IR 05000272/1992004
| ML18096A787 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 06/17/1992 |
| From: | Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Miltenberger S Public Service Enterprise Group |
| References | |
| EA-92-084, EA-92-84, NUDOCS 9206240078 | |
| Download: ML18096A787 (5) | |
Text
Docket Nos. 50-272 50-311 50-354 EA 92-084 June 17, 1992 Public Service Electric and Gas Company ATTN: Mr. Steven Vice President and Chief Nuclear Officer Post Office Box 236 Hancocks Bridge, New Jersey 08038
Dear Mr. Miltenberger:
SUBJECT:
EXERCISE OF ENFORCEMENT DISCRETION FOR MAIN STEAM LINE BREAK REANALYSIS (NRC Combined Inspection Report 50-272/311/354/92-04)
This refers to the subject NRC re~ident inspectio~ to assure. public health and safety conducted by Mr. T. Johnson and other members of the NRC staff at Salem and Hope Creek Generating Stations between March 22, 1992 and May 2, 1992. Within the scope ofthe inspection, the inspectors verified that activities conducted at both Salem and Hope Creek were safely performed. The inspectors also reviewed the restart activities of Salem Unit 2, and concluded that your ~taff was appropriately prepared for the startup of the reactor. The findings of the inspection were previously discussed with Messrs. C. Vondra and J. Hagan of your staff. The inspection report was sent to you on June 8, 1992.
Duripg the inspection, the NRC also reviewed the circumstances associated with a nonconservative analysis of a design basis accident at Salem Unit 2. The condition was
_ ipentified during a Westinghouse reanalysis and reported by your staff to the NRC. This nonconservative analysis, wh~ch existed since the initial licensing of the facility, led to the:
operation of Salem Unit 2 during cycles 5 and 6 in a condition outside the design basis, as described in your Licensee Event Report (LER) (No.91-036), dated January 9, 1992.
Specifically, the auxiliary feedwater system was configured such that the flow it would deliver during a steam line break accident could exceed the flow assumed in the most limiting steam line break analysis for the plant. The nonconservative assumption did not result in a condition outside the design basis prior to cycle 5 because excess margin to design limits ort shutdown margin and containment pressure were available to more than compensate for the effects of a higher auxiliary feedwater flowrate without modifying the design. However, for Cycles 5 and 6, sufficient excess shutdown margin did not exist and thus, this condition that existed during cycles 5 and 6 appears to be a violation of NRC requirements. Details are 9206240078 6;886~12 PDR ADOCK PDR G
Public Service Electric and Gas Company provided in Section 10.6.A of the referenced repon. Unit 1 was not affected, since the c0re analysis revealed that sufficient shutdown margin was always available during each operating cycle at that unit.
The nonconservative assumption was identified by a licensee-initiated Westinghou5e reanalysis of the posttilated containment pressure during a design basis accident. As part of that reanalysis. a current piant specific model of auxiliary t"eedwater t1ow was used in the analysis that was based upon empiriCal operating data at Salem. as opposed to mod~ling assumptions used by Westinghouse for rhe original Salem steam line break analysis during initial licensing of the facility. During the reanalysis using the updated auxiliary feedwarer flow numbers, it was determined that the worst case steam line break accident could result in exceeding the containment pressure design iimit.
.
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-Normally, enforcement action is considered for such a condition because it involved the plant operating outside of its design basis. However. after consultation with the Director, Office of Enforcement. as well as the Deputy Executive Director for Nuclear Reactor Regulation, Regional Operations. and Research. the NRC has decided to exercise enforcement discretion, -
pursuant to Section VII.B of the NRC Enforcement Policy (10 CFR Part 2. Appendix C),
and not issue an enforcement action in this case because ( 1) the condition was identified during a licensee-initiated reanalysis and promptly reponed to the NRC; * (2) detection and *
idemhication of the condition was not likely by either your staff or the NRC during routine inspection. review. surveillance, or quality assurance activities; (3) comprehensive corrective -
actions were initiated within_ a _reasonable* time period following identification of this issue:
and (4) the failure to assure that Westinghouse used modeling assumptions consistent with the Salem facility does not rer1ect on current performance.
- *
As pan of your corrective actions. yourApril 22, 1992 letter to the NRC committed to a series of immediately effective administrative controls to assure that sufficient safety margins
- were maintained. In addition. you requested modification of your Technical Specifications for both units to formally establish more conservative Limiting Conditions of Operations, and you also identified planned engineering and design changes to further mitigate the effect of a postulated steam line break event on containment pressure. Furthermore, your supplement to LER No.91-036. dated May 22, 1992, identified the root cause of the use of inaccurate auxiliary feed water values in the steam line break analysis as inadequate design review.
perfonned in 1978. The supplemental LER also indicated that you have initiated an Accident Analysis Profile program (to be completed in 1992), that is expected to augment your current Configuration Control Documentation process in an effort to re-verify and document the validity of all assumptions and analysis used to support the design basis of your facilities; and determine if other discrepancies exist between the safety and design anaiyses, and aetual plant
- Public Service Electric
- and Gas Company
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performance and experience. Finally, you stated that current standards and procedures require detajled review, evaiuation. and verification of design calculations and analysis, and are expected to preclude recurrence of this type of event.
While this condition did demonstrate the importance of proper verification of initial analyses to support the design basis for the facility, your actions in identifying, reanalyzing,.reporting and correeting this-problem demonstrated a high regard for safety and conservatism and you are encouraged to aggressively continue this effort.
No response to this letter is required.
- £~~
Thomas T. Manin
Regional Administrator cc:
S. LaBruna, Vice President, Nuclear Operations
- C. Schaefer, External Operations - Nuclear, Delmarva Power & Light Co.
C. Vondra, General Manager - Salem Operations J. HagaJi, General Manager - Hope Creek Operations F. Thomson,. Manager, Licensing and Regulation L. Reiter, General Manager - QA and Nuclear Safety Review J. Robb, Director, Joint Owner Affairs A. Tapert, Program Administrator R. Fryling, Jr., Esquire M.. Wetterbahn, Esquire J. Isabella, Director. Generation Projects Department, Atlantic Electric Com?any *
J. Lipoti, State of New Jersey Lower Alloways Creek Township K. Abraham, PAO (2)
Public Document Room (PDR)
Local Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector State of New Jersey
Public Service Electric and Gas Company bee:
Region I Docket Room (with concurrences)
Management Assistant, DRMA JJoyner, DRSS EWenzinger, DRP JWhite, DRP JStone, NRR SDembek, NRR BNorris, DRP LCheung, DRS RCooper, DRSS DRS-SALP Coordinator DRSS SALP Coordinator RLobel. OEDO
-. CMiller, PDl-2, NRR WHehl, DRP SStiankman, DRP WHodges, DRS LBettenhausen, DRS
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- Public Service* Electric and Gas Company DISTRIBUTION:
PDR SECY CA JSniezek, DEDR ILieberman, OE TMartin,.RI JGoldberg, OGC LChandler, OGC TMurley, NRR JPartlow, NRR Enforcement Coordinators RI, RII, RIII, RIV, RV.
Fingram, GPA/PA BHayes, OI VMiller, SP DWilliams, OIG EJordan, AEOD JLuehman, OE Day File EA File.
DCS RI:RC Smith DEDR Sniezek R:if :Jlel/f..t
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