IR 05000443/1988010

Insp Rept 50-443/88-10 on 880706-0906.Violations Noted.Major Areas Inspected:Actions on Previous Insp Findings,Nrc Bulletins & Info Notices,Operational Safety,Potential ROs & Operational Events & Maint & Surveillance Activities

ML20155E638
Person / Time
Site:	Seabrook
Issue date:	09/28/1988
From:	Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20155E606	List: IR 05000443/1988010 ML20155E603 ML20155E614
References
50-443-88-10, IEB-87-002, IEB-87-2, IEB-88-005, IEB-88-5, IEIN-86-050, IEIN-86-50, IEIN-88-025, IEIN-88-046, IEIN-88-25, IEIN-88-46, NUDOCS 8810120361
Download: ML20155E638 (66)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No.: 50-443/88-10 License No.: NPF-56 Licensee: Public Service Company of New Hampshire 1000 Elm Street Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit N Inspection At: _S_eabrook, New Hampshire inspection Conducted: July 6 - September 6,1988 and September 21, 1988 Inspectors: A. C. Cerne, Senior Resident Inspector, Seabrook Station D. G. Ruscitto, Senior Resident Inspector, Seabrook Station E. Yachimiak, Operations Engineer (Examiner), PWR Section, Division of Reactor Safety C. J. Conklin, Senior Emergency Preparedness Specialist, Emergency Preparedness Section, Division of Radiation Safety and Safeguards Approved By* O $ ^uo w3 (2AlN Donald R. Haverkamp, Chiaf, eactor Projects Date Section No.3C Inspection Summary:

Areas Inspected: Routine inspection on day and backshirts by two resident inspectors and two regional specialist inspectors of actions on previous inspection findings, NRC Bulletins and Information Notices, operational safety, licensee potentially reportable occurrences and operational events, maintenance and survelliance activities, design changes, allegations, training, and electrical c(nfiguratior, contro G810120361 881006 POR O ADOCK 05000443 PDC

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Inspection Summary (Continued) 2

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Results:

1. General Conclusion A repetitive weakness was identified in the implementation of the tagging program involving physical removal of a section of non-safety related piping containing a valve which was caution tagged. While the non-safety nature of the equipment indicates that regulatory requirements were not violated, the recurrent nature of the incident indicates that further management attention in this area is warranted (Refer to paragraph 8.b).

A weakness was identified in the licensee's reporting system with respect to diesel generator failures (Refer to paragraph 4.k)

A weakness was identified in the calculations associated with ncn-class 1E loads powered from class 1E power sources. Licensee evaluat',on of this problem is continui19 and is being tracked under existing unre solved item 88-06-01 (Refer to paragraph 4.j).

A licensee strength was demonstrated in the handling of testing and inspection of flanges and fittings in accordance with NRC Bulletin 88-0 Strong participation by quality assurance and engineering personnel con-tributed to the licensee's ability to respond to this industry wide problem in a timely fashion (Refer to paragraph 6.b). Violation A violation was identified regarding the failure to report diesel gener-ator failures in accordance with the technical specifications (Refer to paragraph 4.k.).

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TABLE OF CONTENTS Page Persons Contacted.............................................. 1 Summa ry of Faci l i ty and NRC Acti vi ti es. . . . . . . . . . . . . . . . . . . . . . . . . 1 Resident Inspector Activities............................. I Visiting Inspector Activities............................. 1 Plant Status.............................................. 2 Operational Safety............................................. 2 Plant Inspection Tours (71707, 71710)*.................... 2 Operational Events (93702) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 License Action on Previous Inspection Findings (92/01). . . . . . . . . 5 Unresolved Item 86-54-02: CBS Piping Design.............. 5 Unresolved Item 87-10-02: RHR Valve Alignment Question Unresolved Item 87-16-03: Operation of the SUFP on an Emergency Bus........................................... 7 Inspector Follow-up Item 87-22-01: Siren Modifications... 8 Inspector Follow-up Item 88-09-01: TSC/E0F Technical Support................................................. 8 Inspector Follow-up Item 88-09-02: TSC/OSC Multiple Access Po1nts........................................... 11 Inspector Follow-up Item 88-09-03: Departing Shift 0osimetry............................................... 11 Inspector Follow-up Item 88-09-04: Media Center Responses to Press Inquiries...................................... 11 Unresolved Item 88-02-01: SI Accumulator Isolation Valve Control Circuitry................................. 12 Open Item 88-06-01: Non-Class IE Loads Powered From Class IE Sources........................................ 13 Violation 88-06-02: EDG Failure Reporting................ 16 5. ' Licensee Reports (92700)....................................... 17 Construction Deficiency Report 86-00-09: Veritrak/Tobar Transmitters............................................ 17 CFR 21 Report 87-88-04: Gould Relay Failures.......... 18 CFR 21 Report 87-88-03: Service Water System Valve Liners and Seats........................................ 18 Station Information Reports............................... 20

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Table of Contents (Continued)

Page NRC Bulletins and Information Notices (92701).................. 20 NRC Bulletin 87-02: Fastener Testing to Determine with Applicable Material Specifications................. 20 NRC Bulletin 88-05: Nonconforming Materials Supplied by PSI and WJM.......................................... 21 NRC Information Notice 88-46: Licensee Report of Defective Circuit Breakers.............................. 22 NRC Information Notice 88-25: Minimum Edge Distance for Expansion Anchor Bo1ts.................................. 23 IE Information Notice 86-50. Inadequate Testing to Detect Failures of Safety-Related Pneumatic Components or Systems.. .............................................. 23 Surveillance / Maintenance (61840, 61726, 62703)................. 24 OX 1456.81: Operability Test of ISI Valves ............... 24 EX 1804.044: Safety and Relief Valve Setpoint Pressure Test.................................................... 24 EX 1804.016: Diesel Generator Auxiliary Coolant System Quarterly Test ...................................... 24 IX 1680.921: SSPS Train "A" Actuation Logic Test ......... 24 EX 1804.015: Diesel Generator 1B 18-Month Operability and Engineered Safeguards Pump and Valve Response Time Testing Mode 5 Surveillance ............................ 25 X 1406.02: CBS Pump and Valve Quarterly Test and 18-Month Remote Position Indication ............................. 25 Residual Heat Removal (RHR) System........................ 26 Design Changes and Modi fications (37700, 37701). . . . . . . . . . . . . . . . 27 Post Accident Sample System (PASS)........................ 27 Secondary Component Cooling Water (SCCW) System. . . . . . . . . . . 07 Allegation Followup (92701).................................... 29 1 Training (41400, a1701)........................................ 32 General Employee Training.......... ...................... 32 Operator Training......................................... 33 1 Electrical Configuration Control (92701)..................... . 33 12. Management Meetings (30703,30702)............................. 34

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Table of Contents (Continued)

Attachments: Meeting Attendees, Meeting conducted August 17, ik Meeting Slides, Meeting conducted August 17, 1988

The NRC Inspection Manual inspection procedure that was used as inspection guidance is listed for each applicable report sectio t til

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DETAILS 1. Persons Contacted - New Hampshire Yankee (NHY)

E. A. Brown, President and Chief Executive Officer

1. W. A. DiProfio, Assistant Station Manager

• T. C. Feigenbaum, Vice President, Engineering, Licensing and Quality Programs W. J. Hall, Regulatory Services Manager

• D. E. Moody, Station Manager G. S. Thomas, Vice President, Nuclear Production

• J. M. Vargas, Manager of Engineering

• J. J. Warnock, Nuclear Quality Manager

Attended exit meeting conducted on September 9, 1988

1. Attended exit meeting cond" ted on September 22, 1988 Interviews and discussions with other members of licensee and contractor management, and with their staf f s, were also conducted relative to the inspection of items documented in this repor . Summary of Facility and NRC Activities Resident Inspector Activities On August 8-11, 1988, the Resident Inspector attended a Resident Inspector Seminar in King of Prussia, Pennsylvani On August 8-19, 1988, the Senior Resident Inspector travelled to Rockville, Maryland for a temporary assignment with the NRC Office of Nuclear Reactor Regulatio On August 17, 1988, the resident inspectors attended a management meeting between the NRC and NHY in King of Prussia, Pennsylvani (Refer to paragraph 13 of this report)

On September 1, 1988, the Senior Resident Inspector was reassigned to another duty station. The Resident Inspector was assigned as Senior Resident Inspector, Visiting Inspector and NRC Management Activities On July 18-22, 1988, an NRC Region I operations engineer (examiner)

conducted a routine inspection of plant operations and previously identified item His inspection findings are included in this repor .

