ML20137H528

Partially Deleted SECY-94-242,discussing W/Commission Proposed NOV & Proposed Imposition of Civil Penalties in Amount of $500,000

ML20137H528
Person / Time
Site:	Salem
Issue date:	09/16/1994
From:	Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
Shared Package
ML20137H409	List: ML20137H405 ML20137H528
References
FOIA-96-351 EA-94-112, SECY-94-242, SECY-94-242-R, NUDOCS 9704020271
Download: ML20137H528 (76)

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POLICY ISSUE

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(NEGATIVE CONSENT)

SECY-94-242 September 16, 1994 l

E0.t: The Commissioners Erns: James M. Taylor l

! Executive Director for Operations Subiect: PROPOSED $500,000 CIVIL PENALTY TO PUBLIC SERVICE ELECTRIC i & GAS COMPANY CONCERNING VIOLATIONS AT SALEM UNIT 1

(EA 94-112)

Purpose:

To consult with the Commission regarding a proposed Notice of Violation and Proposed Imposition of Civil Penalties in the cumulative amount of $500,000.

i / Consultation with the Commission is required, in accordance with Section i III.(2) of the enforcement policy, since the proposed penalties are in excess of three times the Severity Level I values shown in Table IA of the policy.

i Backaround:

On April 8-26, 1994,'the NRC conducted an Augmented Inspection Team (AIT) inspection at Salem, Unit I to review the circumstances associated with an automatic reactor shutdown and two automatic actuations of the safety injection system on April 7, 1994. While safety-related systems functioned as

'/ required during the event, the occurrence provided challenges to operators and equipment due to preexisting equipment problems and workarounds (i.e., methods

. R1 for coping with known procedural or equipment-related problems that have not been corrected), inadequate training and procedures, and operator errors

.,J Yi i during the transient. Inadequate management attention to these areas f

; i contributed to these problems.

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l i The apparent violations and problems identified as a result of the in e

are described in the enclosed proposal, and were discussed during a

, 7 J !} enforcement conference with the licensee on July 28, 1994. The conference was fa attended by representatives of the media, congressional staff, state l

representatives from New Jersey, Delaware, and Maryland, and members of the

a ,. public.

fk Contacts: J. Lieberman, OE N_QII: ENFORCEMENT RELATED - LIMITED  ;

? - 504-2741 TO NRC UNLESS THE COMMISSION

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1a$S J. Beall, OE DETERMINES OTHERWISE i

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- u J E- 504-3231 9704020271 970324 PDR FOIA O'NEILL96-351 PDR - - . - = = - -

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1 The Commissioners  ;

i Discussion:

Six violations and problems were identified based on the staff review of the The AIT report and a follow-up inspection conducted on May 1-June 25, 1994.

four most significant problems involve three failures to identify and correct significant conditions adverse to quality at the facility and the failure by

' the Senior Nuclear Shift Supervisor (SNSS) and the Nuclear Shift Supervisor (NSS) (the two Senior Reactor Operators on duty in charge of the operations crew) to exercise appropriate command and control of the operations staff and the reactor during the transient. The event and related violations are

' described in the enclosed proposal. The four most significant problems are A cumulative penalty of $500,000 is each classified at Severity Level III.The normal appitcation of the factors in proposed for these four problems.

i the Enforcement Policy resulted in a civil penalty of $200,000. For the l

i reasons described in the enclosed proposal, the staff proposes to exercise enforcement discretion pursuant to 10 CFR Part 2, Appendix C, Section VII.A.

l The staff proposes to escalate three of the problems from the base. civil penalty of $50,000 by $100,000 each to reflect the NRC's conern and to convey the appropriate regulatory message to the licensee, which is the need for the

licensee to change its apparent practice of tolerating degraded conditions and workarounds.

After the July 28, 1994, enforcement conference with the licensee, the staff also scheduled and conducted an enforcement conference on August 2,1994, with Ahe Senior Nuclear Shift Supervisor (SNSS) on duty at the time of the event.

/ The conference was held because of his performance of an activity (bypassing a vacuum permissive interlock by temporarily changing a switch position for about one second) that was not authorized by procedure. During the enforcement conference, the individual indicated that he believed that he could perform the function (which involved a non-safety related system), and /

did not recognize, at the time, that he violated the NRC-required procedure p g

for temporary modifications i

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3-i l Recommendation:

If the Connission does not direct otherwise within two weeks from the date of this paper, the staff intends to issue the proposed civil penalties to PSE&G, ,

and the proposed letter to the SNSS.

Coordination-  !

The Office of the General Counsel has no legal objection to this action. l l '

HQIE:

This paper and its issues should not be publicly disclosed because the matter involves predecisional enforcement issues.

i ecutive Director for Operations l

Enclosure:

Proposed Notice of Violation and Proposed Imposition of Civil Penalties

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i p UNITED STATES

. = n NUCLEAR RERULATORY COMMIS810N i 4

U I WASHINGTON, D.C. SOBBS WM f-l i

Docket Nos. 50-272; 50-311 License Nos. DPR-70; DPR-7T,

! EA 94-124 . .

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Mr. Steven E. Miltenberger i Vice President and Chief Nuclear Officer

! Public Service Electric and Gas Company

' Post Office Box 236 j Hancocks Bridge, New Jersey 08038

Dear Mr. Miltenberger:

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SUBJECT:

NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY -

t $500,000 l (Inspection keports Nos. 50-272/94-80; 50-311/94-80; and

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50-272/94-13; 50-311/94-13) i l This letter refers to the NRC Augmented Inspection Team (AIT) inspection

conducted on April 8-26, as well as a follow-up inspection conducted on May 1

- June 25, 1994, at the Sales Nuclear Generating Station, Hancocks Bridge, New l Jersey. The ins action reports were sent to you on June 24 and July 1, 1994, '

respectively. Tse AIT inspection was conducted to review the circumstances i associated with an event that occurred at Unit 1 on April 7, 1994, involving i an automatic reactor shutdown and two automatic actuations of the safety j injection system. The reactor shutdown was complicated due to problems with spurious steam flow signals; failure of the atmospheric dump valves to .

function properly,' which caused safety relief valves to open; the pressurizer  !

i 9oing solid due to the safety injections; the Pressurizer Power Operated i l Relief Valves opening numerous times to relieve system pressure; and the l Pressurizer Relief Tank rupture disc rupturing due to being overpressurized, i

j Based on staff review of the AIT report, as well as the follow-up inspection, y violations of NRC requirements were identified. The apparent violations were

! provided to you in a letter, dated July 6,1994. On July 28, 1994, an open

enforcement conference was conducted with you and other members of your staff

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to discuss the violations, their causes, and your corrective actions. '

The event occurred on April 7,1994, when marsh grass caused clogging of the circulating water systems travelling screens at the intake structure, which l

) led to a trip of the circulating water pumps. Although the control room

operating staff initiated a power reduction (at a rate of between 1% and 85 i per minute), the Senior Nuclear Shift Supervisor (S455) lost command and L control focus when he personally performed the actions to bypass a vacuum i

! permissive interlock to restart a circulating water pump that had tripped. j

! That activity was not authorized by procedure and required him to leave the i

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! Public Service Electric and Gas Company l

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control room during a critical stage of the transient. Furthermore, while the SNSS was involved with bypassing the pump interlock, the Nuclear Shift Supervisor (NSS), who was in charge at the time, also abandoned his command and control function for a short period when he became directly involved in  ;

the withdrawal of control rods for the purpose of recovering reactor coolant temperature from an overcooling condition. Subsequently, the reactor tripped automatically when power was increased beyond 25% without clearing a control setpoint. j l

The reactor and turbine trip resulted in the sudden closure of the turbine i

stop valves. This caused a pressure wave in the main steam piping that, '

coincident with the low reactor coolant temperature at the time, caused an automatic actuation of the safety injection system and the consequent filling  ;

of the pressurizer steam space. Although the reactor coolant temperature and i i

steam generator pressure subsequently increased (due to core decay heat) to a point where the atmospheric relief valves should have opened, the valves did l not function. As a result, a steam generator code safety valve opened, resulting in a rapid decrease in reactor temperature and reactor pressure, which initiated another safety injection. This safety injection, with an .

already solid pressurizer, required the pressurizer power operated relief i valves (PORVs) to cycle numerous times, and led to the eventual rupture of the pressurizer relief tank's rupture discs. These incidents all provided complications to the transient that could and should have been avoided.

The related violations and problems are described in the enclosed Notice of  ;

Violation and Proposed Imposition of Civil Penalties (Notice). The four most significant violations or problems are described in Section I of the Notice  !

and involve the failure to identify and correct significant conditions adverse to quality at the facility related to spurious steam flow signals and the atmospheric relief valves, both of which led to unnecessary safety injections during the transient; the failure to identify and correct significant conditions adverse to quality at the facility related to providing adequate training, guidance, and procedures for the operators to cope with the event; and the failure by the SNSS and NSS to exercise appropriate command and control of the operations staff and the reactor during the transient.

With respect to items I.A and I.B in the attached Notice, your staff was aware of degraded or impaired equipment performance and tolerated or " worked around" the conditions. Spurious high steam flow signals were known to have occurred during previous reactor trips at the facility. However, the Rather, cause of the prior condition was not identified and appropriately addressed.

assessments focused on the cause being related to a change in the high steam flow setpoint and the expected response of the related instrumentation. As a result, the actual cause (a pressure wave actuating the high steam flow instruments as a result of a turbine trip) was not identified and corrected.

With respect to the second problem, your staff was also aware of the less than optimum performance of the atmospheric dump valve controls as a result of

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j Public Service Electric ,

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! implementing a design modificat' ion in 1977. Consequently, the automatic . f functioning of the valves in response to actual pressure conditions was i i impaired. However, notwithstanding that the FSAR provided for the automatic actuation of these valves, this condition was not corrected, but rather i

tolerated by relying on operator actions to assure the proper operation of the .

i valves. These conditions provided an additional challenge to the operators

during the April 7, 1994, event.

With respect to item I.C in the Notice, the SNSS decided to personally ,

! override a circulator pump protective interlock (vacuum permissive) so as to l l restart a pump that had tripped, and he left the control room to complete this  !

j activity. In doing.so, the SNSS failed to maintain the command function at a .;

critical time during the plant transient. During his absence, operator action ,

1 resulted in reactor coolant temperature decreasing below the minimum .

! temperature for criticality, which eventually resulted in, (1) satisfying part -

i of the protection logic for an automatic safety injection, and (2)

! operating errors during subsequent recovery actions that caused the reactor trip. In addition, while the SNSS was absent from the control room, the NSS, 4 l designated as responsible for the control room command function, assumed the 2 6 duties of a reactor operator for a short period by performing control rod i movements. As a result, for the~ period of time the NSS was manipulating the controls, no individual was responsible for the control room command and j

control function. j With respect to item 1.D in the Notice, four violations were identified.  !

f l Degraded conditions or workarounds were accepted rather than fully addressed.

Procedures and. training were not adequate to assist the operations staff in l j coping with the transient caused by the grass intrusion, including the maximum j

rates for reducing power during transient conditions, recovery and control of 3

the safety injection system, and recovery from a solid pressurizer condition.

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In addition, the SNSS violated a procedure for maintaining configuration l

control for the circulating water pump when he momentarily bypassed an

! interlock for condenser water box vacuum in an attempt to start the pump.

i This procedural concern is significant in light of management's response to l l prior events. Similar grass intrusion events had occurred prior to i April 7, 1994. In response to those prior occurrences, station management

! provided specific procedures and instructions for dealing with the effects of j the events on operation of screenhouse equipment. However, station management j did not provide adequate training or specific guidance and procedures for j activities performed by the control room staff, and this failure was significant since the potential safety effect of the April 7,1994, transient

- was aggravated by control room staff errors. Also, the plant's nuclear steam I system supply vendor (Westinghouse) had reported the potential for inadvertent

! safety injection system actuations filling the pressurizer to solid

conditions. However, station management had not taken actions regarding the Westinghouse information.

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Public Service Electric i 1

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These four problems demonstrate that management provided inadequate direction ,

J to the staff at the facility, which contributed to the operator errors, non- l 1 conservative operational decision-making by line management, and deficiencies  !

in command and control. It also raises serious questions regarding the manner I

in which management's expectations are established and communicated to the  !

Salem staff regarding their performpnce at the station. Furthermore, these l r deficiencies are particularly disturbing since this was the fourth significant

event at the Sales facility since November 1991, when an overspeed condition-caused catastrophic damage to your turbine at Unit 2. The recent history also

raises serious questions regarding the adequacy of management's commitment to

! conducting appropriate root cause analysis of the problems at the Sales

' facility, as well.as improving performance and precluding such events in the future. Given the significant programmatic deficiencies exemplified by these l

violations, each of the problems is individually classified at. Severity Level III in accordance with the " General Statement of Policy and Procedures for NRC 3

Enforcement Actions," 10 CFR Part 2, Appendix C, (Enforcement Policy).

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! The NRC recognizes that subsequent to the identification of these problems, i

actions were taken to correct the problems and prevent recurrence. These

actions, which were described at the enforcement conference, included, but i j were not limited to, (1) convening of a significant event response team to I

review the event and its ramifications; (2) verbal reinforcement that all j staff are expected to identify and participate in the correction of problems

! at the facility; (3) additional training to individuals whose performance was l 1ess than expected, including simulator training sessions for all the 1 operating crews; (4) issuance of an Information Directive to all shift i personnel to reinforce and clarify management's expectations regarding

' command, control, and communications; (5) revision of certain procedures; and (6) development of long-term program improvement actions as part of your Comprehensive Performance Assessment Team (CPAT).

Although the NRC found each of your immediate corrective actions to be acceptable, the NRC is unwilling to predict or assume success for your long-term actions because historically, the implementation of such actions for past problems has proven to be ineffective. While your actions are aimed at proximate causes, the NRC is not yet confident that lasting actions have been established that will prevent recurrence. Specifically, although senior management generally has established appropriate expectations for staff level performance, those expectations were neither clearly communicated to the staff, nor effectively reinforced by middle management and first line supervision. The NRC is particularly concerned because communications between senior station management and the staff have been deficient for some time, and

'the deficiencies have contributed to a number of prior events.

