ML18100A632

Provides ninety-day Response to GL 93-04,as follow-up to Initial Response to Subj Generic Ltr on 930729,including Status of Compensatory Actions Detailed in 930617 Request for Emergency Amend

ML18100A632
Person / Time
Site:	Salem
Issue date:	09/21/1993
From:	Labruna S Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-93-04, GL-93-4, NLR-N93152, NUDOCS 9309290297
Download: ML18100A632 (10)

Text

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• ' Public Service Electric and Gas Company

. /

Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1700 Vice President - Nuclear Engineering SEP2 l 1993 NLR-N93152 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:

RESPONSE TO GENERIC LETTER 93-04 ROD CONTROL SYSTEM FAILURE AND WITHDRAWAL OF ROD CONTROL CLUSTER ASSEMBLIES SALEM GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-272 AND 50-311 This letter is to provide Public Service Electric and Gas's

{PSE&G) ninety-day response to the subject generic letter concerning the Rod Control System failure and withdrawal of a rod control cluster assembly which occurred at our Salem Unit 2 facility. This response is submitted in accordance with the requirements of 10CFR50.54(f).

Initial PSE&G response to the subject generic letter was submitted on July 29, 1993 (Ref. NLR-N93122) and referenced a previous submittal (Ref. NLR-N93098) which included assessment of the single failure on the rod control system, impact on the current licensing basis, and compensatory short-term actions to address the degraded or nonconforming conditions. Since the response indicated that the initial licensing basis was not satisfied, the generic letter requires a plan and schedule for the long-term resolution be provided within ninety days. This letter is to provide the requisite information.

On June 17, 1993, PSE&G submitted a request for an Emergency License Amendment. This amendment request conservatively treated the identified failure as a single failure and classified the event as a ANSI 18.2 Condition II (Moderate Frequency) event. It was concluded that Departure from Nucleate Boiling {DNB) design limits for the fuel continued to be met for all affected accident analyses and that this failure is detectable based upon the measures taken. As a result, contJnued compliance with General Design Criteria 25 was ensured as w~tl as the safe operation of Salem Uni ts 1 and 2. '1-(

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I 9309290297 930921 PDR ADOCK 05000272 P PDR D  :

1-SEP~ l 7993

• Document Control Desk NLR-N93152 PSE&G has been actively working with the Westinghouse Owners Group (WOG) since the identification of the circuitry failure.

Various actions have been initiated to address the long term resolution of this event, including: rod control system testing, generic accident analysis, and failure assessment. As a result of these further efforts a potential long term resolution is being developed involving a timing adjustment in the current orders of the rod control system. This enhancement is described in Attachment 1.

Also as a result of the efforts undertaken since the event's occurrence, it has been concluded that the recent Salem commitment for dilution to criticality is no longer necessary. A Salem specific Justification for Continued Operation (JCO) is being developed so that reactor start-ups for Salem Units 1 & 2 can be performed by pulling rods to achieve criticality. The bases for this is also included in Attachment 1.

As a result of this letter, PSE&G has fulfilled the requirements of the generic letter.

Sincerely, Affidavit Attachment (1)

SEP21 7993

• Document Control Desk NLR-N93152 c Mr. T. T. Martin, Administrator - Region I

u. s. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. J. c. Stone, Licensing Project Manager

u. s. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. S. T. Barr (809)

USNRC Senior Resident Inspector Mr. K. Tosch, Manager, IV NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nucle.ar Engineering CN 415 Trenton, NJ 08625

• Mr. Mark Proviano Westinghouse Electric P.O. Box 355, ECE 4-08 Pittsburgh, PA 15230-0355

REF: NLR-N93152 STATE OF NEW JERSEY )

) SS.

COUNTY OF SALEM

s. LaBruna, being duly sworn according to law deposes and says:

I am Vice President - Nuclear Engineering of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Salem Generating Station Unit Nos. 1 and 2, are true to the best of my knowledge, information and belief.

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7 Subscribed and Sworn to before me

_'this ~ day of ~, 1993

_J  : (l,. 'lf . Notatj~Jif., Jersey My Commission expires on ElliABETH J. KIDD Notary Public of New Jersey My Commission Expires April 25, 1995

.NLR-N93152 ATTACHMENT 1 NINETY DAY RESPONSE TO GENERIC LETTER 93-04 SALEM UNITS 1 & 2 BACKGROUND:

Generic Letter 93-04 requested that licensees of Westinghouse designed reactors take actions based on a failure in the Rod Control System (RCS) experienced at Salem Unit 2. It was found that the system was potentially susceptible to a single failure that could cause an inadvertent withdrawal of control rods in a sequence resulting in a power distribution not considered in the design basis analyses.