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On August 16, 1988, the Director, Office of Nuclear Reactor Regula-tion visited the site. He held discussions with the Resident Inspec-tor and toured the plant. The NHY inventory department staff was requested to provide information concerning the Seabrook program for material receipt inspection and identification of fraudulent or substandard part On September 21, 1988, an NRC Region I senior emergency preparedness specialist conducted a routine inspection of previously identified item His inspection findings are included in this report, Plant Status During this r~orting period, the plant remained in operational Mode 5, cold shutJown, with primary temperature between 105 and 140 degrees F and depressurized. Major maintenance was conducted on ser-vice water cooling tower pump SW-P-110A, the reactor trip breakers, the chemical and volume control system, the control building air handling sy stem , the waste gas system, the diesel generators and switchyard circuit breakers and bus duct Major 18-month surveillance was conducted on the emergency diesel generators, emergency core cooling systems, engineered safety fea-tures actuation systems and ventilation filter On July 19, 1988, while performing surveillance testing on the train

"A" containnent building spray system, an improper valve lineup caused approximately 5,000 gallons of water from the refueling water storage tank to flow to the suction of the operating train "A" residual heat removal pump suction and into the reactor coolant sys-te Details of this event may be found in paragraph 7.f of this repor Significant design changes were initiated on the secondary component cooling water and post accident sampling systems. Further discussion of these changes may be found in paragraph 8 of this repor A major licensee activity involved identification and testing of flanges and fittings in accordance with NRC Bulletin 88-05. Further inspection of this bulletin may be found in paragraph 6.b of this repor . Operational Safety

, Plant Inspection Tours The inspectors observed station activities and plant status during general inspections of the plan The inspectors examined work for any apparent defects or noncompliance with regulatory requirements or license conditions. The inspectors interviewed station staff and contractor personnel in their work area .

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During control room observation periods, during both normal working hours and on backshifts, the inspector reviewed control room logs and records including night orders, shift journals, shift turnover sheets, the temporary modifications log, and control board indica-tions. Specific note was taken of equipment in "pull-to-lock" condi-tions, equipment tagged, alarm status and adherence to technical specification (T.S.) limiting conditions for operation and action statements. Also, boron samples, taken from the reactor coolant system and connected water supplies, were spot-checked for concen-tration, sample frequency and documentation in accordance with specified zero power license condition The inspector verified the proper position, in accordance with oper-ational procedure or work controls of various valves, switches and breakers during system . walk-downs and checked the valve and switch status in the control room. Similarly, temporary modifications and component tagging, maintenance work, and design change implementation activities, as observed during plant inspection tours, were evaluated for evidence of both proper field controls and coordination of the subject work activity witn the control room and operations personnel on shif In certain cases, the operability of specific components and the applicability of the observed work to the T.S. requirements were discussed with the op?rator The inspector identified several minor discrepancies in material conditions. A list of items was provided to the license Action taken on each issue is described belo (1) Design coordination report (DCR) 87-0185 changed out certain switches on the main control board (MCB). The inspector ques-tioned when the new identification labeling will be complete The licensee provided work request (WR) 87 WOO 7159 initiated on September 30, 1987 to have the labeling finishe (2) The startup rate meter for nuclear instrument channel N310 on the MCB f requently sticks downscale and requires manual agita-tion to free the pointer. The inspector questioned the status of resolving this issue since it has been a recurring problem.

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Request for engineering services87-452 was initiated on

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January 6, 1937. Meter operation under normal neutron flux will be observed during the upcoming test program to verify that the present condition is being caused by Icw core activity level (3) The lens on the indicating light on the MCB for safety injection accumulator SI-TK-9C nitrogen vent valve (SI-FV-2477) requires engraving. The licensee initiated WR 88-2514 to accomplish this task.

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(4) The inspector identified a disassembled conduit clamp on instru-ment rack KM-IR-73 in the service water pumphouse. The licensee took corrective action to reclamp the condui (5) The 345kV schematic drawing posted on the wall of the relay room was not being controlled as an approved operator ai The licensee provided a new controlled copy of the drawing and posted it in accordance with NHY guidelines delineated in the Operations Management Manual, Chapter 8, "Operator Aids".

On July 13, 1988 while touring the tank f arm, elevation 20'-0", the inspector noted valves CBS-V39 and CBS-V44 unlocked and closed. These valves are normally locked open. The inspector verified that the locked valve log in the control room reflected the current status of the valves and determined that adequate controls were in place to ensure that the valves would be returned to their proper positions when require While touring the control room on July 20, 1988, the inspector noted that suction pressure for train "B" emergency feedwater (EFW) pump FW-P-378 indicated 6 psig, while the suction pressure for the train

"A" pump (FW-P-37A) indicated zero psi The inspector verified by inspecting the EFW pumphouse that the suction valves to each pump were danger tagged closed and that plastic isolation "pancakes" had been installed downstream of the suction valves to keep the pump casings dry. Since the tap for the FW-P-378 suction pressure instru-ment is between the "pancake" and the closed suction valve, any leak-age past the suction valve or trapped pressure would be sensed by the suction pressure instrument. Based on this information, the inspector had no further question While touring the essential switchgear rooms the inspector noted that the indicators for containment building spray (CBS) system sump level were not identified. These level indicating tranmitters CBS-LIT-2384 and CBS-LIT-2385 were installed by engineering change authorization 03/109038H in 1985. The inspector reviewed the above ECA along with the applicable design change notice (DCN 65/0259A) and budget expense revision (BER 742A). The licensee stated that the indicators will be

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labele b. Operational Events (1) Paragraph 4 9 of this report details a reporting deficiency con-cerning diesel generator failures. As described in that para-graph the licensee instituted a new reporting procedure utiliz-ing the station information report (SIR) process. Subsequent to this procedural modification, two additional failures occurre The inspector reviewed the preliminary SIRS on the failures which occurred on August 11 and 12, 1988 on the train "B" engin These failures will be the subject of 30-day reports to the Commission in accordance with Seabraak Technical Specifica-tion _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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l (2) On August 10, 1988 the electrical load dispatcher offsite opened up 345 kV circuit breaker No.163 in the switchyar At the time, 345 kV circuit breaker No.11 was open and out of service

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I for maintenance. Train "B" emergency diesel generator (EDG) was l also out of service for maintenance as permitted by Technical l Specification The result of this breaker opening was an undervoltage condition to buses E5 and E6 and the resulting automatic start of the train "A" EDG. As expected, no transfer of power from the unit auxiliary transformer to the reserve ;

auxiliary transformer occurred, and power was restored by manual '

operator action without inciden The licensee made a non-omergency report to the NRC operations center in accordance with ;

10 CFR 50.7 The inspector reviewed the preliminary station information report and will followup licensee activities under the licensee event report when issue (3) On July 8, 11, 13, 15, 20, 26, August 3, 1988, the licensee made 48-hour, non-emergency calls to the NRC Operations Center via the emergency notification system pursuant to NRC Bulletin 88-0 Additional information on this issue may be found t paragraph 6 b of this repor I 4. Licensee Action on Previous Findings l (Closed) Unresolved Item 86-54-02: Containment Building Spray (g S).

Pump Suction Pipina Desian Question The primary issue _ raised with this unresolved item involved questions of code compliance and ;

adequacy of the overpressure protection of a portion of the CBS sys- f tem piping. Since the residual heat removal (RHR) system piping is designed to higher system pressure requirements than that of the CBS system, the adequacy of a single check valve in each of four lines interconnecting the RHR and CBS systems was evaluated with respect to design commitments, American Society of Mechanical Engineers (ASME) ,

Boiler and Pressure Vessel Code interpretations, and current ASME Code guidanc The inspector held several meetings, including telephone conferences, with licensee engineering and licensing personnel during the first half of 1987 to discuss the subject design questions. The original temperature / pressure design data for the CBS piping was reviewed and an ASME Code subcommittee member was interviewed in regard to precise interpretation and requirements of Section NS-3612.4 of the ASME Code,Section III (1971 Edition, Winter 1972 Addenda). Furthermore, !

the NRC Office of Nuclear Reactor Regulation (NRR) became involved in the question of original design adequacy and FSAR commitments. As i stated in Supplenent No. 7 to NUREG-0B96, the Seabrook Safety Evalua-tion Report (SER) issued in October, 1987, the NRC staff concluded that:

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"Although the current guidelines in Section 12I of the Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers (ASME Code) stipulate the use of two series connected check valves for such system interface applications, the appli-cant is in compliance with the ASME Code requirements under which the Seabrook RHR and CBS system piping was designed and constructed."

Therefore, the question of the code compliance of the original CBS system design was reviewed and determined to be adequate by NR i

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However, based upon concern over the potential for RHR system leakage to the CBS system pump suction piping, as had been noted to occur in late 1986, the licensee committed te implemer,t both short and long-term corrective action The major element of the licensee's short-term actions involved the installation of a piping thermal monitoring system (PTMS) which generates an alarm in the control room when the CBS system piping temperature profile indicates that leakage from the RHR system is occurring. Operators could then evaluate and estimate '

the RHR-to-CBS system leak rate and respond with the appropriate valve and system realignment The inspector witnessed field activities associated with the instal-lation of the PTMS, examined the final thermccouple locations and revi eweri the operator alarm response actions, As documented in Supplement No. 7 to the SER, the NRC staff concluded that the licen-see's short-term actions were sufficient to resolve concerns of CBS system overpressurization due to RHR check valve leakage and to allow operation with the present CBS/RHR pressure isolation configuration until the first refueling outag The performance of longer-term corrective measures, such as the installation of redundant motor operated gate salves in series with the existing check valves, is currently being scoped and analyzed by the licensee. The need for such action is a full power licensing issue / condition, as noted in SER Supplement No. 7, which resides under the purview of NRR for future evaluatio With respect to the acceptability of existing field conditions and to the adequacy of licensee contingency actions in response to the subject RHR check valve leakage, no concerns remain and no additional safety questions have ben identified. While NRR has further licen-sing action on this matter, as an inspection issue all the releunt parts of this item have been resolve This issue is considered closed.