Simply stated, it appears that you have tolerated an atmosphere that accepts degraded conditions and resultant workarounds such as those that contributed 1

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to the event, rather than establish an atmosphere of a quality licensee

environment. Therefore, to emphasize the importance of aggressive management of the Salem facilities to ensure that (1) management expectations are i established, effectively communicated, and followed by your staff, (2) problems are promptly identified and corrected, and (3) operations supervision, particularly the Senior Reactor Operators, maintain appropriate conmiand and control of the reactor at all times, particularly during transient conditions, I have issued the enclosed Notice of Violation and Proposed Imposition of Civil Penalties (Notice) in the cumulative amount of $500,000 for the violations set forth in Section I of the Notice.

The base civil penalty amount for each Severity Level III violation or problem is $50,000. The adjustment factors in the Enforcement Policy were considered and no escalation or mitigation for those factors was considered appropriate.

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The NRC considers items I.A, I.B. and I.D to involve multiple or continuing violations that existed for more than one day. Enforcement discretion was applied to escalate the civil penalty for items I.A. I.8, and I.D an additional $100,000 each,.per Section VII.A of the Enforcement Policy, to further' emphasize the importance that the NRC places on the need for each licensee to identify conditions adverse to quality, determine the root causes, and promptly effect lasting, comprehensive corrective actions rather than tolerating degraded conditions and workarounds.

You are required to respond to this letter and the enclosed Notice and should follow the instructions specified in the enclosed Notice when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. In your response, you should also describe the corrective actions that you have taken or planned to ensure management expectations are clearly communicated to, understood by, and consistently carried,out by, your staff; how you will measure the

- effectiveness of those corrective actions; and how you will reinforce your i expectations. Your response should also address in particular what actions l

you will be taking to promptly terminate this apparent practice of tolerating known degraded conditions such as those that contributed to this event. We l

expect an aggressive and prompt response to this matter as neither you nor the NRC can accept (1) this performance in the future, and (2) the large number of  !

j equipment related events that have recently occurred at Sales. After '

reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements.

In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room.

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.' The responses directed by this letter and the enclosed Notice are not :.ubject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Pub. L. No.96-511.

$ Sincerely, James L. Milhoan l

Deputy Executive Director for i Nuclear Reactor Regulation, Regional j.

Operations and Research j

Enclosure:

! Notice of Violation and Proposed Imposition of Civil Penalties

-cc w/encls:

J. Hagan, Vice President - Operations / General Manager - Salem Operations S. LaBruna, Vice President - Engineering and Plant Betterment C. Schaefer, External Operations - Nuclear, Delmarva Power & Light Company R. Hovey, General Manager - Hope Creek Operations F. Thomson, Manager - Licensing and Regulation J. Robb, Director, Joint Owner Affairs A. Tapert, Program Administrator R. Fryling, Jr., Esquire M. Wetterhahn, Esquire P. Curham, Manager, Joint Generation Department, Atlantic Electric Company i Consumer Advocate, Office of Consumer Advocate l W. Conklin, Public Safety Consultant, Lower Alloways Creek Township ,

K. Abraham, PAO-RI (2)  !

Public Document Room (PDR)

Local Public Document Room (LPDR)

Nuclear Safety Information Center (NSIC)

NRC Resident Inspector State of New Jersey State of Delaware 3 State of Maryland i

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Public Service Electric l and Gas Company l

DISTRIBUTION:

PDR SECY CA l JTaylor, EDO **

JMilhoan, DEDR ,

JLieberman, OE ,

TMartin, RI l Stewis, OGC WRussell, NRR l RZimmerman, NRR Enforcement Coordinators RI, RII, RIII, RIV FIngram,GPA/PA JFitzgerald, 01 PLohaus, SP l DWilliams, OIG )

EJordan, AE00 LTremper, OC JBea11, OE l EA File (2)

DCS JStone, NRR SDembek, NRR VMcCree, OEDO <

MThadani, NRR MShannon, ILPB i

SECY NOTE: In the absence of instructions to the contrary, SECY will notify the staff on Tuesday, October 4,1994 that the Comunission, by negative consent, assents to the action proposed in this paper.

l NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES Public Service Electric and Gas Company Docket Nos. 50-272; 50-311 Salem Nuclear Generating Station License Nos. DPR-70; DPR-75

. Units 1 & 2 EA 94-112 d

As a result of NRC review of the results of an Augmented Inspection Team inspection conducted on April 8-26, 1994, as well as a follow-up inspection conducted on May 1-June 25, 1994, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, the Nuclear  !

- Regulatory Comission proposes to impose civil penalties pursuant to Section 3

234 of the Atomic Energy Act of 1954, as amended (Act), 42 U.S.C. 2282, and 10 l d

CFR 2.205. The particular violations and associated civil penalties are set forth below:

I. Violations Assessed Civil Penalties A. 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, i requires, in part, that licensees promptly identify significant 4

conditions adverse to quality, determine their causes, and take i corrective action to preclude recurrence. l

- Contrary to the above, prior to April 7, 1994, the licensee did  !

not promptly identify and correct the cause of spurious high steam i flow signals that occurred during reactor / turbine trips on l June 10, 1989, July 11, 1993, and February 10, 1994, a significant condition adverse to quality. As a result, this condition i i recurred on April 7, 1994, during a transient, and led to an unnecessary Safety injection system actuation in response to a reactor trip.

This is a Severity Level III problem (Supplement I). (01013)

Civil Penalty - $150,000

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i B. Licensees are required by 10 CFR 50.71(e) to assure that the updated Final Safety Analysis Report (FSAR) accurately describes the design, construction, and operation of the licensed facility and the operations. Specifically, licensees are required (in part) to revise the updated FSAR "to include the effects of... all 4

changes made in the facility or procedures as described in the

FSAR."

Licensees are required to maintain and operate the facility as described in the FSAR unless, per 10 CFR 50.59, changes are made after completing a safety evaluation which provided the bases for the determination that the changes did not involve an unreviewed safety question.

The Salem Generating Station Updated FSAR, in section 10.4.4.1, states that "[s)hould the condenser not be available as a heat sink, the steam generator safety valves and power operated relief

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i valves (PORVs) will open to dump steam to the atmosphere." This

' automatic opening of the steam generator PORVs is credited in section 15.2.9.1.2 in the mitigation of a loss of offsite power ,

and a turbine trip. ,

Contrary to the above, the licensee had modified the controller  ;

i for the steam generator PORVs such that, prior to April 7, 1994, the PORVs would not open to dump steam to the atmosphere following i the loss of the condenser heat sink without manual operator action. The changes were made and the plant so operated without the completion of a safety evaluation and without revising the Updated FSAR. The inability of the steam generator PORVs to open automatically, as described in the FSAR, was recognized and 1-tolerated by the licensee since 1977 and contributed to the

! severity of the April 7,1994, plant transients, including an 3

unnecessary automatic actuation of the Safety Injection system.

R This is a Severity Level III problem (Supplement I). (02013)

! Civil Penalty - $150,000

( i i C. Technical Specification 6.1.2 requires that th6 Senior Nuclear l Shift Supervisor (SNSS) or, during his absence from the control room, a designated individual, shall be responsible for the

control room command function.

j Technical Specification 6.8 requires that written procedures be established, implemented and maintained covering the activities i

referenced in Appendix A of Regulatory Guide 1.33, which includes i

administrative procedures.

PSE&G Administrative Procedure NC.NA-AP.ZZ-002(Q), Attachment 32, l " Shift Management Responsibility for Station Operation," (a 4

! procedure written to satisfy the requirements in Appendix A of

, Regulatory Guide 1.33) requires, in part, that the SNSS shall remain free to survey and analyze all operating parameters, noting 4 that intense involvement in any particular detail may run the risk i

of losing control and perspective of the overall operation.

I Contrary to the above, on April 7, 1994, during a grass intrusion i event at the circulating water system intake structure, shift management personnel did not remain free to survey and analyze all operating parameters, and for a short period lost control and i

perspective of the overall operation, in that:

i 1. The SNSS involved himself in a particular detail when he left the control room during the loss of circulating water transient to override a circulating water pump protective i-1 1

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Notice of Violation interlock so as to restart a circulating water pump, and in doing this activity, failed to maintain his control room command function during a portion of the plant transient; and

2. While the SNSS was absent (rom the control room, the nuclear shift supervisor (NSS), designated as responsible for the control room command function, assumed the duties of a reactor operator by performing control rod movements for a l

.short period. As a result, for the period of time the NSS j was manipulating the controls, no individual was responsible for the control room command function.

. This is a Severity Level III problem (Supplement I). (03013)

Civil Penalty - $50,000 D. 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that licensees promptly identify significant conditions adverse to quality, determine their causes, and take corrective action to preclude recurrence. '

Contrary to the above, on Apri1~7, 1994, inadequate training, guidance and procedures were provided to the operators to cope with plant transients resulting from grass intrusion events, events which had previously occurred frequently at the Salem facility and which had caused safety system challenges and reactor trips, significant conditions adverse to quality, as evidenced by ,

the following violations:

1. Prior to the reactor' trip on April 7, 1994, the SNSS determined that an important action to mitigate the  ;

ongoing plant transients due to the grass intrusion event' was 'to bypass an interlock to allow the quick restart of a station circulating water pump, as had been done on previous occasions. Existing procedures .

did not authorize this action. The SNSS did not use i j

station procedure, NC.NA-AP.ZZ-0013(Q), " Control of Temporary Modifications," when bypassing the interlock, in that he changed a switch position to allow the pump to restart, without satisfying the i requirements in Section 3.9 of the procedure.

Specifically, the SNSS did not approve the bypass of the circulating water pump interlock as a temporary modification, including initiation of attendant logs l and reviews. Further, the SNSS did not discuss the i bypass of the interlock with a System Engineer as required by Section 4.1 of the procedure, nor was a  ;

Temporary Modification Package prepared for the  !

bypassed interlock, as required by Sections 5.1.7 and 5.2 of the procedure. Finally, and because the bypass of the interlock was not identified and. treated by the I

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l Notice of Violation f SNSS as a temporary modification, appropriate reviews

of the modification were not conducted as required by Section 5.4 of the procedure.

L j 2. While experiencing the loss of condenser vacuum, a rapid power reduction was initiated from approximately.

l 75 percent power, using procedure Sl.0P-10.ZZ-0004(Q),

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j. '" Power Operation." The procedure was'not adequate for i

the circumstances in that it did not provide specific guidance to support the rate of power reduction that was used during the April 7, 1994 transient, and it i

did not state the management expectation as to when

! operators should cease the effort to maintain plant operations and instead stabilize plant conditions 'by either a turbine or reactor trip._ Additionally, l

operator training was not adequate to compensate for j the lack of specificity in the procedure. As a result, operators exceeded shutdown rates fncluded in

their training, and allowed the reactor temperature to

! drop below the minimum temperature required for I critical operations.

3. During the initial' actuation of the safety injection' system

on April 7, 1994, an emergency core cooling system partial train actuation occurred, namely, the "A" train of the Solid State Protection System actuated, and operators implemented l

j- emergency operating procedure,.1-EOP-TRIP-1. Neither the procedure nor the operator-training provided was adequate

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for the circumstances in that guidance was not sufficient to

allow operators to recover from the resulting safety

" injection logic disagreement and to control the safety injection system in a timely manner. The plant's nuclear steam system supply vendor (Westinghouse) had previously communicated to the licensee the potential for inadvertent safety injection ~ system actuations filling the pressurizer to solid conditions. However, station management had not taken actions regarding the Westinghouse information. As a result, the plant pressurizer filled with water from the ,

safety injection, significantly complicating the event ,

because it challenged the reactor coolant pressure boundary.

4. During the recovery from the solid pressurizer condition, neither plant procedures nor operator training was adequate in that operators were unable to utiliza any procedure relating to existing plant conditions. Although Emergency Operating Procedure, E0P-CFST-1, " Coolant Inventory Status Tree" Hs available, it was not used because the operators l

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i" pressurizer level, FRCI-1, could not be used because of a requirement to secure all reactor coolant pumps to ensure l that the Reactor Vessel Level Indication System indicated 1

i greater than 100 percent full.

l This is a Severity Level III problem (Supplement I). -(04013)-

! Civil Penalty - $150,000 Violations 'not Assessed a Civil Penalty

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! A. 10 CFR 50.57 requires, in part, that emergency plans and l procedures for event classification and notification of offsite

, authorities be implemented. Salem Emergency Plan and Event t Classification Guide, Attachment 8, NRC Data Sheet, requires that L specified information regarding the event description be l- completed, approved, and provided to the designated communicator i for transmission (to the NRC) within 60 minutes. The specified i information includes systems affected, actuations and their initiating signals, causes, effect of event on plant, and actions taken or planned. It also requires that anything unusual or not understood be noted.

Contrary to the above, on April 7,1994, although the NRC was informed of an unusual event at the facility involving a reactor trip and safety injection, specified information was not communicated to the NRC within 60 minutes, as evidenced by the following examples of information that were not provided:.

1. The apparent logic mismatch of the protection system and j resultant unexpected (or lack of) operation of the (ECCS) )

flow path valves and the unexpected condition of the main  !

steam and feedwater isolation systems; j

2. The cause of the reactor trip;

3. The effect of the event on the plant (namely, the consequent ,

solid RCS condition); and, j

4. The operators' plans to recover from the solid RCS condition.

This is a Severity Level IV violation (Supplement VIII). (05014) j B. 10 CFR Part 50, Appendix B, Criterion VIII, Identification and ,

Control of Materials, Parts, and Components, requires in part, l that measures be established for the identification and control of parts and components. These measures must assure that identification of the item is maintained throughout installation j and prevents the use of incorrect parts.