Salem's conservative assessment concluded that this postulated single failure in the rod control system was a Condition II event and, therefore, involved an Unreviewed Safety Question. PSE&G requested an Emergency License Amendment on June 17, 1993 which included evaluation of the safety analyses contained in the UFSAR. The evaluation concluded that the DNB design limits for the fuel continued to be met.

In addition, the submittal included several compensatory actions taken prior to startup. These included:

a. Enhanced Rod Control System surveillances prior to startup,

b. For Salem Unit 2, the Technical Specification surveillance was to be performed weekly for two weeks, biweekly for two cycles, and monthly thereafter to provide added level of confidence of the RCS.

c. Modification of the startup procedure to preclude an asymmetric rod withdrawal from the subcritical condition by first pulling control rods while still highly borated to the estimated critical position, then diluting to criticality,

d. Classroom and simulator training addressing the effects of potential single failures in the Rod Control System,

e. Issuance of standing orders to heighten operator awareness of potential rod control system malfunctions, and

f. A review of event response procedures to assure adequate guidance to operators in the event of a Rod control System malfunction.

.NLR-N93152 Attachment 1 The remainder of this response is to discuss the actions taken or planned to resolve this issue and provide a schedule for completion.

DISCUSSION:

The Westinghouse Owners Group (WOG) has undertaken the following initiatives to support the response to NRC Generic Letter 93-04:

conducted Rod Control System testing in the Salem training center, examined the existing Rod Control System Failure Modes and Effects Analysis (FMEA), and performed a generic assessment of the DNB consequences of uncontrolled asymmetric rod withdrawal using a three-dimensional spatial kinetics/systems transient methods. As a result of these efforts, Salem is able to update the previously submitted evaluation for asymmetric rod withdrawal from subcritical and to provide a potential modification that would preclude any asymmetric rod withdrawal as a result of a failure similar to that experienced on Salem Unit 2.

I. Assessment of Asymmetric Rod Withdrawal From Subcritical In the Emergency License Amendment Request of June 17, 1993, a conservative evaluation of a single failure which resulted in a spurious motion demand coincident with the direction command logic failure was provided. The potential for a spontaneous, asymmetric rod withdrawal was considered extremely unlikely and it was also felt that the operator would have sufficient cognizance of rod movements to preclude significant asymmetric rod movement. An additional administrative measure was included which conservatively modified the approach to criticality procedure by diluting the Reactor Coolant System boron concentration to achieve criticality with all control rods withdrawn to a predetermined configuration.

PSE&G feels that this method of diluting to critical is no longer necessary for the following reasons:

1. An evaluation of possible single failures has been conducted and has concluded that a spurious, (i.e., rod movement without a valid motion demand signal), asymmetric rod withdrawal from subcritical is a multiple failure scenario; one failure resulting in corrupted current orders and a second causing rod movement.

2. Rod surveillances with current signatures, instituted as a compensatory action following the event, is capable of detecting logic card failures.

-NLR-N93152 Attachment 1

3. During the approach to criticality the RCS is in the manual mode and operator attentiveness to available indications would readily identify any abnormal rod movements.

4. Generic safety analyses have been performed by WOG which demonstrate that the DNB design Basis is satisfied for the Rod Withdrawal from Subcritical event.

Per UFSAR Section 3.1, Salem Generating Station is committed to the intent of the General Design Criteria of 10CFR50 Appendix A.

Criterion 25 (GDC 25) , "Protection System Requirements for Reactivity Control Malfunctions," states that the protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

A failure assessment of the RCS logic cabinet current order circuits was performed by WOG to determine which logic cabinet printed circuit board failures can result in movement of less than a group of rods. The identified failures of concern were grouped into eight scenarios, as presented by WOG to the NRC on 9/13/93. None of these failure scenarios included spurious rod movement, (i.e., a demand signal must be present for rod motion).

The RCS is maintained in the manual mode while the reactor is subcritical. Therefore, as a result of the failure assessment and application of GDC 25 stated previously, a single failure resulting in a spurious motion demand coincident with the corrupted current orders need not be considered since they are independent and considered two separate failures. The Asymmetric Rod Withdrawal at Power event was previously evaluated and determined not to exceed DNB safety limit or result in fuel failure.

Consistent with Westinghouse safety analysis methodology, random single failures of control systems are not considered provided they are detectable during normal operation or surveillance testing. Of the eight failure scenarios, all can be detected by the rod surveillance that was implemented at Salem before every startup which includes taking current traces of the RCS while moving control rods. PSE&G's review of the current traces is based on Westinghouse guidance. Since all of these potential failures are detectable, it is beyond the present safety analysis basis to assume an additional separate failure in conjunction with corrupted current orders which might result in a spurious rod movement.