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. (0 pen) Unresolved Item 87-10-02: NRC Information Notice 87-0 "RHR Valve Misalignment Causes Degradation of ECCS in PWRS": Tlii s Information Notice (IN) addressed the degradation of the FSAR four-loop emergency core cooling systems (ECCS) injection flow rate if RHR crossover line vslves were closed. As documented in NRC:RI Inspec-tion Report 50-443/87-10, the licensee's Independent Safety Engineer-ing Group (ISEG) recommended that NHY Engineering perform an analysis, based upon Westinghouse Owners Group (WOG) data, which

would address the problems associated with the normal RHR shutdown

. cooling configuration during Mode 4 operation with a closed crossover valv Due to a delay in the WCG response to IN 87-01, the licensee's analysis has yet to be performe In a licensee memo dated July 14, 1988, a tammitment to initiate the WOG solution to the IN 87-01

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generic problem was mad If the current RHR shutdown cooling pro-cedures are not in accordance with the new solution, then they will be revised with appropriate correction As a result, this item remains ope Additional inspection effort was devoted to the follow-up of operator training in this area. Discussions with on-shift operatars revealed that they were familiar with the problems associated with degraded emergency core coolinJ systems ECCS operability and the closure of the RHR crossover line valves. Procedures which address the valve alignments for shutdown cooling (051000.01, 051013.03, and 0S1013.04)

and RHR technical pecification surveillance testing (0X1413.01) were reviewed and found to have incorporated the appropriate cautions /

statements regarding this problem.

2 (Closed) Unresolved Item 87-16-03: Ope _ ration of the Startup F_eed-on an Emergency Bus.

1 water Pumptest oTerational (SUFP)_/T-39.2, Toss oF0f f site Power with SI," reviews ofBase procedure 0X1426.02, "C/G 1A 18 Month Operability Surveillance," and subsequent discussions with both the licensee and NRR, two concerns regarding the operation and testing of the SUFP on emergency bus E5 l were identified, The licensee's corrective action for the operations concerns was to

, revise the applicable emergency operating procedures to ensure that i

operators would verify that emergency diesel generator (EOG) 1A would have adequate load carrying capability before loading the SUFP on to bus E5. This was verified by a review of the following procedures:

E-0 ES-0,1, E-3, FR-H.1, ECA-0.1, and ECA- In each of these procedures, the maximum allowable EDG 1A load of 3600 kW is addressed as either a caution on the summary page or has been incorporated into the procedure as a required step / action.

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The testing concern for EDG 1A and the SUFP loading will be addressed by interpreting Technical Specification 4.8.1.1.2 in accordance with a proposed NRC Generic letter, which clarifies the description of auto-connected loads. The inspector had no further questions in this area and considers this item to be close d. (Closed) Open Item 87-22-01: Siren Modifications. This item indi-cated that the sirens located in Rye, New Hampshire required modified

antenna ground planes and that several addi.ional sirens required ,

application of the anti-icing coatin The inspector reviewed the

) repetitive task sheets for the antenna change outs and application of

anti-icing coatings for seven Rye sirens.

Based upon the above, this item is close ,

e. (Closed) Open Item 88-O'9-01: TSC/ EOF Technical Support. The inspec-tor participated in the NRC evaluation tean, which observed the 1983 ,

Annual Graded EP Exercise on June 27-28, 1988, as documented in

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NRC:RI Inspection Report 50-443/88-0 Several open items were generated concerning exercise weaknesses. The following presents amplification and clarification of certain technical concerns iden-

, tified in paragraph 3.1 of the above repor Inspection Report 50-443/88-09 stated,

" The Technical Support Center (TSC) and Emergency Operations Facility (EOF) staff displayed questionable engineering judge-1 ment and/or did not recognize or address technical concerns j (50-443/83-08[9]-01)."

Several issues addressed below were cited as examples. Overall engi- !

neering judgement displayed in both the TSC and EOF was adequate,

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however, the following activities were noted to be isolated areas of weakness which were intended to be addressed by the license In

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! follow-up subsequent to the exercise with licensee technical support, operations and emergency preparedness staff, the following additional

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inspector follow-up iten 88-09-01 is dcscribed individually belo ,

(1) "Efforts continued to restore the emergency feedwater pump l (EFW) af ter a large break LOCA" ,

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The licensee correctly stated that the EFW pump would be required to operate to support steam generator cooldown in *

the recovery phase and continued repair efforts were pru-dent. The inspector agrees and determined that the stated activity did not detract from the overall recovery ef fort, !

nor did it diminish other high priority recovery action in progross or planned, and that TSC judgments were made with long-term recovery in min l

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(2) "A questionable fix for the containment building spray (CBS) system" The inspector met with the Technical Support Manager and a Technical Support Engineer and discussed the rationale behind the corrective action taker to rig an alternative water source for the CBS system. Although the capability of the proposed modification to the system to reduce con-tainment pressure was never proven due to the eventual repair of a CBS pump, the inspector determined, based on this additional information, that the engineering judgment and methodology involved in the proposed system and opera-ting procedure changes were acceptable. The licensee actions were appropriate since this fix was considered to be a "last resort" measure aftcr all prudent and subsequent extraordinary reasures had failed to provide containment spray by other means due to additional scenario controller interventio Additionally, the licensee had previously determined that the composition of the present TSC engineering staff, while adequate, could be enhanced by providing an augmented staff roste NHY has committed to implement this initiativ (3) "A lack of effort to locate and isolate the release path" This apparent lack of effort was the resu,t of licensee decision; not to pursue entry into the containment encirsure due to high radiation level Discussion with the licensee confirmed that indirect measures, such as remote temperature, pressure and sump level indications, were taken in a timely fashion to provide an alternate assessment of potential leakage paths. The inspector was unaware of these activities during the drill. The licensee decision to postpone entry into the containment enclosure was intentional, based upon other recovery ef forts associ-ated with depressuring the containment. Restoration of a CBS pump was imminent and activation of this system would have stopped the releas CBS restoration was subse-quently, and repeatedly, delayed by controller intervention so that the operators were prevented from affecting repair The licensee decisions in this regard were appropriat (4) "No effort was noted to bisdown ste.m generators (S/G) to lessen the heat load in containment" This comment implied that S/G blowdown was appropriat The actual concern was that a step in the emergency proced-ure required the S/G to be depressurized. This step was not performed because the TSC staff was unsure of the integrity

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of the S/G tubes because no sample was available due to blowdown system isolation. This TSC staff concern was expressed to the inspector when he questioned them during the exercis The NRC position in this area is that improved guidance to the operator may be warranted and >

should be evaluated, however the decision not to vent or i blowdown the S/Gs without sampling appears to have been  :

reasonable and appropriat (5) "Neither the E0F or TSC staff questioned a release of f greater than 7C30 curies per second with only clad damage and no core uncovery" The inspector reviewed the player and controller logs for selected TSC, E0F and engineering support center (ESC) '

staff. These logs revealed that several staff members did question and/or comment on the mismatch between the reactor coolant activity and the release rat Subsequent discussions with the TSC and EOF controllers and players also indicated that they were aware of this mismatc In actuality, the ESC staff made very accurate core damage assessments based upon the data supplied by the TSC. The E0F dose assessment staff made accurate dose projections

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based upon the release rate, as well as correlation of field data to the release rate. A review of previous drill 4 comments, as well as the player instruction for this exer-cise, indicated that this level of activity is recognized ,

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'o be an unrealistic number, which is required to provide

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the offsite dose rates necessary to exercise the entire ,

emergency planning zone. The technical staf f s had repeat-

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edly identified and questioned these mismatches in previous drills and were told by the controllers that this high

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release rate was necessary to test the off-site plans, and that they should not challenge the dat Although NRC review of the specific scenario used for the ,

exercise was acceptable, the above described problem indi-

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cates that the licencee should place more effort in

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developing exercise scenarios where core damage and release l rates are consistent.

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tion confirmed that the TSC/ EOF staff possesses adequate capabil-ities to protect public health and safety. This open item is con-sidered close l

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. (Closed) Open Item 88-09-02: TSC/OSC Multiple Access Points. This item indicated that the TSC and Operational Support Center (OSC) have multiple entrances and exits tt t are not controlled. As a result, contamination controls were inettective at times as personnel entered t without frisking and it couldn't be determined if continuous account-ability was, or could be, maintaine l The TSC has a main entrance where contamination controls and initial and continuous accountability is established and maintained. The TSC also has a back entrance which is not locked. Although this entrance is not normally used, the licensee agrees that it could be used, in effect bypassing the controls established at the main entrance. The licensee has agreed to change ER 3.1, "Technical Support Center Operations", to control access through this entrance as well as move the main entrance control The OSC also has multiple entrance However, this was a condition that was artif f:ial to the exercis At the time of the exercise, the radiological control area (RCA) had not been implemented at the

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station. The licensee procedures clearly show that when the RCA is I

implemented there will be only one entrance into the OSC from the RCA, l

The inspector noted that the licensee established and maintained habitability throughout the exercise. Althougn some minor contamina-tion could have occurred in the TSC, it is clear it would have been <

prcmptly recognized and would not have adversely impacted TSC operations.