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_ _ . _ . . . _ _ _ _ . . _ - . ~ _ . _ . _ _ _ . _ _ _ . _ . _ _ . _ _ _ _ . _

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!. Notice of Violation  ;

1 i Contrary to the above, prior to April 7, 1994, measures were not _1 i established for the configuration control of certain parts and

!, components, as evidenced by the following examples:

! 1. During the 1993 Unit 2 outage, power operated relief valve i

! (PORV) internals made of 17-4PH stainless steel (original  !

design material) were installed in valves 2PRI and 2PR2, rather than internals made of 420 stainless steel as

recommended by the vendor, and approved by the licensee as i design change replacement material.

! 2. The post-trip investigation for the April 7, 1994 event, j

identified that the installed summator module for the high e

steam flow setpoint did not have the proper identification.

l Specifically, a regular module was installed instead of a specifically-modified variant. ,

. l I

' This is a Severity Level IV violation (Supplement I). (06014) ,

Pursuant-to the provisions of 10 CFR 2.201, Public Service Electric and Gas I l l l Company (Licensee) is hereby required to submit a written statement or

! explanation to the Director, Office of Enforcement, U.S. Nuclear Regulatory  !

Commission, within 30 days of the date of this Notice of Violation and i

Proposed Imposition of Civil Penalties (Notice). This reply _should be clearly ,

I

marked as a " Reply to a Notice of Violation" and should include for each

alleged violation: (1) admission or denial of the alleged violation, (2) the reasons for the violation if admitted, and if denied, the reasons why, (3) the l corrective steps t6.at have been taken and the results achieved, (4) the

. corrective steps that will be taken to avoid further violations, and (5) the

date when full compliance will be achieved. If an adequate reply is not ,

received within the time,specif'ied in this Notice, an order or a demand for i information may be issued to show cause why the license should not be

! modified, suspended, or revoked or why such other action as may be proper  ;

should not be taken. Consideration may be given to extending the response i l

i time for good cause shown. Under the authority of Section 182 of the Act, 42-

U.S.C. 2232, this response shall be submitted under oath or affirmation.

i Within the same time as provided for the response required above under 10 CFR l

2.201, the Licensee may pay the civil penalties by letter addressed to the i Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, with a check, draft, money order, or electronic transfer payable to the Treasurer of l l

the United States in the cumulative amount of the civil penalties proposed

{

above or may protest imposition of the civil penalties in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S.

tl Nuclear Regulatory Commission. Should the Licensee fail to answer within the time specified, an order imposing the civil penalties will be issued. Should L the Licensee elect to file an answer in accordance with 10 CFR 2.205 a

protesting the civil penalties, in whole or in part, such answer should be

clearly marked as an " Answer to a Notice of Violation" and may

(1) deny the j violations listed in this Notice, in whole or in part, (2) demonstrate

extenuating circumstances. (3) show error in this Notice, or (4) show other i

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i Notice of Violation  !.

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• j reasons why the penalties should not be imposed. In, addition to protesting the civil penalties in whole or in part, such answer'may request remission or

! mitigation of the penalties. l In requesting mitigation of the proposed penalties, the factors addressed in Section V.B of 10 CFR Part 2, Appendix C (1992), should be addressed. Any

written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may j incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g.,  ;

citing page and paragraph numbers) to avoid repetition. The attention of the 4

) Licensee is directed to the other provisions of 10 CFR 2.205, regarding the i procedure for imposing civil penalties.

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Upon failure to pay any civil penalties due which subsequently have been j determined in accordance with the applicable provisions of 10 CFR 2.205, this

matter may be referred to the Attorney General, and the penalties, unless

compromised, remitted, or mitigated, may be collected by civil action pursuant j to Section 234c of the Act, 42 U.S.C. 2282(c).

! The response noted above (Reply to Notice of Violation, letter with payment of

- civil penalties, and Answer to a Notice of Violation) should be addressed to:

Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region I, 475 Allendale

Road, King of Prussia, Pennsylvania 19406 and a copy to the Senior Resident Inspector, Salem Generating Station. ,

i i Dated at Rockville, Maryland l this day of September 1994 t

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p y..m *, UNITED STATES s* j t

NUCLEAll REGULATORY COMMISSION WASHINGTON, D.C. SomeH001 I

~s...../

Docket No. 55-8909 License No. SOP-10582-1 4 EA 94-147 Mr. Michael D. Gwirtz

[HOME ADDRESS DELETED UNDER 2.790) )

j Jear Mr. Gwirtz:

On August 2, 1994, th'e NRC conducted an enforcement conference with you at the Nuclear Training Center in Salem, New Jersey, to discuss your actions during an April 7, 1994, event at Salem Unit 1, The event was described in an Augmented Inspection Team (AIT) Report (No. 50-272/94-80) issued on June 24,

. 1994. Apparent violations identified as a result of the inspection were sent to your employer, Public Service Electric and Gas Company (PSE&G) in a letter dated July 6,1994.

The conference was held to discuss the circumstances associated with your

bypassing of a circulating water pump interlock (not safety-related) in order

to restart the pump. At the time you bypassed the interlock, you were the senior nuclear shift supervisor (SNSS) on duty during which a significant

grass intrusion event was occurring. In addition, by becoming involved in the  !

i bypassing of the interlock, you failed to maintain your command and control  !

function during a portion of the event. Your actions exhibited poor judgement i on your part. Further, although we believe you performed this activity because you felt it was in the best interests of safety to maintain the main condenser available, that activity was not authorized by procedures governing control of temporary modifications.

You are reminded that you hold a license from the United States government which confers upon you the special trust and confidence of the American people in the safe operation of a nuclear facility, and places you in a position where sound judgement is expected, and your performance must be above reproach. This includes, as a Senior Reactor Operator license holder, and in your case, the Senior Nuclear Shift Supervisor, your responsibility for ensuring adherence to all conditions of the license issued to PSE&G, as well as maintaining your command and control f1nction at all times. Your actions on April 7,1994, although they occurred for only a short period, did not meet those expectations, and did not provide an appropriate example for the reactor operators and other senior reactor operator under your supervision.

s Accordingly, I have given serious consideration as to whether specific enforcement action should be taken against you. However, the NRC has decided CERTIFIED MAIL RETURN RECEIPT RE0 VESTED

_ . - . . - -- -- . = . _ - . _ .- - -. . - - -- .

Mr. Michael D. Gwirtz

[ that the principal concern related to this event, and the related violations, was your management's failure to clearly communicate its expectations to staff j regarding expected performance at the Salem station. Accordingly, the NRC is

issuing a Notice of Violation and Proposed Imposition of Civil Penalties on  ;

this date to your employer, Public Service Electric and Gas Company, for the  ;

i violations identified by the NRC related to this event. With respect to your '

license, I have decided, after consultation with the Director, Office of

- Enforcement and the Commission, not to take any enforcement action, in part, t

because you received an oral reprimand from your management, you were provided remedial training, you appeared candid, contrite, and remorseful during your enforcement conference at which you acknowledged your errors and because our prior dealings with you do not suggest a problem with your competence or integrity. However, the NRC notes that any similar occurrence in the future could result in enforcement action against you.

No response to this letter is required. However, if you desire to respond, j please provide that response, in writing within 30 days, to Mr. Glenn Meyer, i

Chief, BWR/PWR Section, NRC Region I.

l . A copy of this letter will be placed in your individual operator license

docket file. Also, in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter will be placed in the NRC Public Document Room with your address deleted.

i Sincerely,  !

i I

James L. Milhoan  :

Deputy Executive Director for l Nuclear Reactor Regulation, Regional l Operations and Research j

! cc:

i J. Hagan, Vice President - Operations / General Manager - Salem Operations Public Document Room (PDR)

~: Local Public Document Room (LPDR)

Nuclear Safety Information Center (NSIC)

NRC Resident Inspector State of New Jersey State of Delaware K. Abraham, PAO-RI (2) e i ,

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1

. - j Mr. Michael D. Gwirtz 1

DISTRIBUTION:

. PDR SECY 1 CA JTaylor, EDO JM11hoan, DEDR JLieberman, OE TMartin, R'l Slesis, OGC WRussell, NRR

RZimmerman, NRR

. Enforcement Coordinators

~,

RI, RII, RIII, RIV FIngram,GPA/PA ,

JFitzgerald, O!  !

PLohaus, SP

. DWilliams, OIG EJordan, AE00

, JBeall, OE

, EA File (2) i

DCS l

! JStone, NRR l SDembek, NRR j

i VMcCree, OEDO CMiller, NRR MShannon, ILPB i

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• l Public Service Electric & Gas Co.

Firewa"ch Investigation .

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Withhold From Pu "c Disclosure In - or cea ith 10 @-2390g

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i EXECUTIVE SUNNARY Publ c Servic ric and Gas Company employs a contractor, l

! , to provide regulat;ory-required fire watch services that compensate for ' impairments of fire protection '

systems at the Sales and Hope Creek Generating Stations. The i

Site Protection Department in the Nuclear Services organization -

provides direction and oversight for this function.

a Field Supervisor for $ observed a firewatch employee fa 1 to properly complete the Hope Creek patrol route #

( rove ).Water The supervisor covered the missed post (Hope Creek l Service Building) and subsequently confronted the 1

{ The employee admitted missing that and other posts and e r j false data on the firewatch log. The employee stated he resigned, was escorted off site and his acce.ss was revoked' The su rvisor immediately notified $nagement of the I

M corporate managesent immediately initiated an investigation

to determine the extent of the problem. M compared a sample of-firewatch rove logs with security door access records, and l followed up with interviews of firewatch employees and supervisors. Based on a one week sample period, the @

! investigation found,17 of 35 firewatch personnel with discrepancies between locations denoted on their logs and security door access records. Those 17 individuals were sus nded Wand their site access was revoked by PSE&G.

reports contained a number of recommendations which were a uated by PSEEG in formulating corrective actions.

PSE&G undertook a comprehensive investigat n and review of this investigation, j matter. The purposes were to assess the examine independently the causes of the problem, including the l

role of $ as well as its own oversight of the firewatch

! functions, determine root causes and assess the appropriateness of corrective actions taken and the necessity for additional i

corrective actions, and examine this issue in conjunction with As part of its other possibly related occurrences at the site.

' investigation, PSE&G Quality Assurance selected two additional d sample periods from earlier in the year and performed an independent analysis of the firewatch records and security door access records. This confirmed that the problem was neither j in the solely recent nor isolated. The PSE&G review resu suspension of two additional firewatch personnel 4

)

In addition, PSE&G Corporate Security Services was asked to J

conduct an independent investigation, using experienced i

investigative personnel with law enforcement backgrounds, into the apparent falsification of rove logs by firewatch personnel.

Their investigation identified a number of issues related to the missed roves and falsified logs. These issues were considered in the root cause analysis and in defining corrective actions. The j

investigation also identified several peripheral issues which are

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! b The result of ti investigation

! being followed up separately.

i was that there was a programmatic breakdown in the firewatch' process. However, no relationship was identified between the results of this investigation and previous firewatch related allegations.

• n e events and issues identified by the investigative processes i

j were evaluated in accordance with the documented facts from three {

l perspectives: people, plant and procedures. Me root cause j categories and causal factors of the program breakdown were derived using the following analytical tools: causal factor analysis, fault tree, and hazard / barrier / target analysis. He root causes determined from the root cause methodology were .

compared with the root cause categories originally identified by l PSE&G management. W e three root cause categories and causal i l i

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factors are:

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j 1. Imssthanadequatesupervisionoffirewatchpersonnel$

i 6 j a. Failure to consider the limitations of human performance, especially routine repetitive tasks, in l

planning rgve routes and work assignments. .

l b. Less than adequate field supervision, minimal field 4 contact.

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! c. Practice of covering emergent work with field .

! supervisors.

! d. Program Manager performance: ignored feedback, I i

favoritism, economic pressure, no clear expectation 1 articulated.

t 1 l 2. IAss than adequate performance of firewatch personnel. l

a. Poor process for training and indoctrination in ,

j expectations of new employees. l l

f b. Expectations communicated only once and not reinforced. j

c. Low self-esteen. j

3. Less than adequate oversight by PSE&G. l 1

1 a .- Line management resource constrained.

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! b. QA audits and surveillances not structured to be able to detect the problem.

l f c. Assignment of emergent work to $

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! A detailed discussion of the causal factors and associated 4 corrective actions is included in the main body of the report. ,

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A number of short term corrective actions, such as rove route

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redesign, were taken to assure the integrity of the firewatch program. In addition, a number of broader actions have been or

' are being taken to prevent recurrence of similar instances.

significant long term corrective actions are as follows:' ~

1. corporate staff will develop a standardized firewatch ndoctrination training program to be given for all current and new employees. This training will provide consistency of performance expectations and cover the basic duties of and methodologies for accomplishing firewatch responsibilities. The training will be provided by @

management and will not be delegated to supervision on the l

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job. Training for current employees is scheduled for Refresher training to be completion by November 30, 1992.

conducted annually.

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2. W corporate staff will conduct supervisort training structured toward supervisor / employee relation skills, quality post inspections, communications, and quality Refresher training to be (Completed 9/3/92) circles.

I conducted annually.

3. (llbcorporatequalityassuranceandsafetypersonnelwill conduct after hour inspections on a quarterly basis.

(Commencing 4th quarter 1992) j

4. R.tndom card histories of at least one firewatch person assigned to each shift will be conducted by ggErsite management on a weekly basis. (currently implemented)

5. Random card histories of at least three firewatch personnel will be conducted by,3EE corporate staff personnel on a 3

' quarterly basis. (Commencing 3rd quarter 1992)

) 6. The Rove Challenge Program will be utilized on at least a i once a week basis for all shifts. (currently implemented)

7. Commentary will be added to General Employee Training j

stressing the importance of the firewatch program.