As expressed in the evaluation presented in the requested Emergency License Amendment, it is reasonable to conclude that operator action would be expeditiously taken to prevent

.NLR-N93152 Attachment 1 challenging fuel integrity. As the operator withdrawals rods during the ascension to reactor criticality, identification of any abnormal rod movement would almost be immediate due to the continuous observation of the Individual Rod Position Indications, neutron count rate, and the bank demand counters changing both audibly and visually. The action taken would be to cease rod withdrawal and take the applicable actions as required by the Abnormal Operating Procedure S1(2).0P-AB.ROD-003(Q),

Continuous Rod Motion," and reinforced by training exercises.

Once criticality is achieved, any asymmetric rod movement that might occur would be bounded by the Multiple RCCA Withdrawal at Power (Asymmetric) analysis provided in the submittal of June 17, 1993.

It is judged that the above reasons are satisfactory in demonstrating that dilution to critical is no longer a necessary measure to ensure that-the Asymmetric Rod Withdrawal from Subcritical event meets the DNB design basis. However, in addition, WOG has performed a generic set of safety analyses using three dimensional techniques for uncontrolled asymmetric rod withdrawal which concludes that the DNB design basis is met.

This analyses was included in reports WCAP-13803, Rev. 1

[Proprietary] and 13838 [Non-proprietary] entitled "Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal" and was provided to the NRC for information purposes on August 26, 1993 via WOG letter OG-93-70. Though this WOG analysis is generic to all Westinghouse designed plants, PSE&G has informally determined that it encompasses Salem specifically and demonstrates that Departure from Nucleate Boiling (DNB) safety limit will not be exceeded. A formal determination will be incorporated into a Justification for Continued Operation (JCO) which is discussed further in the following paragraph.

Based on the above, there exists a sufficient technical basis for no longer performing the compensatory action described previously of diluting to criticality, that is: 1) two failures are required for a spurious asymmetric rod movement from subcritical,

2) logic failures are detectable via the surveillance performed,

3) operator actions would prevent challenging fuel integrity, and

4) generic analysis has shown that the DNB safety limit will not be exceeded even if an asymmetric rod withdrawal event is postulated with no operator response credited. PSE&G will incorporate this basis into a Justification for Continued Operation (JCO) for Salem Station. Our commitment to dilution to criticality will continue until the JCO has been implemented.

- NLR-*N93152 Attachment 1 II. Long Term Enhancements The WOG is pursuing a possible design change to the rod sequencing circuitry that would ensure a failure in the Rod Control system would not result in asymmetric rod withdrawal.

This possible design change involves timing adjustments of the current orders such that if corrupted current orders are present, the rods will not move asymmetrically. PSE&G is working with WOG and the industry on the feasibility of this resolution. It is anticipated that validation of this design change will be completed by the end of 1993.

If successful, implementation by PSE&G would be expected during refueling outages 2R8 and 1R12, currently scheduled for fall of

'94 and spring of '95 respectively. Should a modification be implemented, this event would then be considered a Condition III event bounded by the initial Salem analyses, (previous to the Emergency License Amendment Request of June 17, 1993). It is anticipated that a change back to the initial Licensing bases will be submitted at that time.

The schedule for implementation of the proposed long-term action is based on the successful validation of the timing adjustments and receipt of the official technical bulletin from Westinghouse.

III. Status of Compensatory Actions The status of the compensatory actions detailed in the Emergency License Amendment Request is included below.

a. Salem shall continue to perf~rm the Rod Control System surveillance with current traces prior to startups.

b. A more frequent periodic surveillance was performed on Salem Unit 2 and has demonstrated correct operation of the RCS. This surveillance has returned to the normal monthly frequency as a result of its satisfactory completion.

c. It has been determined that diluting the Reactor Coolant system boron to achieve criticality is not necessary as detailed in Section I, but will continue until implementation of a JCO.

d. Classroom and simulator training addressing the effects of potential single failures in the rod control system have been incorporated into the licensed operator training program.

e. The standing orders to heighten the operator awareness of potential single failures in the RCS shall remain in place.

-NLR-N93152 Attachment 1

f. Review of event response procedures was completed prior to the request for the Emergency License Amendment. It was concluded that the existing procedures provide adequate guidance and no further action was required.

CONCLUSION:

To summarize, the Emergency Licensing Amendment Request submitted on June 17, 1993 remains valid at this time. Sufficient bases has been developed to eliminate the administrative measure to dilute to achieve reactor criticality. A Salem specific JCO will be developed providing this j*ustif ication prior to discontinuing this measure. A system modification is being pursued which will preclude an asymmetric rod withdrawal event in the rod control system. Completion of this long term resolution is currently scheduled for the fall of 1994 for Salem Unit 2 after the eighth refueling outage and the spring of 1995 for Unit 1 at the conclusion of the twelfth refueling outage.