I (Closed) Open Item 83-09-02: Departing Shift Dosimetry. This item indicated that no apparent consideration was given to the departing

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first shift to account for possible dose when leaving the plant l during the release, as they were not given dosiretr ;

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A subsequent review of the TSC logs, as well as discussions with TSC and OSC staf f, indicated that consideration was given to the depart- ,

ing shif Contamination and radiation surveys were ordered and

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taken. Results indicated all areas were below backgroun Because of this and the current wind direction, the TSC staff elected to al-l low the departing shift to exit the site without dosimetr [

Based upon the above review, this item is closed.

l (Closed) Open Iten 83-09-04: Media Center Responses to the Press l Irlqui rie s . This item concerne'd the licensee representa W e 5 responses to some questions in the Media Center which were not con-

.sidered adequat The licensee has agreed that these questions were not fully answered. Although the answers given were current, they did not have enough substance. The licensee has agreed to upgrade the

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training for the Media Center spokesperson, including more informa-tion on the NRC Incident Response Team capabilities and roles. Addi-tionally, during a real emergency, federal spokespersons would have been available to provide clarification as the need arose. This item is close . (Closed) Unresolved Item 88-02-01: Acr.umul a to r Isolation Valve Actuation Logic Question In meetings with licensee operations and engineering representatives in June and August, 1988, the resident inspectors discussed questions regarding the "maintain CLOSE0" switch, its function and design feature Licensee personnel ade-quately addressed the compliance of the current design with Institute of Electrical and Electronic Engineers (IEEE) Standard 279 and IE ;

Bulletin No. 80-06 guidance. Additionally, the inspector reviewed system test packages for the wiring veri?ication and functional checks (reference: general test procedure, GT-E-21) of the subject valve circuitry to confirm the opening of the accumulator isolation valves upon receipt of a safety injection signal with the switch in the "maintain CLOSE" positio '

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The licensee stated that the FSAR described a valve capability for future operational testing which, while currently available, was prohibited from use by technical specification requirement The inspector evaluated this position and determined that the governing administrative and LCO controls were adequate to prevent safety prob-lems during routine operation and shutdown activities. Only specific plant transitional situations and mode changes (particularly entry into Mode 3) represent potential problem area It was noted that the Westinghouse Owners Group is evaluating accident scenarios in Mode 3 below 1000 psig reactor coolant system (RCS) pressure and in i Mode 4 on a generic design basi .

In order to address the inspector's specific concerns regarding the adequacy of current orocedures/ drawings and of future operational i controls if technical specification requirements are revised to allow accumulator isolation valve closure in higher modes for testing in accordance with FSAR provisions, the licensee implemented the fol-lowing actions:

(1) Issued Revision 10 to the "SI-Accumulator Isolation Valves Logic Diagram", 1-NHY-503907, to delineate the pressure setpoint above which an alarm is actuated if the valve is not fully open.

] (2) Initiated resisions to the affected alarm response procedures to correct the recommended action references relative to the

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proper RCS pressure setting at the safety injection (SI) unblock j pressure.

(3) Recommended revision to the SI system description, SC-NAH/

NCH-284, Foreign Print No. 52005, for the accumulator tank iso-i lation valves discussing valve closure af ter resatting an SI signal with the valve controls in a "maintain CLOSE0" positio .

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The inspector reviewed licensee engineering memoranda, including one issued by the Yankee Atomic Electric Company, Nuclear Services Division, on the accumulator isolation valve actuation logic and considered the adequacy of the current Emergency Response Procedures to the SI valve respense design, including SI signal rese No problems with existing controls were identifie The inspector determined that th9 questions on the subject system design and controls nave been adequately addressed and that the licensee has taken steps to ensure the continued adequacy of design control if the technical specifications are amended to incorporate the full accumulator desigr. features discussed in the FSA This unresolved item is considered close J. (0 pen) Open Item 88-06-01: Non-Class IE Loads Powered from Class 1E S o u rc_e s . This item was origi ally opened to resolve the is:ue sur; rounding the tachometer on the emergency feedwater cump (EFW) tur-bine. Subsequently the NRC concern has been expanded to include the entire program for design, identification and testing of non-class 1E loads powered off of class IE source (1) Background NRC:RI Inspection Report 50-443/88-06 described a non-class 1E circuit (EFW tachometer) which was not included in the NHY Technical Requirements Manual (NYTR) list of devices to be tested per technical specifications (T.S.).

The T.S. involved in this issue consists of two parts which deal with containment penetration conductor overcurrent protective devices and protective devices for class 1E power sources con-nected to non-class 1E circuit This discussion concerns only the class IE power s;urces connected to non-class 1E circuit This specification states that each protective device for class

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IE power sources connected to non-class 1E circuits shall be operable in Modes 1-6, With one or more of the protective devices inoperable, the cir-cuit mv;* be de-energized by tripping the circuit breaker or

racking out ue *emoving the inoperable device within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, j In addition, the above status must be verified every seven days thereafter. The surveillance requirements necessary to declare operability include periodic testing, inspection and preventive maintenance of the device. The list of protective devices to be

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tested per T.S. Surveillance Requirement 4.3.4.2 were incorpor-l ated into NYTR Table 16.3-10 (Technical Requiremer.t 15) under

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the T.S. Improvement Progra .

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The NHY Systems Support Department Manager reported on May 2, 1988 that his review of the circuit indicated that the tach-ometer for the turbine-driven emergency feedwater purrp was a non-class 1E load connected to safety-related bus E5 via 120 vac ;

motor control center E515 distribution panel F.3E, circuit t Request for engineering services (RES)88-226 was w-itten on May 6, 1988 to determi.ie wnether this circuit should be included in Table 16.3-10 of the NYTR. A station information report (SIR) was initiated on July 26, 1988 to document this situation and further clarify the reporting requirement Licensee event report (LER)88-002 and its supplement riocument previous instances where other ncn-class 1E circuits were omitted from Table 16.3-10 of the NYTR. Additional NRC inspection of this previous LER may be found in NRC:R1 Inspection Reports 50-443/

88-06, paragraph Sc and 50-443/88-07, paragraph Licensee evaluation of this issue was conducted as an SIR fol-low-up. Engineerin3 review of calculation 9763-3-E0-00-46-F,

"Failure of non-class 1E Loads on class 1E Buses" revealed several additional loads requiring immediate resolution to en-sure compliance with the T.S. As of the end of this reporting period temporary modifications had been made te nearly all of those circuits and a permanent design change is in progres (2) Chronology January 1988 Licensee review indicates that the supply breaker to inverter 28 off of unit substation E51 is not on the list in the NYT February 1988 Following evaluation of preoperational testing previously conducted on the breaker, it is deter-mined that the breaker must be teste It fails the test, is repaired and the system is restored to operable statu March 1988 LER 88-002 is submitted indicating that a review of all unit substations reveals that the above finding is an isolated cas April 1988 ine inspector providos a copy of a January,1988 daily report frcm another nuclear facility about the power supply to the auxiliary feedwater pump tachometer which is similar to the above findin ,

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May 1988 Request for engineeriig review of Seabrook EFW pump turbine tachometer is issued by NHY (RES88-226). The licensee determines that the EFW pump tachometer is not class 1E. The tachometer circuit is not disconnected electrically from its 1E power source as required by the T.S. action statemen Licensee discovers the breakers between 2 pairs of unit substations are also not on NYTR lis Substation tie breakers are added to list. Sup-plement I to LER 88-002 issue July 1988 Licensee review of the relevant engineering cal-culation determines that two separate problems exist:

(1) Coo-dination of the tie breakers in the unit substations (2) EPd tachometer circuit Circuit breaker for EPd pump is opened per after discussion with the inspecto August 1938 Continued review of calculations indicate that trains "A" and "B" have additional circuits which are not analyzed and are required to be discon-nected per Temporary modifications are initiated so as to be completed prior to expira-tion of the 72-hour LC A permanent design change is in progres (3) Inspection The inspector held frequent discussions with the Technical Sup-port Vanager and Lead Technical Support Electrical Engineer con-cerniog progress of the analysis and installation of the tempor-aiv .todifications. A licensee event report will be submitte Prwiiminary fMC review of the train "B" temporary modifications revealed no concern (4) Findings Based on the above, the following issues remain unresolved:

(a) Adequacy of the original determination of which components were to be incorporated into the NYTR lis _ - _ _