(Completed 9/21/92) j 8. The telephone system will be used for call-ins from isolated l areas. (Currently implemented)

A communications policy letter has been issued. Firewatch 9.

are encouraged to suggest human factors rove enhancements to j

supervision for possible implementation.

l (Completed 9/18/92) with and oversight PSE&G is strengthening the relationsh l 10.

ofdgEB5 This includes: (1) moving supervision into the same facility with PSE&G's fire p tection supervision, l thereby enhancing face to face communication; (2) l

-III-

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! rainforcing the expectations of 6 accountability for j implementation of a successful firewatch program; (3)

' reinforcing PSE&G's Fire Protection supervision and i

management responsibility for ultimate performance of the l l

firewatch program; and (4) Flias committed to reduce their Program Manager's administrative duties permitting him to focus on firewatch performance.

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l 11. During expected periods of emergent work (i.e., outages)

I additional firewatch will be added. (currently implemented) i, i 12. Salem and Hope creek Quality Assurance will include ,

l-verification performance o'f firewatch roves in~the i appropriate surveillance checklists. (currently implemented) i l

13. -The Quality Assurance Audit Plan for Fire Protection will be 4 revised to include field verification of firewatch roves and

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a sample comparison of rove logs with security door transaction records. (scheduled completion 12/31/92) j! ~

As noted in the list of corrective actions, in order to fully f close out this issue, there are a number of actions to be completed. In addition, certain of the findings from the j

firewatch investigation are being evaluated in conjunction wi'th -

the results of another investigation. PSEEG has determined.that these items may be pursued in an orderly manner without

negatively affecting the ability to provide adequate firewatch

! services. Upon completion of the implementation of corrective j actions, and the development of new initiatives and the i

modifications made to ongoing programs, PSE&G is confident that l

the potential for recurrence will be minimized and its ability to i detect similar deficiencies will be significantly enhanced.

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EDO Principal Correspondence Contr 1 FROM: DUE: / / EDO CONTROL: 001D374 DOC DT: 08/08/9 FINAL REPLY:

TO:

Chairman Selin FOR SIGNATURE OF : *

• GRN **

CRC NO: 94-081 DESC:

ROUT .

NRC VIOLATIONS'AT SALEM NUCLEAR PLAN Taylor Milhoan

.v. . . . ..., :.,, c. ,. s .,

. .. . a,_ < . .

, ,., Thompson Blaha

i ....;,- , ; c. . . z.at.;H

/.c t, 4.fr.. /. . . s Y TTMartin, RI DATE: 08/16/94 Lieberman, f0lA- ' N .._._ _ --.___.

ASSIGNED TO: CONTACT:

NRR Russell p

S,. - - - , - OR _ :

FOR APPROPRIATE ACTION 9

OFFItE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET

'/ PAPER NUMBER: CRC-94-0812 IDGGING DATE: Aug 15 94

! ACTION OFFICE: E l AUTHOR: h l AFFILIATION:

CHAIRMAN S \

ADDRESSEE: j i  !

LETTER DATE: Aug 8 04 FILE CODE: IDR-5 SALEM #

}

SUBJECT:

NRC VIOLATIONS AT SALEM NUCLEAR PLANT 4

1

! ACTION: Appropriate DISTRIBUTION:

SPECIAL HANDLING: NONE f CONSTITUENT:

i l NOTES: ENCLS To: EPO DATE DUE:

SIGNATURE: . DATE SIGNED:

AFFILIATION:

e 9

EDO - 010374

i

) o y TO: HonorableMrIvan Selin, Chairman NuclearRegulatory Commission

, Washington, DC 20555 FROM: '

1

) DATE: August 8,1994 J

I

SUBJECT:

NRC Violations at Salem Nuclear Plant '

Your closing remarks with respect to the Wag with the PSMG management have been misunderstood by the PSMG management and have resulted in the n=+:-:e-sy termination ofperfectly fine employees The supervisors and managers have used this opportunity to get rid ofemployees who are not willing to conform 100 percent with the i

' supervisor. Also an employee who constantly points out Tech. Spec. and license violation j

is target ofsupervisors' wrath IfPSMG management does not exercise some review of supervisors' aptions, it will be very ddficult for employees to raise nuclear safety issues j

within PSMG It is within the NRC' regulatory responsibihty to ensure nuclear industry employees are not intimidated and reinhated for raising nuclear safety issues as normal i work responsibilities.

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i l 1 The managemem has acted with haste and without due proc 6ss in violation ofseveral

! Federal and State statutes Management review not  !

b i $'

{ t must operate

'; without violating i NRC rules and regulatens andliomse

( 'ons. It is the intent of this submittal to prove I

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!. that any further operation by this licensee of Salem Nuclear Plant is a violatisn of NRC j rules , regulations and conditions oflicense

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1. April 7,1994 event at Salem -1 has caused substantial anxiety in the general public i around the plant . He river grass problem has been there for ever and will continue to be there in future. Salem UFSAR, Chapter 15 does address consequences ofloss of l

offsite power but the probabihty of all circulators Gryk.g due to river grass problem is j much higher. Therefore any further operation of Salem Nuclear Plant is an Umeviewed i Safety Question (USQ )in acecch with 10CFR 50.59 process i

Constmetion ofcooling tower as proposed by State of New Jasey was stonewallb by

! the PSE&G management and significant false I b-rdion as to the cost ofthe cooling j tower was presented to the State ofNew Jersey. With the substantial resources spent

{ on the defeating of the State ofNew Jersey requirement regarding building of the

! cooling to , PSE&G management must have total for 7

} at Salem-1 a result ofPSE&G amant commitment to gf 2

2 Derefore, Salem management is totaHy responsible for mot buBding the cooling i i tower and resmiting April 7 accident. Amy Aring of the low level employees by the management is unwarranted and should be opposed the NRC.

1 l Salem Plant should not be allowed to run until USQ is addressed by the management l and approved by the NRC if cooling tower building is not committed by PSE&G. Even i ifPSE&G commits to building cooling tower, a Justi6 cation for CAM Operation l (JCO) by the Licensee is required by NRC regulations.

2. PSE&G management committed violation of10CFR 50.5 on or about April 21,1994.

Review of a Potential Issue ( PI ) as reported by Westinghouse (At+h 1 )

regarding non-conservative calculation of set point for Pressurizer Overpressure l

, Protection System (POPS) indicated Salern-1 was in violation of the Technical i SpaA% tion and its basis which are naaMaan oflicense An incident (IR),

j gd was written by the * (Ken Ogara g . . .

i fh i

This IR stated that Salem was outside its design basis as stated in Salern UFSAR and its Tech. Specs. Salem need not shutdown h= additional overpressure protection

' is provided by the safety valve RH-3, installed on RHR pump common suction line.

j. This valve does provide overpressure protection but its use in the analysis has been

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! h NRC until Salem makes changes to its Tech. Specs. Salem IJcensing bas not submitted either relief reguests or changes to the Tech. Specs.

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evensb6 ugh I had submitted the formal request several nwwehe back.

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~ Dey da==aad options to courseofactions. Itwas p -

damaaad that ifIR i(delivered to the control room , the NRC resident and the NRC j AIT team would get the copy of R next day and Salem would not be able to restart l as planned. This could cost PSE&O several miBion DoBars and may even shutdown l Salem-2. If however, the NRC discovers the R ister and management actions in i wkhm thesinformation from the NRC, there could be's 50-100K fine which is

! signi6cantly less than the loss to the PSB&O if the NRC delays the startup. NRC and

! media attention was on these same valves which ihnetion as Power Operated Valves ( PORV ) at Power and as POPS at,. cold temperature than 31 PSEAG management was able to receh e the NEC perumlasian to start by IEegaEy

g-withholding sigalScast inforunaties by the licensee and its sealer persommeL his is a violation of 10 CFR 50.5 and must be investigated by Federal and State agencies with fab resources.

3. De POPS issue as stated above was reported to PSE&G by Westinghouse by thcar Letter PSE-93-204 (Attar 4=nant 1) received by PSE&O in March 1993. His issue was reviewed by Vice President S LaBruna ( a Director of PT.E&G ) when he read the Tech.

Spec. revision by Diablo Canyon as reported in industry publication (Attachnwww 4).

Several other plants have reported this problem and the NRC has issued an NRC Information Notice, IN- 93-58 ( Attar 4=nant 5 ) dated July 26,1993 based on these reports. The NRC should have issued this as a Bulletin or Generic letter requiring i inaaaaan to justify fbrther operation and take timely action on a issue which can result in brittle rupture of the Remotor Vessel .

E&PB reported to the Salan station in December 93 ( Aetne4=nant I A )that the as-calaulated pressure in RCS artmadad the eBowable pressure at low RCS temperature as

! provided in Saleen Tech. Speca. Hey went ahead to use a ASME Code Case to state t there was no problem. EAPB management known fuDy wee that a Code Case or later edition of ASME Code can not be used by licensees unless reviewed and approved by l Code j the NRC for a specific plant. The NRC staff has serious reservatious

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case and has approved its use rebwantly at few other ,h 1 3 I I

)

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.- appewal of the Code Case and gets Tech. Specs and

.;. . is operating ostelde its design basis.

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'  ?- reganhng Incident Reports and definitely in cover-up l . ij. ytalmiana nere is constant conflict between the obli dos of a staffengineer the viala+iaan in accordance with the regulations and as required by law and j .

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!.g ibattion of marimite the factor of b

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)f la resuked in a very long shutdown of both U " dp l

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In conclaston, the PSEAG anamagessent has wBfmEy and BiegaEy acted agalast those who identify the nuclear safety concerns or the NRC violations.

In June 94, E&PB reported final results ( Awh 1B ) of their calculations with weak and non-consavative assumptions that as-calculated pressure will exceed the Tech. Spec. by 0.5 PSIG. These assumptions are different than the assumptions in j { Wenghnuse Letter and are being made so that the problem at Salem can disappear 4

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i l 9/Ithout alicense change 1 (

l No violation has been swgs.d or incident report written because E&PB thinks j they have license to violate Tech. Spec. by 0.5 PSIG. It is my understanding that the j NRC does not have any provision for violations by the Licensees by a small factor.

l Even today, Salem-1 is operating outside its design basis as stated in Salem UFSAR and Tech. Specs and requires inunadiate licensee and NRC action..

l If Sales plast-specific calculation indicates non-eenservatism in the set pollet of i POPS valves, them it is reportable to the NRC within 5 days under 10CFR21.

! Salem Units have been outside their old pressure and temperature limits for last 10-12

! years. The NRC can plot all old limits and locate the currently calculated peak pressure j as required to be calculated by the Westinghouse letter or simply ask PSE&G to i provideit .  !

l If Salem-1 had cooled down hvely in April 7 accident due to safety iqiection and I blowdown of PORVs and Pressurira Safety Valve, then there was a good possibility l of a Reactor Vessel rupture due to' cold overpressurization his may and may not I have been realized either by the NRC or the Ikensee

('" 4A. Salem bas found a problem in their stress calculation which is resulting in redesign i of the existing hangers and supports. Dere are hundreds ofsupports involved on each l Unit. All these supports will be moddied on their next refueling outages There is no l NRC approved JCO which will allow their present operation. In similar situation found at Stone and Webster in 1977,4 or 5 plants were shutdown until required wwwlineadons were made NRC must investigate who in management has been involved is this coversey.

4B In 1991, Salem determined that the support loads exceeded the ASME code eBowables when Pressurizer loop seals and PORV loop seals discharged during an overpressurization event . A design change drained the loop seals and moddied the supports on both Units during the refueling outages No incident report (IR) ,IER or a 3C0 for running the plant until the end of fuel cycle was prepared by EAPB. A week FM Evaluation was to the the NRC' '

His issue was did . M. Morroni John that there was no to hh it Salen Project Manager at the NRC was aware of the Salem problem his informal arranganent may not meet resulatory requirements for meeting the ASME Code whichis statelawin New Jersey.

ASME code vietation is a Federal 10CFR regulation violation and state of New Jersey laws.

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( 5.During startup aAer April 7 accident in Mode 3 , an inspection on presmirizer safety valve supports was made by EAPB. At about 8 PM , Control Room was notified that a pin was missing its clips and its pin was partially out. This support could not have functioned during an carthquake since the pin would have fallen out. Control room requested determination oroperabihty ofthis component at about at 10.00 PM. At about midnight, E@ieg called the support operable since clips are non-safety related con = unable itesn. The appropriate Tech. Spec. Action State (TSAS) was not entered which has only 1 e

9 K'[j PM the clips were installed and job was NRC can dannnantation of this probleen from Technical Department files on NON-DRs and outage shiA managers' log. EAPB should have evaluated not whether the clips are NSR or consumable item but its contribution in j

disabling the Safety Related component in caraying out its safety function.  !

This was a Tech. Spec. violation by Engineering and Operations  !

6.Approximaiely in 1990, a crack was found in one 8" cantainemt spray pipe , close to containment in peru.iion area. 'Ihis piping was cut and replaced by new spool piece l' under a design change prepared by me. The 50.59 safety evaluation prepared by me required a license an=admaat to change the Tech. Spec. on containment isolation since r

the piping being replaced was part ofcontainment Another Tech. Spec. reliefwas the NRC n-1 j action statement from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to I g

'sissue willbe seu management meetinginstead ofin the RC as is case with all 50.59 Safety Evaluations 'Ibe SORC --:-Ag approvmg this design i

l change was attended by the NRC resident Kathy Gibson but no mention was made of the Tech. Spec. violation. Ahhough a blank was installed inside the containment on this line but it was not tested for leakage at containment design pressure. Without this testing, the blank could not become new containment pressure boundary.

The signi&*w of above Tech. Spec. violation can be understood from the Salem SALP report for that year. During that year, an NRC graded drill used this scenario.

This was cited in the SALP report as the basis for Emergency Preparedness Dep-i-era receiving a grade 2 instead of1. Nobody noti 6ed the NRC that actual job at power was carried out in violation of the Tech. Spec.., the concern expressed in the NRC SALP report Contrary to the regoirements in the Tech. Spec. on costalament isolation, the licensee carried est the modifications risking the benith and safety of general public.

