the failure identified at Salem. The modifications will be made at the next scheduled refueling shutdown of each unit following receipt of the official technical bulletin from Westinghouse.

The schedule for implementation of the modification 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 existing rod motion surveillance tests provide assurance that the failure scenarios of.an uncontrolled asymmetric rod withdrawal will be detected, and the analysis program performed and documented in WCAP-13803, Revision 1, concluded that there was no safety significance for affected Westinghouse plants for a Salem-type rod withdrawal.

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