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(b) Licensee actions taken upon discovery of the non-class IE EFW tachometer powered from a class 1E bu (c) Reportability of the above findings in accordance with 10 CFR 50.7 An additional question that must be resolved concerning the NYTR is whether non-class IE loads which meet seismic design criteria may be omitted from the NYTR listing. Licensee and NRC activ-ities are ongoing and will be the subject of continuing evalua- l

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tio This item under expanded scope remains ope l t (Closed) Violation 88-06-02: Emergency _ Diesel Generator (EDG) Failure

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Reporting

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(1) Background. NRC:RI Inspection Report 50-443/88-06 described a t?ip of the train "B" emergency diesel generator which occurred on February 24, 1988. Open Item 88-06-02 was written to docu-ment NRC questions related to the reportability of this failer Based upon the NRC questions, NHY conducted a comprehensive review of the diesel generator logs and determined that seven

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i j failures had occurred since issuance of the zero power license in October 1996. The failures were analyzed and summarized in a

letter to the NRC (NYN-89102) dated July 22, 1938. The informa-tional requirements of T.S. 4.8.1.1.3 were addressed for the most recent failure on February 24, 19S3. Additionally, the six previous failures were reported to bring the record up to dat (2) ,iviremen The above T.S. is applicable in Modes 5 and >

survetilance Requirement 4.8.1.2 states that the required ac electrical power sources shall be demonstrated operable by per-

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formance of Specification 4.8.1.1.3. This surveillance specifi-cation states that all diesel generttor failures shall be reported to the Commission in a Special Report within 30 day ,

i (3) Findin2 None of the above f ailures were reported within the 16-day time frame required by T.S. 4.8.1.1.3 and this failure to ,

, report constitutes a violation of the Saabrook Technical

$pecifications (SE-06-02).

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(4) Licensee Corrective Actions. Licensee corrective actions as a

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result of this violation aid actior.5 to prevent recurrence were

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, provided to the NRC in letter NYN-8310 hHY reporting proced- .

ures have been revised to address EDG f ailure The station information reporting system will be utilized to ensure that appropriate post failure actions are taken.

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Based upon the above and appropriate licensee actions initiated on two recent diesel f ailures, the inspector considers this issue closed and no additional resronse is required.

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5. Licensee Reports (Closed) Construr: tion Deficiency Report (COR) 86-00-09: Veritrak/

Tobar Transmitters. NRC:RI inspection reports 50-443/87-24 and 88-06 both document the progress made in the installation of Rosemount transmitters to c.orrect this deficienc Design coordination report (OCR)86-340 was implemented to control the rework and complete the corrective action documented in the final 10 CFR 50.55(e) report to the NR During this inspection, the inspector examined the completed field installation of all 23 Rosemount transmitters in the Unit 1 contain-ment building. The rework associated with change authorization No. 7 to DCR 86-349 was checked and specific installation details (e.g. ,

compression fittings) were examined. The inspector also noted that the installed components were Rosemount Model 1154 transmitters, dif-ferent from the Model 1153 transmitters that have exhibited manuf ac-turing deficiencies at other nuclear power plant The inspector reviewed the DCR for calculations affecting instrument setpoints and determined that certain technical specification tabular data and limiting condition for operation setpoints require revisio ' he iiconsee submitted letters to the NRC dated May 27, July 8 and

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August 0, 1938 (NYN-83075, NYN-88091, and NYN-88109 respectively),

which discuss the methodology used in the Rosemount setpoint analysis and transmit the proposed techr.ical specification changes and a sup-plemental analysis of the relevant safety consideration The inspector reviewed these documents, noting consistency with the Westinghouse setpoint methodology (also discussed in NRC:RI inspec-tion report 50-443/87-24) and with the values calculated in DCR 86-34 The inspector's review of the proposed technical specifica-tion revision > vere discussed with NRR project and technical reviewer personne The inspector confirmed that system operability considerations will be adequately controlled by the proposed technical specification changes, that a license amendment has been requested and is being processed, and that the licensee has completed all corrective actions relevant to its final 10 CFR 50.55(e) report. Adequate consideration of the level measurement error due to reference leg heatup for the steam generator level reactor trip and emergency feedsater actuation setpoints was also verified to have been included in the Rosemount data calculations. A licensee request (NYN-88082) dated June 9,1938, regarding the need for operator action in response to level measure-ment errors also has been transmitted to NRR for revie All corrective measures commitments have been completed and no fur-ther action is req"ired cf the licensee at this tim This CCR is considered close _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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< (Closed) 10 CFR 21 Report (87-88-04): Gould Relay Failure The failure of Seabrook-specific modified Telemecanique J-10 relays in April and August, 198'/ resulted in a licensee investigation into the number and use of relays installed at Seabrook Station. NHY engi-neering evaluation 88-001, "J-10 Relay System Evaluation", concluded that plant operation with the defective relays in service was accept-able during Modes 5-6, but was unacceptable during Modes 1- Of the 112 J-10 relays which were found to be in service in the

! plant, 57 were installed in safety-related applications. These were replaced in accordance with DCR-87-39 . Because of the unique voltage requirements specified for the original

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relays, Telemecanique was unable to ensure a qualified 4r y3ar opera-i tional design life for the replacement relays. Analysis showed that i

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a design life of only 4.3 years could be guaranteed. This reduction in design life resulted in the generation of maintenance procedure j MS0514.17, "Telemecanique J-10 Relay Magnet Block Replacement". This

procedure provides the instructions necessary to change out all j safety-related J-10 relays prior to the end of their design life.

, To verify that these changes were made, the inspector conducted a field walkdown of selected replaced relays with the cognizant tech-

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nical support engineer. This sampling included the following relays:

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System Relav Work Package i CBA E42/9a-3-3 87W003095 CBA E42/9a-3-4 87W003096 PCCW RYY-2192-1L, 2L, 3L 87 WOO 3132, 8133, 8134

, PCCW RYY-2292-1L, 2L, 3L 87W003135, 8136, 8137

. EAH E3E/3-R1 $7 WOO 3112 l'

EAH E3F/Sa-R2 87 WOO 3113 EPA RBC7a 87 WOO 3114 All of the above listed relays were verified to have been replace A document review of the above listed work packages was performed.

! No discrepancies were identifie The inspector has nc further

! questions in this area and considers this item to be closed.

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f (Closed) 10 CFR 21 Report (87-88-03): Service '(ate _r_ System Valve

Liners and Seats. A generic problem was icentified with the cil-covery in May, 1987 of the premature deterioration of the liner / seats of certain butterfly valves supplied by Fischer Contiels. The sub-

! ject valves, installed in the service water system, had been modified l previously as corrective action in .:ccrdance with a 10 CFR 50.55(e)

i report (85-00-13) in which liner detachment problems wera noted. The i root cause of the most recent deterioration problem was attributed to l inadequacies in the modifie d seat design and in the elastomer liner bonding process applied to correct the original detachment proble ________________

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This issue was first opened in NRC:RI inspection report 50-443/87-13 and was reviewed by an NRC:RI specialist inspector, as discussed in report 50-443/87-1 The licensee submitted a 10 CFR 21 report (NYN-87091) to Region I on July 28, 1987. The inspector reviewed the ,

licensee's "Summary Report on Service Water System Valves", dated July 29,1987, noting discussion of both short term and long term corrective action programs. With respect to the short term, NRC inspectors, over the past year, have witnessed licensee implemanta-tion of a repair and test program for the subject valve Twenty-

, eight valves were modified with an improved valve liner / seat design t

which has increased the liner thickness to preclude deterioration (reference: DCR 87-249). Also, the instClation of design modifica-tions (D R's87-315 and 87-401) to the piping downstream of certain of the valves was inspecte These changes alloweet for the subject valves, previously utiliied in throttling applications, to be posi-tioned either fully opened or closed, thus reducing the potential for i

! future deterioration. By July, 1988, all the design changes asso-ciated with the servica water valve rework and system redesign had been complete i Longer term corrsctive action consists primarily of a monitoring ,

program to ensure that short term corrective action has been effec- l tiv Tne licensee plans to conduct an inspection of four of the modified valves, including two that were changed from a throttling application, during the first refueling outage. The inspector verif-ied that this activity has been formally noted in the licensee's

integrated ccmi tment t ra c k i .9 g system (action no. RED 2082). The l l inspector also reviewed scheduled raintenance data sheets which pre- '

i scribe the insp2ction of two additional codified valves for seal /

liner damag Such checks will occur each time the servics water strain 1rs in proximity to the valves are removed for cleanirg, at a -

frw uency of about every two months or whenever differential pressure indications dictat Also the licensee has fabricated test coupons of the modified elasto er liner material burded to valve-like meta l

These test coupons have been immersed in the circulating nater pump house basin to ecnitor the effect of seawater on both the elastomer and the bonding process. The inspector examined two work requests describing the removal Of the test coupons to be conduc' ed in the

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latter part of 1953 for transmittal to the elastomer manufacturer, Belzona Molecular Laboratory, for pull testin The insepctor noted that both the . nrt and long term corrective  !