7 Similar Design Change Package (DCP ) was prepared, as contingency, by the E&PB .

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l Thisissue was brought usi engineer / preparer of the DCP in PRE-SORC l

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( meeting which was attended by top B&PB managers including the then-Ucensing j Manager. A comnwat was made that it was an interesting problem and let Salem j Station handle it as it deemed aaaammary.

l E&PB high level management including the thaine Manager made a decision to {

ignore the violation of the Tech. Spec. ( the lh- ). l i

I 8 Reactor Head Vent Valve Position indication in control room is designed Safety j Related and EQ. The limit switches are classiSed in the MMIS systeen as Non-Safety i

Related. His may have resulted in wrong type ofrecordkeeping for work orders and

purchasing activities.

9 During early July 1994, one RX bead vent valve did not fully open on demand. The j flow through this was approximately 1/10 offlow through other three valves Valve

! vendor investigated this anomaly and root cause of the sticky stem was found to be the l boric acid caking in the pilot area of the valve. His root cause was issued by the I&C i Engineer This root cause has identified a common mode failure of all trains of the j flow path. Buch a failure can result in unavailabihty of any vent path after LOCA for j venting the hydrogen and other nop- condensable gases Also, Salem has failed to l identify this as an Incident Report even though this event resulted in identiScation of a i common mode failure affecting all r+= hat flow paths

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j Salem Station has ignored the identification of a common mode failure which can l result in complete loss of safety system function.

i 10 Salem Power Operated Relief Valves ( PORV ) in Reactor Coolant System which

) were investigated by the AIT team have inadequate flow area as reported by the valve l vendor. By Tech. Spec. this valve should have a flow area of 3.14 square inches while l vendor letter indicates 1.716 sq. in. as per At'#raaat 3. His results in either a Tech.

l Spec. violation or work around by the operators. The operators may or may not

! remember informal work around in an emergency such as April 7,1994 event ne

, Ucensing Manager and Technical Department

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i beesulone so

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i Contrary to requirements to revise Tech. Spec. la a thnely mammer, Kal- has i shown clear disregard for complying the Tech. Specs.

11 Tech. Spec. 3.4.9.3 requires provision of 3.14 square inches vent area when one or ,

i both POPS are not operable. If the POPS valves have failed say due to air operated 1

diaphragm problems, the same valve could not be operated as vent , specially in Mode  !

4 when RCS is still at pressure.

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His is o Generie Problem at aR PWRs k 12 PWR Tech. Specs provide overpressure protection at power by Presanizer Safety Valve, set at 2435 PSIG. PORVs ahhough set at 2335 PSIO have very small area j such that most plants do not take credit for the operation of PORVs in their i' overpressure protection analysis They also have POPS valves or RHR suction safety valve set at 375 PSIO. There is no protection at intennodiste RCS pressure of 400 to 2300 PSIG. Pressure and Temperature limits in Tech. Spec. provide limidag j condition ofoperation at all temperatures For example, at 360F, RCS pressure may l . be limited to say 1600 PSIG. A mass or heat addition transient will result in ROS <

pressure of 2335 or 2435 PSIG. 'fhis overpressurization may resuk in a brittle rupture of the reactor vessel. 3

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This is a Generic safety problem for a5 PWRs and aR plaats should be saade to l address it. ,

t 13 Reactor head leakage detection system , MCSLAPAM , has not worked at Salem since it was installed in response to the leakage at Salan and as a commitment to the i GL-88-04,. NRC has issued a closure of the PS E&G respense based on this 1 commitment and understanding that Salem leak detection systeen is working and is being malatalaad by the Licensee' In fact the NRC and industry are dap= ding on

] carly detection of the Alloy 600 CRDM tube cracidag ruth which is being reviewed by the NRC. Salem has decided to scrap the systeen without informing the j

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NRC. The Licensing Fagiaa~ feels there is no commitment to the NRC to inform even when the original iMa11=+ ion was made under JCO before startup i

l The significance of early leak detection can be compM by recollecting the loss ofmetal at Salem in 1988 which approximated 0.4 inch per month Reactor vessel  ;

integrity is utmost important since ahnost all plant safety analysis do not address the i reactor vessel failure. The NRC should mandate installation and proper operation of

the leak detection equipment on the reactor head with appropriate Tcch. Spec.

revisions otherwise E-aa~ will continue to violate nuclear safety in favor of

capacity factor.

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PSE&G is not la compliasee with a writtes comanitment made to detect borie acid leak on the reactor head. .

14 When ' Fishmouth' problem was detected in Salesn s'.eam generators, Salem made vabal commitment to remove explosive plugs next outage in a teleoon with Mr. Emit Murphy ofNRR. This onmmitment was not kept by Salem nor did Salam make any effbrts to notify Mr. Murphy when removal of explosive plugs was postponad to next outage.

15.When' Fishmouth' problem detected at Calem, a 50.59 Saf Evaluati

( on was ' appropriate personneliW 'ag ice esident &

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em 1 ( safety evaluation has been asked . ' Murphy, N' l was and revised completely after SORC presentation. This is violation of administrative section of Salem Tech. Specs.

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l 16 PSEAG delays in reporting the Saety & License violations nonna!!y to next day .

i morning so that appropriate relief can be obtained frorn the NRC. This practice is a j willful violation ofthe NRC regulations on the reporting recpa ma in 10 CFR.

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l 17 Salem has been violating accident analysis assumption which is the basis of Tech.

j Spec. on Controlled I4akage. All other plants have been surveilling the seal irsection j

Sow while Salem aantianan to record the # 1 ses! lenkoffflow in this sury.in.w j The acceptance criteria for this surveillmane is 40 GPM which would imply that # 1 j eenl leskoff flow of 10 GPM for each RCP is acceptable. This is not true since

Wantiagk=m RCP manual requires RCP shutdown at # 1 sealleskoff flow of 6 j OPM. Wo. tid.suse RCP manual also provides range of 6 to 13 GPM for seal injection , flow. The intent of the Tech. Spec. limit and sury.itimane is to prevent abnormal plant operation in which all 4 RCPs may have limiting flow of 13 OPM.

l Salem ECCS analysis for mininn!m safety injection flow namunes 78 GPM flow in ,

1 j seal injection branch if normal seal irdection does not exceed 40 GPM. If seal injection j flow in a LOCA event exceeds 78 GPM, then minimum seal injection Sow will not be dehvered to the core.

{

J l Therefore, Salem operation in the past with seal injection flow exceeding the 40 GPM limit is a very sesions violation.  !

l This problem has not been corrected for last 3 years The Lhaiao Fagiaaer who is

! a SRO licensed by the NRC, has been sitting on it so that it may not cause plant

! shutdown. While working on submittal to the NRC he is intentionally delsying

! ivyw&ig it to the NRC so that Salem may not have to shutdown ifPSE&G decides j to acknowledge the existence of the problem This is violation of the license C h 18 There is s incorrect statement in Tech. Spec. on Cantainawed Spray System. This l

i error has been reported to the Lhaiao F=iaaar. No action has been taken by the

! I h_aiao Fani-who also is alicensed SRO.

i i 19 Salem has been performing the surveillance on temperature monitoring inside the j containment differently than stated in the Tech. Spec. from startup days. The

( deviation may not be conservative and there is a potential for Tech. Spec. violation.

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l Licensing Fagine who is a licensed SRO has not submitted the License Change l

( Request even though all required input has been provided to the Licensing Department.

l I This is a Salem License violation and must be corrected immediately.

20 Containment Spray header inside the containment requires filled to mininunn Tech.

Spec. level of the Reft-ling Water Storage Tank (RWST). This beeder has been l found to be drained before. The signi6cance of this item is that it may take longer to I initiate the spray in case of a LOCA or main steam line break. In that case, containment design pressure may be exceeded since calculated peak containment l

i pressure is so close to the design pressure at Salem. ,

! I i 21 As-found set point deviation from Tech. Spec. limit ofPressurizer Safety Valves has l . not been reported to the NRC while other li<wan= have reported such de6ciency.

Y j team prevented this startup and was correct in opening the Pressurizer Safety Valves j and PORVs. This was clear indidation of how Salem management cares about health

! and safety of the general public. PSE&G needs to decide whether they want to run l the Salem facihty for the general good of the public or for making profit for the

{ shareholders.

23 Salem management uses filthy language in workplace. Although use of such language among employees may be an internal matter for PSE&G to address, use of such language for its regulators is una"*ptable. In this incident, this supervisor who has

, been relicensed by the NRC, used foul language and body motions for an NRC lady inspector who was at Salem to conduct an audit of the activities related to the closure

, of Generic Letter 88-17 and midloop operations. This licensee needs to cleanup foul l language usein the normalworkplace

! In conclusion, public laterd wm be served if Salem is shutdown ladefinitely, l ante sigalficant changes are made at the top and middle managessent. It must

] have its regulators' confidence la operating the facility within the law and j licence. He NRC has been compassionate la the past la dealing with this

! licensee but if it does not act quicidy and pacipitately la shutting down this j facElty, the NRC wB loose the confidence of the general public.

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( C: - Honorable Senator Joseph R Biden, Jr' 6209 FederalBuilding 844 King Street WI"iayon, DE 19801 oinas % Adtninistrator  !

Region 1.USNRC King ofPrussia, Pa.

- Charles W Senior Resident Inspect U S Nuclear Regulatory Conunission )

kan Generating Station f Hancocks Bridge,NJ 08038

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y Westinghouse Energy Systems Electric Corporation a.. m 3

mee<eP m f m m wan March 15.1993 PSE 93 204 Mr. Dase Perkins. Manager Quality Auurance and Engineering Procurement Public Service Electric and 04< Company P. O. Box 236 Hanweks Bridge. New Jerse) 08038 l l

Public Senice Electrie and Cat Company Salem Urdts No I and 2 Cold Otemreuure Mitiration Setem (COMS) Noncon<ervatism

Dear Mr. Perkin<:

71w (ubject Nu6fe.tr Safety Advicory Lener is forwarded for your information and use. Thit letter i

provides the Wesunghouw e<melu<lon regarding 10CFR21 reportability, plant applicabilit sigruficanee and recommended actions.

Wettinghou<e 10CFR21, h.Ls determined that thi< issue is not reportable purtuant to the requirement < of if you hase any quesunnt corwerning thit Adsisory Letter, plea <e let me know.

l

. Very truly yours, '

M L R. Gasperini, Manager Regional Sales Support l

I Attachment two

,ff,

. PSE-M 2td March 15,1993 g

ce: J. A. Nichok (M/C S 24) !L I A D. A. Peruru (M/C N 14)- IL, I A T

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I THl315 A NOTIFICATION OF A ILECENTLY IDENTIFIED POTENTIAL SAFETY 1&$VE PERTAIN!NO TO SASIC i

COMPONENTS SUPPL!!D BY WESTINCHOUSE. THis INFOILMATION 15 DEINC PROVIDED TO YOU so THAT A I RIVIEW OF THis 11sCE CAN BE CONDUCTED BY YOU TO DETER.MINE IF ANY ACTION 12 REQUIRED. I 4

I F.O. Ses 355. Feehurgh. FA 132XH035 i

Subject:

Cold Overpressure Mitigation System (COMS) Nonconservatism Number: NSAL 93 005B Basic Component: COMS Setpoint Date: 03 10 93 i Plants: See attached 2

Substantial Safety Huard or Failure to Comply Pursuant to 10 CFR 21.21(a) Yes O No It i

Transfer of Information Pursuant to 10 CFR 21.21(b) Yes O Advisory Information Pursuant to 10 CFR 21.21(c)(2) Yes O

References:

St)MM AR Y  !

! (

The purpose of this tener is to inform licensees that Westinghouse has identified a potemial issue  !

i regarding a nonconservatism in the Cold Overpressure Mitigation System (COMS) serpoint development.  !

. i I

l Adateut informemen, if equired, may be ebsauwd from dw onsnator. Telephees a12 3744557.

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4 Originator (s):

I,d R. Hsdwick H. A. 'Sepp; Mander Strategic Licensing Issues Strategic Licensing Issues 4

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i 1 - Sheet 1 of 5 f

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PLANT APPLICABILITY ,

i NSAL.93 005A (Westinghouse COMS Setpoint Analysis) 4

^

( D. C. Cook Units 1 & 2 Turkey Point Units 3 & 4  !

Byron Units 1 & 2 Braidwood Units 1 & 2 i' i V. C. Summer ' South Texas Units I & 2 l Zion Uniu 1 & 2 Almaraz Units 1 & 2 i 4

R. E. Ginna Beaver Valley Units I & 2

Vogtle Uniu ! & 2 Korea Units 3 & 4  ;

! Yongswang Units 1 & 2 Doel 4

• i Millstone 3 Vandellos 2 l

Diablo Canyon Uniu I & 2 Trojan Napot Point Wolf Creek i Callaway Besnau Unlis 1 & 2

- Comanche Peak Units 1 & 2 Sequoyah Units 1 & 2 j Manshaan Units I & 2 i

Asco Units 1 & 2 t Sirewell 8 Watts Bar Units I & 2

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Tihange 3 Jose Cabrera (Zorita)

NSAL 93 005B (Non Westinghouse C051S Seapoint Analysis)

J. M. Farley Units I & 2 H. B. Robinson 2 Shearon Harris !  : Haddam Neck W. 8 McGuire Units 1 & 2 Catawba Units 1 & 2 Indian Point Unlu 2 & 3 . Seabrook 1 ,

Prairie Island Units 1 & 2 t Millstone 2 Qalem Units I & 2)

San Onofre i Surry Units 1 & 2 North Anna Units 1 & 2 Point Beach Uniu ! & 2 Kewaunee Yankee Rowe Arkansas Nuclear I & 2 Davis Besse ! Calvert Cliffs I & 2 -

Oconee 1,2, & 3 Rancho Seco Fort Calhoun i Palisades Palo Verde 1. 2 & 3 Waterford 3 Crystal River 3 St.1.ucie ! & 2

'three Mile Island i Maine Yankee Angra D. Reis i Doel I & 2 Trmo -

Krsko Ringhals 2, 3, & 4 Tihange 1 C. N. des Ardennes C. N. BR3 Kori I & 2 Kyushu Shikoku -

Mihama I & 2 Ohi I & 2 Takahama 1 Pilgrim i Brunswick I & 2 Perry 1 Dresden 2 & 3 ,

Quad Cities ! & 2 LaSalle I & 2 Fermi 2 Hasch I & 2 Oyster Creek River Band 1 Clinton !  !

I i aw =

Sheet 2 of 5 l

J Arno!d 1 FinParick Coopa Nine Mile PL 1 & 2 j Millstone 1 Montecello j I. Susquehanna i & 2 Limerick I & 2 l Peach Bonom 2 & 3 uMopeCreek13 i

Grand Gulf i Browns Ferry 1. 2, & 3 Vermont Yankee WNP2 i Chinshan I & 2 Kuosheng 1 & 2

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j ISSUE DESCRIPTION

! Wutinghouse has identified a potential issue regarding a nonconservatism in the Cold Overpreuure

I hiitigation System (COMS) setpoint development. De pressure difference from the wide range prusure

! transmine c. the reactor vessel (where the Appendix 0 Limit is defined) had not been considered in the I Westinghouse analyses. His pressure difference effectively results in the pressure in the reactor vesael being greater than that seen by the wide range pressure transminers (used to actuate the PORVs when in se cold overpressure mode), and therefore, potentially resulting in violation of the Appendix G Limit.

i j As a result, se fo!!owing nonconservatisms were identified:

i

!  !) For the design basis mass irg}ection transient under normal full flow circumstances, the pressure at the reactor vessel core midplane elevation (used for Appendix G pressure limit)is anticipaud

! to be ten than 100 psi above that measured at the wide range pressure instrument.

j 2) ne design basis heat input transient for COMS is defined as the plant is initially operating with i . a!! reactor coolant pumps (RCPs) off and is cooling down on the R mdual Heat Removal System (RHRS). His crestes a secondary to primary temperature differr a. One RCP is then started, l

which results in reverse heat transfer from the steam generato

;ito the RCS and causes a i pressurization transient. If the RCP is started in the same loop that the wide range pressure i transmitter is used to control one of the PORVs, the transminer will be seeing s lower pressure j than se reactor vessel by approximately 25 psi.

1 i TECHNICAL EVALUATION ,

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Nuclear pow er plants designed by Westignhouse address a plant safety criteria for maintenance of reactor I

vessel integrity by enabling a cold overpressure mitigation system at relatively cold RCS temperatures.

( ne reactor vessel integrity limits for which this system provides protection are defined by the Appendix G curves generated for each plant. To compensate for the identified nonconservatism in the Westinghouse cold overpressure protection analysis Westinghouse has calculated the AP associated with the elevation difference of the wide range pressure transmitter and the mid plane of the reactor vessel.

De .iP values were calculated based upon a conservative set of assumptions for possible plant conditions.

nese assumptions include RCS temperature at 70*F. no pressurizer steam bubble and all reactor coolant pumps in operation. His information is provided so that you may make se determination of applicability and resolution of this issue relative to your cold overpressure protection analsysis basis.

De additional pressure drop, if applicable, is equivalent to the increase in pressure at the critical location for reactor ves:el integrity of the core midplane as opposed to the location of the pressure transmitter shich controls operation of the relief valves during a cold overpressure event. his is the pressure increase which needs to be compensated.

SAFETY SIGNIFICANCE Nonconservative cold ove@ressurs protection setpoints have the potential to result in violation of the l Appendia C semperature/ pressure limits should a cold overpressure event occur. Exceeding the Appendix C criteria requires additional evaluation in order to confirm continued structural integrity of the reactor vessel. Until such time as you can determine applicability of this issue to your plant, reasonable assurance of safe plant operation is provided by the results of post event stress analyses performed by Westinghouse of reactor vessel eaa have been subjected to cold overpressure events with significant excursions above the Apperdia C limits. Upon completion of these stress analyses, in no case has a problem been noted in regards to sessel lategrity. .

as***me Sheet 4 of 5 !

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~ as TJ. ,a .,id....s...b. .. M.. JJia.65pki .n scuresWpusc ticajdand safety. This information is provided so that you may review the bris cf this nonconservatism and the j anurance of safe plant operation and determine applicability to yort plant.

(, RECOfc2NDATION$

i Several potential methods to compensate for this pressure ircrease are:

1) Reduce the maximum allowable relief valve setpoint (throughout the entire temperature range for which the cold overpressure protection system is enabled) by an amount equivalent to the applicable AP determined for your plant, or

2) Maintain RCS pressure below the Appendix G hearup/cooldown curves during a COMS event by

, an amount equivalcat to the applicable AP when the reactor coolant pumps are in operation, or

3) Restrict the number of reactor coolant pumps which can be in operation below a defined RCS i

1 temperature without drawing a steam bubble in the preuvriser. Dependent upon the' number of reactor coolant pumps that are operable, this configuration may reduce or alleviate the issues associated with the additional AP.

4) The heatup/cooldown curves in the technical specifications typically may include a pressure allowance for instrumentation uncertainty. Based on Westingbouse methodology, this uncertainty is not considered a requirement for COMS overpressure criteria since it has been established that the conservatism inherent in calculation of the heacup/cooldown curves per Appendix G provides i sufficient margin. If your plint's COMS serpoinu were based on curves which included Instrumentation uncertainty, this may be availabic to offset the applicable AP.

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l O PSEG Pubhc Service Ebetric and Gas Company P.O. Box 236 Hancocks Bridge, Nea Jersey 08038 NE I i

MEC-93-917 Q Nuclear Department b, h 4 I

TO: F. Schnarr l Reliability & Assessment Group I

. - gerq;r i46- i FROM: Howard Berrick Salem Mechanical Engineering Supervisor 1

SUBJECT:

NONCONSERVATION IN POPS SETPOINT ATS OPEN ITEM - WESTINGHOUSE NUCLEAR SAFETY ADVISORY LETTER PSE-93-204 (NSAL-93-005B)

DATE: December 30, 1993

Background

Westinghouse NSAL-93-005B transmitted via PSE-93-204 identified'a potential issue regardinga nonconservatisim in the POPS setpoint development. The pressure difference from the wide range pressure transmitter to the reactor vessel midplane (where the

( Tech. Spec. heatup and cooldown pressure / temperature limits are defined) was not considered in Westinghouse analysis.

pressure difference effectively results in the pressure in the This reactor vessel midplane being greater than that seen by the wide range pressure transmitters used to actuate the PORVs, potentially resulting in violation of the Tech. Spec. heatup and cooldown pressure / temperature limit curves.

The Salem POPS analysis (SGS/M-DM-042 and 062) used methodology

provided by Westinghouse in their report " Pressure Mitigating Systems Transient Analysis" (July 1977). The methodology in this report did not consider the pressure difference of concern and therefore the subject NSAL applies to Salem 1 & 2.

Discussion The Tech. Spec. heatup and cooldown P/T limit curves (attached) are determined in accordance with the requirements of Appendix G of 10CTR50 and ensure ' reactor vessel integrity. The pressurizer ,

overpressure protection system (POPS) protects the RCS from exceeding the Tech. Spec. P/T limit curves by opening the PORVs during cold overpressure transients (RCS below 312'F) .

C he pweris in put hxds.

g ' " ' " * *

. - _ . _ - - - . _ . - -- . = - - . - - - - - . . -. . . . . - . . - .

Tha POPS usos tha two wida-rcnga RCS prossuro osnsors PT403 and  !

( PT405 to actuate the PORVs. These sensors sense hot leg  !

}

pressure. Tae pressure at the vessel midplane will be higher i

. than the pressure at the hot leg due to the dynamic and static l 4

pressure difference between the locations. The dynamic pressure  ;

difference depends on the number of reactor coolant pumps (RCPs)

in operation at the time. Westinghouse did not consider the

delta-P associated with the difference in location of the wide-

! range transmitter relative to the vessel midplane leading to a ,

concern that the POPS setpoint may be nonconservative. l The issue of POPS setpoint for actuating the PORVs must be shown to provide adequate protection, with the additional delta-P i incorporated in the setpoint analysis. The Salem POPS analysis a calculated the maximum pressure attained during a cold

. overpressure transient to be 446 psig with the PORV set at 375 psig. Therefore, it must be shown that 446 psig plus the delta-P of concern does not exceed the Tech. Spec. P/T limits, in order

for the POPS PORV setpoint to be adequate.

The Tech. Spec. P/T limit curves define the allowable temperature and pressure combinations ,for heatup rates up to 60* F/hr and for ,

cooldown rates ranging from 0*F/hr to 100*F/hr. For the POPS l analysis a composite curve made up of the heatup curve and the 20'F/hr cooldown curve is used. The use of cooldown rate of

( 20*F/hr is justified because at the low temperatures when POPS is armed, higher cooldown rates are not achievable. A review of the Tech. Spec. P/T limit curves shows that the 20*F/hr cooldown curve is more limiting at low temperatures, on both Units. The pressure that must not be exceeded is 450 psig on Unit 1 and 475 psig on Unit 2.

Additional margin in the Tech Spec. curves can be gained for the  !

POPS application by taking credit for ASME Code Case N514. This  !

cede case-states "LTOP systems shall limit the maximum pressure  !

in the vessel to 110% of the pressure determined to satisfy  !

Appendix G of Section XI, Article G-2215". (LTOP - Low l Tempressure Overpressure Protection is the same as POPS). By taking credit for this Code case, the allowable pressure can be increased by lot. In this case the lowest pressure that must not be exceeded is 495 psig on Unit 1 and 522.5 psig on Unit 2.

l

--; . - - - . . - . - ~ .

Evaluation I. -

Table 1. summarizes the results of the evaluation. Comparing the POPS analysis maximum pressure of 446 psig to the Tech. Spec. P/T limits shows that the margin available to accommodate the delta-P is 4 psig on Unit 1 and 29 psi on Unit 2. An additional 45 psi (Unit 1) and 47.5 (Unit 2) can be gained by taking credit for the code case. Westinghouse indicated that based on generic analyses

. the delta-P is 74 psi with four RCPs operating. It is clear that Unit 1 does not currently have the margin to accommodate.the expected delta-P with four pumps operating.

To quantify the Salem specific delta-P and assess benefit of fewer operating pumps, Westinghouse was requested to calculate the delta-P for one, two and four RCPs operating. The results of the calculation provided delta-P values of 31 psi, 39 psi.and 73 psi for one, two and four RCPs respectively (PSE-93-707).

Westinghouse assumed the transmitters are zeroed out to the RHR-suction line at 92.4 ft. The transmitters are zeroed to the hot-leg (9.7 f t) . To correct for this difference 2 psi was added to the Westin'ghouse results. The maximum pressure including the above delta-P values is ptesented in Table 1.

Table 1 shows that Unit 1 Tech. Spec. minimum of 450 psig is exceeded by the two RCPs and four RCP cases. Taking credit for k, the Code case, the Unit 1 Tech. Spec. minimum is 495 psig which can be met by the two RCPs operating case. The Tech. Spec.

pressure limit increases with increasing temperature and exceeds the four RCPs maximum pressure of 517 psig at 200* F. Therefore, by restricting the number of operating RCPs to two RCPs below 200* F, the POPS PORV setpoint will provide adequate protection.

The temperature of 200* F was selected to coincide with cold shutdown Tech. Spec. operational mode (Mode 5).

Table 1 shows that Unit 2 Tech. Spec. minimum pressure of 522 psig.(taking credit for the Code case) can be met with four RCPs operating. However, to maintain similarity in the operation of the units, and to provide margin for future evaluations of the Tech. Spec P/T curves, the same restriction on RCP operation is recommended on Unit'2.

J It should be noted that 1) restricting the number of RCPs is one of the recommendations in the subject NSAL and 2) taking credit for the 10% margin in the limits as afforded by the ASME Code case was discussed with Westinghouse and this margin has been credited by other utilities (eg FP&L), to address the subject issue.

(

O

Racommendation

(.

In summary to address the POPS setpoint nonconservatisms identified in Westinghouse Nuclear Safety Advisory letter PSE 204, we recommend restricting the number of RCPs in operation while in mode 5 to no more than two RCPs. Procedure change request is being issued to incorporate this change into IOP-2 (Cold Shutdown to Hot Standby) and IOP-6 (Hot Standby to Cold <

Shutdown). The ATS open item NSAL-PSE-93-204 is considered closed by this letter.

GN:

Attachment

.{

C M. Danak 9 K. Pike J. Ranalli J. Serwan j

[

J. Wiedemann ATS File 1 MEC File' l Standards Records Coorindator l

l l

I i

(

-+

. . . - . . . . . . - . - - = . . . . . - . . - . . - . .

, . p~ .

l TABLE 1 l

POPS Maximum Delta P Maximum Tech specs Tech Spec.

PORV Pressure (Vessel Pressure + P/T limits P/T limits Setyt calculated midplane to Delta-P minimum minimum

, Psig In POPS transmitter pressure pressure +10% i Analysis psig Paig 4 RCPs 2 RCPs 4 2 '

(psi) (psi) RCPs RCPs .