, actions taken or planned by the licensee in response to this design l

) deficiency were consistent with the 10 CFR 21 report submitted to the '

NRC and with the discussion of the deficiency documented in NRC:RI inspection report 50-443/87-18. Short term corrective actions have been co pleted and icng term corrective actions are scheduled and j being tracke The inspector has no further questions at this tire

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, _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . __ ______

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with respect to the licensee evaluation of the problem, the testing conducted tv effect a workable design solution, the actual repairs or the plans for future monitoring of the valves to check for liner deterioration. The licensee's overall approach to this problem from a technical standpoint has been methodical and comprehensive. The NRC has been kept informed of new developments and licensee plans. This 10 CFR 21 Report is considered close Station Information Resorts. Licensee station information reports (ilR) are used to internalTy report and evaluate operational events that may require further investigation, notification to a regulatory agency or require root cause analysis. Licensee Event Reports and 10 CFR 21 reports normally originate with an SIR. The reports discussed below were reviewed for compliance with the implementing instructio Supervisory, regulatory. services, r anagement and SORC reviews were verified. Also examined were the technical evaluation of each event, root cause analysis and recommendatio (1) >IR 88-01_0: On January 15, 1988 the train "A" amergency diesel generator (EDG) was unloaded and shutdown during a post mainten-ante test because of a lif ting relief valve in the auxiliary cooling water system. As a result of this SIR several minor design changes were instituted to improve engine reliability and performance. Tne inspectors discussed these modifications with the Systems Support Manager and the cognizant Lcad Systems Enginee (2) SIR 88-054: This SIR was initiated to investigate the root cause of a mispositioned circuit breaker in the service water syste The licensee evaluation revealed minor administrative work control defic;encies and some human factors improvements which should be made in the labeling of the af fected motor con-trol center , NRC Bulletins and Information Notices (Closed) NRC Bulletin 87-02, Supplements 1 and 2: Fastener Testing to Determine Conformance with Applicable Material Speci ficati on ~

As documented in NRCTRI inspection report 50-443/37-26. Bul~1etin 87-02 was closed based upon the conduct of testing and submittal of test results by the licenset to the NRC. The inspector assessed all the actions taken by the licensee in response to this bulletin and determined that they were both complete and adequat . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _

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i Subsequently, the NRC issued Supplements 1 and 2 to NRC Bulletin 87-02, requesting, and then clarifying the request for, additional information on the suppliers and nianufacturers from which the subject fasteners may have been purchase On July 21, 1983, the licensee i responded to the supplemental requests by letter (NYN-88099) to the NR Enclosed with the letter were a list of approved vendors who supplied or may have supplied ferrous fasteners suitable for safety-related applications and a list of vendors who supplied commercial I grade f astener The licensee response also discussed the basis for 3 compilation of the lists and a committrent to notify the NRC of any additional suppliers or manufacturers identified by on going procure-ment record review ; The inspector reviewed the information submitted in response to Sup-j plements 1 and 2 to NRC Bulletin 87-0 No questions or concerns

, regarding this submittal were identifie This bulletin remains closed for inspection purposes, b. (Closed) NRC Bulletin 83-05, with Supplements 1 and 2: knconform-ino Materials Su and West Mrsey_pplied Manufacturin_g by _Pipino_ Supplie g mpany Inc. at Folsom, at Willianstown, New Jersey New Jersey.

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NHY responded to NhC BuiTe~ tin 88-05_byletter (NYN-88114) on August 25, 193 This letter included the detailed results of the licensee effort to determine the impact of suspect materials at Seabroek. The NHY program consisted of the following:

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Identification of af fected materials in safety related systems I -

Verifying acceptability of installed materials

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Reporting to the NRC in accordance with the requirements of the l

bulletin j A total of 369 flanges ard fittings were identified in safety related i system A test program was developed to measure the hardness of carbon steel items and ferite content in stainless steel item Licensee representatives participated in an Electric Power Research j Institute workshop on the use of the Equotip test equipmen NHY

o .lity control (QC) inspectors performed the fiold testing of each

. flange and fitting. The data sheets were evaluated by the cognizant J quality assurance (QA) engiree Om July 15, 1988 in the service

) water cooling tower, the inspe: tor observad field hardness tasting of

the service water system flanges. The testing was corducted in ac-

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cordance with procedure NHY-EHT-1, "Equotip Hardness Testing" (Revis-ion 01, Change 01). The inspector reviewed the procedure and work request SSW3339 and verified that licensee OC personnel were know-ledgeable concerning both the procedure and test equipment.

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Independent measurements were also performed on separate pieces of '

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, suspect material by J. Dirats and Co. and Bechtel Corporation to

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confirm the Equotip test results. Additionally, test results were sent to the Nuclear Mant.gement and Resources Council (NUMARC) for generic industry data compilation and analysis. Of the 369 flanges and fitting tested at Seabrook, 30 were found to be below the minimum Brinell hardness value of 137. This is the minimum value specified  ;

j in the American Society of Mechanical Engineers (ASME) material '

specification SA-10 The 30 fittings were individually evaluated ,

and found to exceed existing tensile strength requirements in accord-  :

ance with the ASME code. The evaluation demonstrated the inherent '

l conservatism of the code as well as the correlation between hardness l

, and tensile strengt NHY made seven calls to the NRC Operations i 1 Center over the course of the testing as required by the bulleti !

I These non-emergency notifications were part of the 48-hour reporting  :

requirements that were subsequently discontinued by the issuance of i Supplement 2 to the bulletin, r

Throughout the course of the test process, the inspector maintained close liaison with licensee OA/0C inspectors, engineers and managers.

i The methodology employed in identifying, testing and analyzing the suspect fittings was labor intensiv The licensee aevoted adequate

! researces to ensure timely completio The two shift testing sched-

! ule was particularly rigorous and the total support of NHY engineer-l ing and quality assurance departments were in evidenc Additional [

NRC Headqua*ters review of this bulletin may occur as a result of j generic evaluation of the PSI /WJM concer For inspection purposes, l this bulletin is closed.

! c. NRC Information Notice 88-46 and Supplement 1: Licensee Report of .

I Defective Refurbished Circuit Breakers. This Information Notice (IN)  !'

l describes discovery by another utility that certain non-safety re-i lated circuit breakers manufactured by the Square D Company were i actually refurbished equipment rather than new stoc It has been ,

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determined that certain suppliers were refurbishing components and '

re-labeling them as new equipment. The licensee is conducting its I own inspection to determine what effect, if any, this IN may have on  ;

l Seabroo During a visit to the facility on August 16,1988, the i D',.ector of the NRC Of fice of Nuclear Reactor Regulation discussed l

this issue with members uf the licensee inventory and material l

requirements department ^

I t The inspector will continue to fc' low this issue and its relationship

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to receipt inspection of comercial grade items as well as any future p additional NRC correspondence such as NRC Bulletins or additional IN ,

Su,S 'ements. For inspection purposes, this is an open ite ;

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. NRC Information Notice 83-25: Minimun Edge Distance for Expansion Anchor Bolts. An analysis of site specific data affecting the capacity factors of Hilti Kwik-Bolts installed at the minimum spec-ified distance from an unsupported concrete edge revealed safety factors greater than twice the allowable design loads, This analy-sis, accomplished by the Yankee Atomic Electric Company (YAEC) for the Seabrook Project, utilized conservative assumptions based upon Seabrook design criteria, Kwik-Bolt installation specifications and concrete compressive strength ti st data, Since no safety concern was identified, the YAEC recommendation *^ mnnart a Nuclear Management and Resources Council (NUMARC) initiative for generic industry-wide action on this issue was adopted, The inspector noted that a previous NRC unresolved item, 443/

82-03-07, had addressed consideration of the Kwik-Bolt shear cone interaction, including the influence of the spacing of anchors at concrete corners, As documented in NRC:RI inspection report 50-443/

85-25, testing was conducted at the Hilti Test Facility in Tulsa, Oklahoma to check the reduction in Kwik-Bolt capacities, in part, at outside corners, The results of such testing, while indicating a reduction in ultimate capacity, were acceptable when considered with respect to the overall expansion anchor design. The unresolved item was therefore closed, The inspector noted that the past testing of the Hilti Kwik-Bolts, while not accomplishtd specifically to address the 10 CFR 21 concerns ra' sed in IN 83-25, has confirmed the conservatism of the design, the acceptability of Seabrook site-specific applications and the assump-tions made by licensee engineering personnel in calculating design loading data, Thus the licensee positions that Kwik-Bolt installa-tions at Seabrook represent no immediate safety concern and that future reviews can be adequately handled through NU,tARC appear to be well founded, No violations were identifie This item is closed for inspection purposes, e, IE Information Notice 86-50: Jnadequate Testing to Detect Failures The inspect 3

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of Safety Related Pneumatic Component s_ or System reviewed internal licensee memoranda providing evidence of engineer-ing review and regulatory cognizance cf the subject information notice. The licensee ccntinues to evaluate their methods of air system and component testing and instrument air quality sampling in accordance with FSAR commitment The inspector confirmed that although no specific action is required