Psig psig Unit 1 375 446 73 39- 519- 485 450 495

. Unit 2 375 446 73 39 519 485 475 522.5 i

+ .

t TO: J. SERW84 DATE: 12 28-93 i . Sponsor ki FRCM: MAHESH DANAr EXTENSION: 1872 j Originator MAIL CODE: N50 l DEPARIMENT: SALEM PECH ENGS.

i l- PROCEDURE Ho.: 10P-r a 10P-6 CURRENT REY. VARIOUS

{

PROCEDURE TITLE

COLD SHUTDOWN TO HOT STAND 8Y AND HOT STAND 8Y TO COLD SHUTDOWN OPS i

PROCEDURES FOR SALEM 1 & 2 l

i BRIFF DESCRIPTION OF PROPOSED PROCEDURE / REVISION:

i RESTRICT THE NUMBER OF RCPS IN OPERATION WHILE IN MODE 5 TO No MORE THAN TWO RCPS. SEE ATTACHED MEC-93-917 FOR DETAILS.

k BRIEF PURPOSE /RENEFIT OF PROPOSED PROCEDURE / REVISION:

1

! THEPROCEDURECHANGEWILLADDRESSPOPS/LTOPSETPO!NTNONCONSERVATISMSIDENTIFIEDIN l

WESTINGHOUSE NUCLEAR SAFETY ADV!50RY LETTER PSE-93-204 l -

t MANAGEMENT APPROVAL: (1) DATE:

F REQUESTED CCHPLETION DATE: 3-31-94 CATEGORY: 1 l

! Forward requests for new VPNs or new NAPS to the General Manager -

l Quality Assurance / Nuclear Safety Review. Torward all other requests to the procedure sponsor.

APPROVED: (2)

- Sponsor Date NOT APPROVED:(2)

- < Sponsor Date Return a copy of the completed Request to the Originator.  !

l (1) Management approval required for VPNs, NAPS, and SAPS.

(2) Document reason for disapproval or partial approval e on attached sheet.

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( l 4

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4 i

i

_ 0 TTACHMENT lt3

( Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department MEC-94-630 I

i I

TO: J. Wiedemann l Technical Engineer I

. l FROM: H. Berrick i Salem Mechanical En inee Sup rvisor

SUBJECT:

NONCONSERVATISM IN POPS SETPOINT I REOPENED ATS OPEN ITEM - WESTINGHOUSS NUCLEAR SAFETY ADVISORY LETTER PSE-93-204 (NSAL-93-005B)

REFERENCE:

MEC-93-917 Dated 12/30/93 from Berrick to Schnarr 1

DATE: f May 26, 1994

Background

Westinghouse NSAL-93-005B transmitted via PSE-93-204 identified a

(, potential issue regarding a nonconservatism in the POPS setpoint I

development. MEC-93-917 referenced above provided the resolution. This memorandum [MEC-94-630] stipersedes MEC-93-917.

The pressure difference from the wide range pressure transmitter to the reactor vessel mid plane (where the Tech. Spec. heatup and cooldown pressure / temperature limits are defined) was not considered in Westinghouse analysis. This pressure difference effectively results in the pressure in the reactor vessel mid plane being greater than that seen by the wide range pressure transmitters used to actuate the PORVs, with a potential to result in violation of the Tech. Spec. heatup and cooldown pressure / temperature limit curves.

The Salem POPS analysis (SGS/M-DM-042 and 062) used methodology provided by Westinghouse in their report " Pressure Mitigating Systems Transient Anal'ysis" (July 1977). The methodology in this report did not consider the pressure difference of concern and therefore the subject NSAL applies to Salem 1 & 2.

C n,s w.e a ggggjiw& M M8 REV 111M

a. M eaeaann a ~ ~^- ~ ~ ~ h h i r ~ ^ ~ ^ "~

i L Discussion The Tech. Spec. heatup and cooldown P/T limit curves (attached) are determined in accordance with the requirements of Appendix G of 10CFR50 and ensure reactor vessel integrity. The pressurizer overpressure protection system (POPS) protects the RCS from exceeding the Tech. Spec. P/T limit curves by opening the PORVs during cold overpressure transients (RCS cold leg temperature below 312*F) .

The POPS uses the two wide-range RCS pressure sensors PT 403 and PT 405 to actuate the PORVs. These sensors sense hot leg pressure. The pressure at the vessel aid plane will be higher than the pressure at the hot leg due to the dynamic and static pressure difference between the locations. The dynamic pressure  ;

difference depends on the number of reactor coolant pumps (RCPs)  ;

in operation at the time. Westinghouse did not consider the delta-P associated with the difference in location of the wide- 1 f

range transmitter relative to the vessel mid ple.ne leading to a concern that the POPS setpoint may be non conservative.

The Salem POPS analysis (SGS/M-DM-042 and 062) calculated the maximum pressure attained during a cold overpressure transient to

(, be 446 psig for the mass input case and 418 psig for the heat input case with the PORV set at 375 psig. Therefore, it must be shown that the above peak pressures, plus the additive pressure based on the Westinghouse notification does not exceed the Tech. ,

Spec. P/T limits, in order to comply with the Appendix G l requirement. l The Tech. Spec. P/T limit curves define the allowable temperature and pressure combinations for heatup rates up to 60'F/hr and for cooldown rates ranging from O'F/hr to 100*F/hr. For the POPS analysis a composite curve made up of the heatup curve and the 20*F/hr cooldown curve is used. The use of cooldown rate of 20*F/hr is justified because at the low temperatures when POPS is armed, higher cooldown rates are not achievable. A review of the Tech. Spec. P/T limit curves shows that the 20*F/hr cooldown  ;

curve is more limiting,.than heatup curve at low temperatures, on '

both Units. The pressure that must not be exceeded is 450 psig on Unit 1 and 475 psig on Unit 2.

( .

l l

1 I

l

_____.7_.~._________.

k Additional margin in the Tech Spec. curves can be gained for the .

POPS application by taking credit for ASME Code Case N514. This I code case states."LTOP systems shall limit the maximum pressure l in the vessel to 110% of the pressure determined to satisfy i Appendix G of Section XI, Article G-2215". (LTOP - Low -

Temperature overpressure Protection is the same as POPS). This i Code Case has been now incorporated in the 93 Winter Addenda of ASME Section XI. By taking credit for this Code case or the 1993 l Code, the allowable pressure can be increased by 10%. In this ,

l case the. lowest pressure that must not be exceeded is 495 [vs l 450] psig on Unit 1 and 522.5 [vs 475) psig on Unit 2. However, j neither the Code Case or the updated version of the Code can be l applied at this time pending NRC approval of them for Salem.

Evaluation -

l l Tables 1 & 2 summarize the results of the evaluation. Comparing I the POPS analysis maximum pressure of 446 psig to the Tech. Spec.

P/T limits shows that the margin available to accommodate the delta-P is*4 psid on Unit 1 and-29 psid on Unit 2.

To quantify the Salem specific delta-P and assess the. benefit of fewer operating pumps, Westinghouse was requested to calculate  ;

( the delta-P for one, two and four RCPs operating. The results of the calculation provided delta-P values of 29 psi, 37 psi and 71  !

psi for one, two and four RCPs,respectively'(PSE-93-707).

Westinghouse assumed the transmitters are zeroed out to the RHR .

suction line at 92.4 ft. The transmitters are zerced to the i hot-leg (97 ft). To correct for this difference 2 psi was added '

to the Westinghouse results. The maximum pressure including the above delta-P values or just the static pressure difference between the transmitter and the core mid plane, as applicable, is presented in Tables 1 & 2.

l l

Table 1 shows that Unit 1 Tech. Spec. minimum of 450 psig is i exceeded by less than 1 psi for the inadvertent start of one SI pump for the mass input case. This minor exceedance of the heat up and cooldown curve is not considered an infringement of the Appendix G concern for the following reasons. [a] Recent informal calculation using GOTHIC has reestimated peak pressure for mass input case to be 438 psig, and [b] Recent calculation using RH 3 valve to provide LTOP mitigation has calculated peak pressure for mass input case as 420 psig [ Salem has removed Autoclose Interlock).

(t l

( . Recommendation In summary to address the POPS setpoint non conservatisms identified in Westinghouse Nuclear Safety Advisory letter PSE 204, we recommend restricting the number of RCPs in operation -

while in Mode 5 to no more than one RCP. This has been already incorporated into station procedure IOP-2 (Cold Shutdown to Hot Standby) and IOP-6 (Hot Standby to Cold Shutdown). Future changes to the Station procedures should not rescind the

, restriction of Mode 5 operation with no more than one RCP. The next capsule on Salem Unit 1 is scheduled for removal during the Spring 1995 refueling outage. To address any nonconservative shift of the Appendix G curves at that time, (1) Licensing should pursue approval of Code Case 514 for Salem and (2) Initiate a License Change Request to take credit for RH3 safety relief valve for LTOP. This ATS open item NSAL-PSE-93-204 is considered closed by this letter.

5 D:

q.C, Attachments' C M. Danak V. Chandra

{' C. Lashkari K. OtGara K. Pike -

J. Ranalli J. Serwin j F. Schnarr D. Smith i l

ATS File MEC File Standards Records Coordinator J

(

e d .m, TABLE 1 SALEM ESTIMATED INITIATING EVENT DELTA P [ VESSEL STATIC HEAD CORRECTED LTOP ALLOWABLE UNIT PEAK PRESSURE FOR LTOP MID PLRNE TO FROM PEAK PRESSURE PRESSURE ET)R NUMBER IN PSIG FOR TRANSMITTER IF TRANSMITTER TO AETER LTOP CONDITION-MASS [M] & [ BASED ON TECH APPLICABLE VESSEL MID ADDRESSING TO MEET APPENDI:

HEAT [H] SPEC BASIS [ BASED IN CORE PLANE IF WESTINGHOUSE G CRITERIA INPUT CASES 3/4.4.9 FOR UNIT DELTA P APPLICABLE NSAC-93-05B IN

[FROM 1 & 3/4.4.10 FOR CALCULATED BY -

PSIG [ BASED ON TECH SGS/M/DM 42 UNIT 2] WESTINGHOUSE: & [NOT TO BE SPEC HEAT UP AN:

AND 62 BASED ON ADDED WHERE [ SUM OF COLUMN COOL DOWN CURVE:

RESPECTIVELY) OPERATION OF DELTA P FROM #2 AND COLUMN PROVIDED IN T.S ONE RCP AS PREVIOUS COLUMN 84 OR 55 AS FIGURES 3.4-3 PROCEDURALLY IS ADDED APPLICABLE] AND 3.4-4]

RESTRICTED IN ALREADY]

• MODE 5)

UNIT 1 446 PSIG THE START OF A N/A 97'MINUS 450.7 PSIG 450 PSIG MASS INPUT SI PUMP AND ITS 86'-2" = 10.8' INJECTION INTO A = 4.7 PSIG WATER SOLID RCS

[NO RCP RUNNING)

UNIT 1 418 START OF AN IDLE 29+2=31 PSIG N/A 449 PSIG 450 PSIG HEAT INPUT RCP WITH THE [29 IS ft)R 1 SECONDARY WATER RCP FROM PSE-TEMP OF S/G LESS93-707 AND 2 THAN OR EQUAL TO PSI WAS ADDED 50 F ABOVE RCS TO CORRECT COLD LEG TEMP TRASMITTER EL OF 92.4' TO HOT LEG EL OF 97']

O

~ . . ~ . _ _ _ . _ . . _ _ _ _ . _ _ _ - - - _ . - _ _ _ - _ - . . _ _ - - - _ _ . . . - _ _ _ _ _ . _ . - - _ . _ . _ . . . . . . . ..

m ~ O  !

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TABLE 2 <

i SALEM ESTIMATED INITIATING EVENT DELTA P [ VESSEL STATIC HEAD CORRECTED LTOP ALLOWABLE  ;

UNIT PEAK PRESSURE FOR LTOP MID PLANE TO FROM PEAK PRESSURE PRESSURE EX)R 6 NUMBER IN PSIG ft)R TRANSMITTER IF TRANSMITTER TO AFTER LTOP CONDITION MASS [M] & [ BASED ON TECH APPLICABLE VESSEL MID ADDRESSING TO MEET APPENDI)l l HEAT [H] SPEC BASIS [ BASED IN CORE PLANE IF WESTINGHOUSE G CRITERIA INPUT CASES 3/4.4.9 FOR UNIT DELTA P APPLICABLE NSAC-93-05B IN 1& 3/4.4.10 FOR CALCULATED BY -

PSIG [ BASED ON TECH

[FROM UNIT 2] WESTINGHOUSE; & [NOT TO BE SPEC HEAT UP ANI I SGS/M/DM 42 BASED ON ADDED WHERE [ SUM OF COLUMN COOL DOWN CURVE: !

l AND 62 OPERATION OF DELTA P FROM #2 AND COLUMN PROVIDED IN T.S.

, RESPECTIVELY] ONE RCP AS PREVIOUS COLUMN #4 OR #5 AS FIGURES 3.4-3

PROCEDURALLY IS ADDED l

APPLICABLE] AND 3.4-4]

RESTRICTED IN ALREADY]  !

MODE 5)

UNIT 2 446

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[NO RCP RUNNING)

\

UNIT 2 418 START OF AN IDLE 29+2=31 PSIG N/A 449 PSIG 475 PSIG ,

HEAT INPUT RCP WITH THE [29 IS FOR 1

• SECONDARY WATER RCP FROM PSE- l TEMP OF S/G LESS93-707 AND 2 ,

THAN OR EQUAL TO PSI WAS ADDED i 50 F ABOVE RCS TO CORRECT  !

COLD LEG TEMP TRASMITTER EL l OF 92.4' Td HOT  !

LEG EL OF 97']  !

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8 4 4 i teettatts TSurtRat W (St$.P) t j Salem Unit 2 reactor coolant system heatup limitations applicabit l j for the first 10 EFPY with maximum heatup rate of 60'F/hr

' FIGURE 3.4-2 I  !

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SA!.Dt UNIT 2

'3/4 4-28 Amendment No. 86

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l CONTROLLING MRTER!AL: LONGITUDINAL WELD COPPER CONTENT: {

0.35 WT5 l NICKEL CONTENT:

l 1.00 WT5 j INITIAL RTgy: -56*F l k- RT ET AFTER 10 EFPY: 1/47, 178.4'r 3/4T, 116.1'F i CURYt3 APPLICA8LE FOR C00LDOWN RATE 5 UP TO 100'F/NR FOR THE 5tRV PERIOD UP 7010 EFPr AND CONTAINS NO MARGIN FOR P055!8LE I i

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! for the first 10 (FPY t

4 FIGURE 3.4-3

sAr.tM - uwrt 2 3/4 4-29 Amendment No. 86

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{ i WECID'BECU/Av.S.A 8%i N..mSSAG.m

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l Q PH: (8L41774-3151 FRX: [8141774-2646 W: Public Service Electric & Gas '

1 Salem Nuclear Station FAX NO* 609~339~2749 i

DATE: December 18,1191 ATT'N: charles Lascares

REFERENCE:

J FROM:

Frank Leichtenberger SUB.1ECT: Minimum Flow Area of Valva per our Dwg. L-138088 MESSAG13 Our l ayoutshowstheminimum.llowareatobe1.716sq.in.