.by this information notice, the licensee appears to be investigating the applicability of the relevant safety issues and tracking regula-tory cemnitments and criteria accordingl No violations were identifie This item is closed for inspection purpose _-- _ - _ _ _ _ _ _ _ - - _ _ _ _ - _ _ -__ _ _ ____ _ __ ______-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _

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7. Maintenance / Surveillance i a 1 i OX 1456.81: Operability Test of ISI Valves. On July 22, 1988 a re-test of the mot".r operated suction isolation valve to the train  :

, "B" safety injection (SI) pump, CBS-V-53, was performed in accordance 3 with surveillance procedure 0X145631, "Operability Test of ISI '

, Valves". The test was completed under work request 88W2735 and con-l' sisted of the stroking of the valve to gather the required inservice testing (IST) valve stroke time dat The inspector observed the test locally at the valve in the residual heat removal vault. The results of this test were an opening time of 10.69 seconds and a closing time of 10.22 seconds. The maximum allowable stroke time was

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15 seconds for each direction. No violations were identified, a

! EX 1804.044: Safety and Relief Valve Setpoint Pressure Tes On j June D,1931 another nuclear facility reported problems associated j with setting main steam safety valve (MSSV) lift setpoints using 4 nitrogen. When these valves were subsequently lift tested with j steam, setpoint drift was noted. The inspector reviewed surveillance l procedure EX1804.044, "Safety and Relief Valve Setpoint Pressure Test" and verified that Seabrook MSSV's are presently tested in place l with system pressure 15-25% below valve set pressur An assist

) motor is used to provide the additional test pressure. Therefore the

! above described problems can not occur at Seabrook.

I i EX 1804.016: Diesel Generator Auxiliary C'ool ant System Quarterly -

l T e_s t . On May 13, 1958 the train "B" emergency diesef ger,erator (EDG)

, was returned to service following maintenance. Operability of the l ED3 is normally verified by four separate surveillance tests; engine i start, fuel oil transfer pump performance, cooling water and air start valve performance and auxiliary coolant performance. An admin-j ist"tive error resulted in declaring the EDG operable on May 16, j 193b prior to completion of the test un the auxiliary cooling system

(EX 1804.016). Station information report (SIR)88-048 was initiated l because of this occurrence. The SIR indicated that the root cause of 1 the problem was inadequate scheduling because of an error in the j Specification Appraisal computer program, The inspector reviewed i licensee corrective ar.tions which included adjustment of tr,e program model and had no further questions.

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IX 1630.921

SSPS Train "A" Actuation ~Looic Tes On August 19.

j 1988 the inspector witnessed portions of IW0epartment Surveillance i Procedure IX 1680.921, SSPS Train "A" Actuation Logic Test. The pur-i pose of the test is to functionally test the train "A" solid state

! protection system (SSPS) in accordance with technical specification i 4.3.1.1 and 4.3.2.1. The inspector witnessed selected steps concern-i ing reactor trip breaker operation locilly in the essential switch-

! gear room. Th3 inspector noted effective communications established

with the control room, the presence of a knowledgeable electrical

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quality control inspector and proper control exercised over the pro-cedure by the control room personne No violations were identifie _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______ _ ___ - _ _ _ _ _ - _ _ _ _

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. EX 1804.015: Diesel Generator 1B 18-Month Operability and Engi-

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neered Safeguards Pump and Valve Response time Testing Mode 5 Sur-veillance. This is a seven-event surveillance test which satisfies several train "B" Mode 5 technical specification surveillance re-

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quirements. The inspector observed portions of event three and event six. Event three involved an emergency diesel generator (EDG) start initiated by resetting the train "B" low steamline pressure safety

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injection ("S") actuation signal from the main control boar The inspector witnessed the diesel start to a standby idling condition and the starting of the train "B" emergency core cooling system

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(ECCS) pumps as well as feedwater isolation and main steam line iso-lation. The test was run twice because of high speed recorder prob-lems which were eventually correcte In all cases the plant

, responded as designed. Event six followed the 24-hour run of the i train "B" EDG and tested the ability of EDG 18 to start and lead upon concurrent loss of of f site power and an "S" signal and to verify that bus E6 s's Js its load. ECCS pump and valve response times were obtained and the EDG's ability to accept a cooling to*.<er actuation ("TA") s.gnal while loaded with auto connected loads was also ver-

ified, following successful service water system .ransfer to the

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cooling tower, the EDG's ability to accepe a large load rejection was

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tested by simultaneous 1-, tripping the cooling tower pump and charging pum The inspector noted that the control room operators and test director were intimately familiar with the procedure and expedit-iously performed the critical post safety injection steps required by i procedur The equipment also was verified to properly perform its intended function. No violations were identified.

. OX 1406.02: CBS Puma and Valve Ouarterly Test and 18 Month Remote I P~osition Indication. on~JW19, 1988 while perf orming surveiflance procedure OX 1406.02, "CBS Pump and Valve Quarterly Test and 18 l Month Remote Position Indication", about 5000 gallons of water was inadvertently transferred from the refueling water storage tank

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(RWST) to the reactor coolant system (RCS) via the residual heat l removal (RHR) system. The event occurred because valve CBS-V-2, the train "A" RWST to RHR ! solation valve was opened with RH-V-22 and RH-V-23, the train "A" RCS to RHR suction valves still opene The

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operator immediately realized that the lineup was incorrect and re-closed CBS-V-2.

i t NRC:RI Inspection Report 86-54 (paragraph 4.a) described a previous i similar event which occurred on September 5,1986 and describes the

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design bases for the syste Also addressed was the standard i Westinghouse design for interlocks in these valves and the NHY posi-

tion on how certain design features (alarms) would be added to pre-

vent recurrence of the September 5, 1936 event.

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l The inspector met with the Assistant Operations Manager and discussed several issues related to this event, The licensee's ongoing correc-

!. tive actions will be observed during a subsequent NRC inspectio Residual Heat Removal (RHR1 System

, NRC Region I Inspection Report 50-443/87-24 described a discrepancy in the dimensional gap between the train "B" RHR pump casing and impeller. The licensee subsequently disassembled the train "A" RHR pump and found a similar problem. The dimensional gaps were found to be 0.0235 inches and 0.025 inches for the train "B" and "A" pumps respectivel The manufacturer (Ingersoll-Rand) specifies a dia-tretrical clearance between 0.030 to 0.036 inches. Both pumps wearing

) rings were machined within specification and the pumps restored to i service, j On March 13, 1938 the inspector observed the clearance measurements

. made on the Unit 2 RHR pumps. These cumps were never installed in

, Unit 2 and were transported from storage to the Unit I turbine J building for disassembl The inspector noted appropriate quality

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control hold points in the procedur Both quality control and

! maintenance personnel were considered to be knowledgeble f n their

! tasks. The Unit 2 clearances as measured were found to be within specificatio The 'icensee conducted an evaluation of this technical issue pursuant

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to 10 CFR 21. Engineering evaluation E3-016 concluded tFat given the i "as found" dimensions under design thermal and seismic conditions,

! pump damage would not have occurred and therefore, a substantial t

safety hazard did not exis This condit* a was therefore not reportable under 10 CFR 21.

1 The licensee conducted a detai'.ed review of all relevant documents

, to determine whether the wearing rings were modified in some way

during the construction or startup phase The NHY effort consisted

of a review of installation and work records and a review of spare i part receipt and inventory record 'ngersoll-Rand documents indi-

! cated that the clearances were within .pecification when shipped from their facilit Construction and maintenance records revealed no modifications or replacements were ever performed on the wearing j ring The cause of the out of tolerance condition could oot be 1 identified even though the records check was extremely detailed and l the quality cf the records was found to be acceptable. The licensee

concluded that all available prudent action had been taken and ttere-fore considers the issue closed. The inspector discussed the results

] of the engineering evaluation with the Manager of Engineering an3 the

Lead Mechanical Engineer and had no further questions.

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t I Design Changes and Modifications Post Accident Samplin In order to meet the require-i ments oGOGlRif37?giMI _ System (PASS).

Action Plan Requirements", (Item II.B.3), a

] PASS was installed at Seabrook. During hot functional testing, dif-i ficulty was experienced in obtaining consistent sample results be-1 cause of 1.9 dequate sample temperature control. As a result, design coordination report (DCR)88-081 was generated to add an additional sample cooler to the system. The inspector reviewed DCR 88-081, as

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well as its DCR implementatica plan, and made frequent field inspec-

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tions of work in progress with special emphasis in the piping sup-ports in the primary auxiliary building (PAB). Although the primary component cooling water lines which cool the new heat exchanger are 3 not safety related, they are constructed to seismic criteria due to

! the design requirenents of the PA The inspector had discussions i with the Systems Engineering Supervisor concerning the identification

of seismic /non-seismic class breaks in relation to licensee commit-i cents documented in NRC

RI Inspection Report 50-443/86-14. Field l inspection of piping and pipe supports revealed no violations of NRC i requirements. Completion of pre-operational testing on the PASS l requires the plant to be hot and is scheduled for accomplishment in the heatup prior to initial criticality. Actual testing of the PAS $

i will be ths subject of future NRC inspection to close out TMI Item l II.B.3.