.. Regards.' '

1 ,

i i Frank Leichtenberger i

for Ken Meyer j FL/kmtl i

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REAE*DR COOLANT SY1?fM

, DVERPREE stfRE PRO?te?f0M EYSTEME 1 LIMITING CONDITION FOR OPERATION i

3.4.9.3 At least one of the following overpressure protection systems shall l be OPERABLE:

. l l

! a. Two Pressuriser overpressure Protection System relief valves j

(POPS) with a lift setting of less than or equal to 375 psig, er

b. A reactor coolant system vent of greater than or equal to 3.14T i j

, square inches. #  !

! l APPLICARILITY: When the temperature of one er more the RCS cold lege is less j- than or equal to 312'F, except when the reactor vessel head is removed. j M8, , , .

I a. With one POPS inoperable, either restore the inoperable POPS to  ;

} OPERABLE status within 7 days or depressurise and vent the RCS l l

, through a 3.14 square inch vent (s) within the nest 8 heures maintain l

{ ' the RCS in a vented condition until both POPS have been restored to

I 1

OPERASLE status.',

I b. l ,

I With both POPS inoperable, depressurise and vent the ACS through a 3.14 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a  !

l (

4 vented condition until both POPS have been restored to OPERA 3LE tus. -

i c.

In the event either the POPS or the ACS vent (s) are used to mitigate a RCs pressure transient, a Special Report shall be prepared and submitted to the cessaission pursuant to specification 6.9.2 within 30 days. The report shall describe the circumstances ,

initiating the transient, the effect of the POPS or vent (s) on the  !

transient and any corrective action necessary to prevent recurrence.  ;

I sURvErz. LANCE REoczRExtNTs 4.4.9.3.1 Bach POPS shall be demonstrated OPERASLE by:

J T

SALEM - UNIT 1 3/4 4-30' Amendment No. 1E w.

. , . , . . . e. "

REACTOR C001#ff SYSTEM NES 3) i i

Finally, the new 10CFR50 rule which addresses the metal temperature of the closure head flange regions is considered. This 10CFR50 rule states that the metal temperatuge of the closure flange regions must escoed the asterial RT

' byatleast120Ffornossaloperationwhenthepressureescoeds20percentT thepreservicehydrostatictestpressure(621peggforSales).

33/4.4-1 indicates that the limiting RT Table i

of 23 F occurs in the closure head l flange,F 1s let at pressures greater than 621of Sales Unit 1 and the minim Figures 3.4-2 and 3.4-3. psig. These limits do not affect 1

i Although the pressuriser operates in temperature ranges above those for which tLere is reason for concern of non-ductile failure,* operating limits are performed in accordance.with the ASME Code requirements.provide

- - ~ -- -

The OPERASILITY of.ggg Pgs o,g an RCS 4 m-t!an}_pagging_of greater than_),di, /

square inches ensures thht the RCS will be protectd free pressure transients

.e(d4Gaterelievingcapability'toprotecttho'RC a the.R.C.S cold legs are less than or.squal to 312_F Either POPS has ~-

en i

the tr'ansient is limited __to either (1) the sta~rt of an

secondary water temperature of the steam generator less, idle.RGP than or equal with the to 50'F j above the BCS cold leg temperatures, or (2) the st

( papanditsinjectioninto',awatersolidRCS.]artofasafetyinigst_fon ~'

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SALIM - WIT 1 3 3/4 4-11 Amendment No.108

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3 CVERPRESEURE PTOTECTION SYSTEMS 4

LIMITIN3 CONDITION TCR OPERATION 3.4.10.3 At least one of the following overpressure protection systems t 1

shall be OPERABLE: '

i

a. Two Pressuriser Overpressure Protection System relief valves  ;

(POPSs) with a lift setting of less than er equal to 375 peig, 1

_o_ r

=... ---...-.~_

b. The Reactor Coolant System (RCS) depressurised with an RCS vent  !

4 of greater taan or equal to 3.14 square inches. , ,

i APPLICARILITY: When the temperature of one or more of the RCS cold legs l is less than or equal to 312'T, escept when the reactor vessel head is  ;

removed.

l 4

&EZZ,Qg -_-

l

a. With one POPS inoperable, restore the inoperable POPS to )

. OPERASLE status within 7 days or depressurise and vent the RCS  !

. through a 3.14 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. )

i With both POP 8s' inoperable, depressurise and vent the RCS

~

b.

, through a 3.,14 square inch vent (s) withih 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c. In the event either the POPSs or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to specification 6.9.2 within 30 days. The report shall describe the circumstances initihting the transient, the effect of the PCPSs or vent (s) on the transient and any corrective action necessary to prevent recurrence.

$URVIILLANCE REQUIREMENTS 4.4.10.3.1 Each POPS sht11 be demonstrated CPERASLE by:

l 1

l SALEM - UNIT 2 3/4 4-31 Amendment No.110 l

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arAcToa cootANT syster h rt '

\

Finally, the new 10c m50 rule which addresses the metal temperat ure of the metal temperatuge of the closure flangea regions the  !

by at least 120 F for normal operation when the pressure exc thepreservicehydrostatictestpressure(621peggforSal percent es). Table j 83/4.4 1 indicates that the limiting RT is 148'T at pressures greater than 621 psig.flangeof'Sa Figures 3.4-2 and 3.4 3. s region

'these limits do not affect i

Although the pressurizer operates in temperature ranges above th l  !

there is reason for concern of non ductile, operating ose for which failure limits are performed in accordance with the ASME Cada {

j l

The OPERASILITY of two POPSs or an RCS vent opening an 3.14 of greater th i

which could exceed the limits of Appendix of the RCS cold legs are less than or equal to 312 F.

G co 10 ansients en one or more adequate relieving capability to protect the RCSEither fromPOPS has overpressuri )

zation when the transient is limited to either (1) the starte of RCPanwith idlthe  !

above the RCS cold leg temperatures, or (2) the pump and its injection into a water solid RCS.

to 50*F ection

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SALEM - UNIT 2' 5 3/4 4-12 Amendment No. ,c6 P'"" ~ -

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' Volume HJNumber 29*

, SA.EM GEN STATION h h "" " "

l

• Newsletter 3 W C "M/o tWe'ekof th1/15/93T

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)roundvoltage relay areaseded to tand output breakers to a tiga

',- a snain generator M (PS 2737) masapsacas open resultingla a v I PS ,, 1 ras

• M Measew

$ f C "' --

r trip. Cause of the IPut breaker ground N CANYON.1/2 (W) An k OE providedinbreation O

tionis being TOP system PORVpressure I.

t

!741) point of 450 psigwas akadated by Westinghouse without consideradaa a' Q L '

U h tmaumlG1[J the dynamicpressuredissrenosI tween the Rx pressurevesseland

'] ,

E)The Assistant pressure sensors. Preitadnarym) ervbor noticed that tions indkats that the current L1 C,

pgb legM and LPS! valves, 3626 and 3645,were system PORVpressuresetpolati psig could resultin a Tech Specv .

. nave been closed. tion in the event of a LTOP at 6{h ggu/ ,/ pd/

()

los indiated that sa oup actuation relay la Mode 4 whenanyRCS tureis at or below c o< to us.

) ;these valves to opea 5 and 6 wi  :: uscihas4 esition. Tbc relay on. (ra uu; instad. (PS 2737)

~ ~: ' =pM:.W4.2 g- During the perf

.c.us.A y o nc.pg . . -

elcalibration

• M E 9 sam ( $ di G i -2 d j sa a RWSTievel 0$N

-4..tes;{4&g*rgr.rpjj h placed al -e-n,

./. ..n%4 sanitterin the trij .. .

sanoswith theis . ' Q$(qgl$" ' ' ' ,.gi,y)*

' *a'e 4 er so verify that il . .,te sees signal $Dr th . .c /.es.;3

,,,g ,,,p g g ., , -'g-asiefwet P. see,reni (609) 339-2068 #

^W u, _ _ _ _ __ _ _ . .

Ip was pressat 11 if a saistyinjessic

._T"-

J'

,, M RG Sowinto the core resultlagin a were to occurwith the plaatla thi:. ,,,

rens Meyadditloa. Theincreased Sow con 8guration,the potentialexism s, e d1 was due to the RCP speedinatasing bind both the residualheat remon . Y.

when the main turbine speed increased (RHR)and containment budding:;- ~; ,

des to aninstantaneous loss ofload.

pumps. The plant was la this coni..t. '

4.,

1heloss ofload was attributed to the ration forless than 10 minutes. TPa. -

,, g, sg ..!TT4

. main transformer sie breaker to the con 8guration placed the plantlaa Tanaicina , rec,ognised the biled sensor switchyard opening due to spurious ac. condition that alone could have pre. and restored thecomputerpoint for tuntion of the high side overcurrent ves.ted the Ad811: seat of the safety RWCU Sowrate. (PS 2738.OE 5776))

seley. (PS 2739) Amction of structures or systems that tl (UQg//w/Yb 1

- - - - - - _. .. _. . ~. - ~ . . ~ . - _- . - - - _ - _ - . - - - . . - . . - ~ _ - - . -

,) hlt (S E*W saf8  :

j

( RI 1182 I FULLBRIGHT (INPO) 30-JUL-93 09:58 EDT j

! A

Subject:

NRC*INFORMATION NOTICE 93-58 '

i 2- l i

UNITED STATES I NUCLEA'R REGULATORY COMMISSION l l OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 a

i .

July 26, 1993 l NRC INFORMATION NOTICE 93-58: NONCONSERVATISM IN LOW-TEMPERATURE OVERPRESSURS PROTECTION FOR PRESGURIZED-WATER REACTORS Addressees All holders of ooeroting licenses or construction permits for pressurized , water reactors. .

i j Purpose ',

j The U.S. Nuclear Regulatory Co,mmission (NRC) is issuing this -'

l information notice to alert addressees to a nonconservatism in the '

low-temoerature overpressure protection (LTOP) setpoint calculation for Westinghouse facilities. It is expected that recipients will review the information for soplicability t'o their facilitica and I consider actions, as appropriate, to avoid similar problems.  !

However, suggestions contained in this information notice are not recuirements: therefore, no specific action or written response is reauired.

Description of Circumstance On October 29, 1992, the Texas Utilities Electric Company, the licensee for Comanche Peak Steam Electric Station, reported that its existing lou temperature overpressure protection may not have provided the required margins of safety against reactor vossal brittle frac,.ture under certain overpressure transients.

The same condern was Irper reported by the licensees f or Byron, Zion, Diablo Canyon, Kewaunee, Sequoyah and Point Beach nuclear stations.

Discussion In reactor units designed by Westinghouse, overpressure protection of the reactor vessel at low temperature conditions is provided by a cold overpressure mitigation system (COMS). This system compares pressure and temperature inputs against a preset setpoint curve and relieves the pressure when the setpoint is reached. This protection is necessary because, while at low temperatures during plant startup and shutdown conditions, certain transients could cause the reactor coolant system pressure to exceed the reactor vessel pressure-temperature (P-T) limitations established for protection against  ;

C 37 [ M __

brittlo frcqturo. A couricus otcrt of a cOfoty injocticn pump,

{ reactor cookhnt pump, or other operational errors could activate this

. system. During such events, the P-T limitations are maintained by I

opening the-h ossurizer power-operated relief valves (PORVs) or

, safety relief valves in the residual heat removal (RHR) suction lines

to relieve system pressure.

The transmitters that provide pressure signals to the COMS are l l located at the primary system hot les piping of the reactor vessel. {

During low temperature operation of the reactor coolant pumps, l

dynamic pressure in the reactor vessel would be higher (by the amount of flow loss in the core and vessel outlet) than that sensed in the i hot leg. Additionally, the static head correction for the dif ference I

) in elevation of-the sensor to the core regi6n was not considered.

i The resulting pressure dif ference between the sensor and the

. vulnerable location in the reactor vessel could be as high as 790 kPa I (100 pais), depending on the number of reactor coolant pumps in

] operation and the location of the pressure-sensing taps. The LTOP setpoint curve that was originally developed by Westinghouse did not j 4 take these factors into consideration.  !

I Westinghousefhassent a letter to licensees recommending one of t e

following methods to compensate for this pressure increase: 1) d i reduce the maximum allo 6able re, lief valve setpoint by en amount eaufvalent to the plant-specific calculated difference in pressure, j

_ 23 maintain RCS pressure below the'heatup/cooldown curves by a value i '~ equal to the plant specific difference in pressure from both flow  !

loss and elevation dif ference when the reactor coolant pumps are in I operation, 3) restrict the number of reactor coolant pumps and, therefore, the flow loss error that can be operated below a defined RCS temperature without drawing a steam bubble in the pressurizer, or

4) demonstrate that the available margin in the LTOP calculation, taking into account instrumentation uncertainty, is sufficient to offset the plunt- specific pressure dif f erence.

l The Westinghouse letter describes interim administrative controls as I well as calculational methods to verif y setpoint adeauscy f or i addressing th,e LTOP concern. The staff notes that the administrative restrictions described in approaches (2) and (3) are intended by Westinghous to provide interim actions to address the concern only until LTOP- 'tpoints are verified to be adeauste or are revised appropriate sin technical specifications.

Although tho' information in this notice addresses the cold overpressure mitigation system at Westinghouse designed plants, espects of this issue may also be applicable to other PWRs.

This information notice requires no specific action or written response. If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

ORIGINAL S'GM70 BY 33

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