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j Sacondary._ Component Cooling Water System (1) _B_a cig round . The secondary component cooling water (SCCW) system provides cooling water to non-safety related secondary loads in the turbine buildin Typical cooling loads are the air com-l pressors and condensate pu.mp air and oil cooler The system

includes three 50% capacity each contrifugal pumps and two 100%

capacity each large horizontal heat exchangers. The heat ex-changer shells and tube sheets are clad with 90-10 copper

! nickel. All other carbon steel inner substances are lined with

! neopren The tubes are 90-10 copper nicke These heat I enhangers are cooled by a non-safety related leg of the service

) water (SW) system.

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System inspections in 1986 and 1937 revealed significant tube

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corrosion due to low fluid velocities at low flow.

(2) Licensee Evaluation and Corrective Action. The N4Y engineering

department prepared engineering evaluation 88-04 in February, j 1933 which proposed several solutions including installation of j low flow heat exchangers for use during low heat load cond4 tion, j This would allow the main heat exchangers to be placed in layup

when not in use. Design coordination report (DCR) 8B-033 was

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i initiated to add two additional low flow heat exchange's to the l SW/SCCW systems. The heat exchangers were procured from exist- ,

ing stock as they are the original Unit 2 air removal heat exhangers. Once the new auxiliary heat exchangers (SCC-E-185A, i B) are installed, the main heat exchangers (SCC-E-29A,B) may be  ;

removed and reworked or replaced with the Unit 2 cooler :

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(3) Inspectio Despite the fact that this system is not safety '

related, this design change is of general NRC interest because j of its relationship to heat exchanger degradation in primary i systems as well as aeneral workmanship and work control through-out the plant. Th inspartor reviewed engineering evaluation 88-04 and DCR 88-088 and maoe frequent inspections of the work- i sit On July 22, 1988, the inspector identified a section of drain l

piping which had been lut off the main SCCW line in preparation i

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for weldolet installation. The lina contained valve SCC-V-344 and a tubing connection for chemistry corrosion monitoring. The  ;

above valve was still caution tagged and the tubing fittings

were identifisd as "Temporary Modification #10-Other". The L

inspector discussed this activity with the shift operators and

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Assistant Operations Manager. The inspector stated that removal ,

l of a caution tagged valve and temporarily nodified assembly appeared to violate station procedures concerning equipment  ;

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tagging and temporary modification s. Saintenance Procedure MA  !

! 4.2, Revision 7, "Equipment Tagging and Isolation" states, "No

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person shall physically remove any equiprnent that is tagged '

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"DANGER / CAUTION". Maintenance Procedure MA 4.3, Revision 7

"Temporary Modifications" indicates that changes to temporary modifications be re-routed with appropriate notations, initial-  !

led and dated by all reviewers or a new temporary modification l be prepared. In light of the nan-safety related nature of this modification activity, no vioi ltion of NRC regulations existed, i

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however, it is noted that corrective action for violation

87-20-01 that occurred in July, 1957, did not prevent recurrence l l of a similar although significantly less serious situatio It !

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15 also noted that anothar related occurrence was reported in f l station information report $7-108 in November,1937, i

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j (4) Conclusien It appears that additional attention is warranted

in this area especially with respect to temporary modification t

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control. These modifications are clearly identified and removal i or modification requires similar procedural controls as instal-  ;

lation. This area will be +.he subject of continuing NRC inspec-  !

tion with respect to routine plant operations as well as readi-l ress for initial criticality.

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9. Allegation Review

i As documented in NRC:RI inspection report 50-443/88-07, a written response ,

on the licensee's investigation by its Employee Allegation Resolution }

(EAR) program personnel of five separate allegations was requested. By  !

letter (NYN-88116) dated August 29, 1988, the licensee responded with the determination that the subject allegations are either inaccurate or relate '

to issues which were identified and dispositioned through internal quality programs. An enclosure to the licensee letter summarized each concern,

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q its review and the licensee conclusions.

4 The inspector reviewed the above letter, its enclosure and additional EAR j files and documents relating to the investigation of each allegation. As l was documented in the 88-07 inspection report, the inspector had pre-l viously conducted preliminary reviews of each allegation and performed l both field inspection and records research where appropriate. During this 3 inspection, the results of the licensee investigation were evaluated not only with regard to completeness and substantiating avidence, but also I with respect to the inspection data independently collected and checked by j the NRC. The following represent the conclusions reached for each of the

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five open allegation (a) U_ncertified pioing material supplied _by Boston Pipe.

The inspector reviewed UE&C audit and nonconformance reports (NCR)

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covering the Boston Pipe & Fittings Co. of Cambridge, Massachusetts

and the material supplied by this company for Seabrook Station. At

! lea,t one of the NCR's documented the receipt of fittings on site

, without certificatio Additionally, a Pullman Power Products NCR j was founw to have identified certain refrigeration system and support l material whicn lacked the appropriate documentation, l Each case of a nonconforming condition resulting frcm incomplete

] certification appeared to be properly dispositioned with evidence of

! completed corrective action and reinspection by quality assurance (QA) persennel. The inspector also noted that contractor receiving inspection reports required and recorded document verification anc traceability of the subject material as a requisite part of the inspection criteri Thus, while the existence of the noted NCR's indicates that this allegation may have some basis in fact, the identification and disposition of these problems by the licensee also indicates that the receipt inspection process was working effectively. The inspector found no evidence to suggest uncertified eaterial supplied by Boston Pipe had been installed in the plan .

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(b) Uncertified electrical equipment supplied by Massachusetts Gas and Electri The inspector checked a sample of purchase orders from the Massachusetts Gas & Electric Light Supply Company, noting that most

, wire and circuit breakers were procured for general jobsite temporary power and lighting. Despite the nonsafety-related use of such mate-

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rial, at least one NCR was issued to document the lack of proper '

material certification. The inspector also noted that both UE&C and Fischbar.h, the electrical installation contractor, conducted receiv- *

ing inspections which required document checks for certificates of compliance of the inspected material in accorcance with specification requirement As similarly discussed with allegation (a) above, the far.t that the i licensee quality programs require receipt inspection checks for pro-  !

per material certification and that NCR's have been issued when com-plete documentation was not available provides one measure of con-firmation that the material installed meets fabrication specifica-l tions. Even in the case of a nonsafety supplier like Massachusetts

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Gas and Electric, evidence of such QA checks are available in licen-

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i see record The regulatory requirements governing certificates of compliance, versus material certifications like mill test reports, i are not in conflict with the licensee position that the manufacturer [

provides the requisite certifying documentatio The inspector identified no information or facts that indicated that

, the Massachusetts Gas and Electric Light Supply Company had impro-

, perly certified material or that electrical components had been installed in the plant in applications for which they were unqual-if;ed.

] (c) Acceptable level installation of the reactor coelan_t_puyp_ The inspector reviewed Westinghouse and contractor records which

substantiated the licensee conclusion documented in the NYN-88116 ,

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J letter to the NR The Westinghouse Nuclear Service Division 1 "Procedure for Setting of Major NSSS Components", Revision 2, issued i in February, 1979, delineates the level criteria for the reactor I coolant pump The inspector checked the Pullman-Higgins installa-tion records for two reactor coolant pumps (RCP), including RCP-lC

which represented the component originally questioned in the tech-

) nical concern addressed in NRC:RI inspection report 50-443/87-07 (reference: UE&C engineering change authori:ation 03/1557A). For i each pu p, the inspector examined the "RCP-Volute Level Data Sheet -

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After Adjustment" and independently calculated the maximum level deviation. Although RCP-lC was slightly more of f-level than RCP-ID, both pu.?ps were measured to be level within the Westinghouse accept-ante criteri <

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Furthermore, the inspector noted that a Westinghouse memorandum issued in March, 1982 acknowledged the adjustment that was made to the RCP support and the resulting change in tne RCP volute main flange differential elevation. Westinghouse engineers approved the change at that tim The inspector reviewed additional evaluation of the RCP level concerns by the licensee corporate engineering-l staff to include recent Westinghouse studies on RCP "tilt" condi-tions. These newer studies appear to indicate that the original Westinghouse level criteria, which the Seabrook RCP's meet, are  ;

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conservativ l Therefore, with regard the question raised by this allegation, the inspector confirmed that the reactor coolant pumps have been instal-

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led and inspected to the Westinghouse design criteria and that I acceptable level conditions for each RCP were verified af ter imple-I mentatien of the engineering change which resulted in the reposition-

) ing of the base of one suppor (d) Weldolet in the emerynty feedwater (EFW) pump room with wrong taper and counterfeit identification numbe _

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j j Visual inspection of weldolets in the EFW pump room by an NRC i 1 inspector revealed no deficient or nonconferming condition Tne inspector also reviewed licensee nuclear quality group evaluations

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of elbolets and weldolets in the EFW pump room to ensure American '

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Society of Mechanical Engineers (ASME) code compliance, acceptable '

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markings and traceability and weld quality and taper. The licensee evaluation included documentation reviews, visual and ultrasonic

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thickness examinations, and inspection tracing of the scribed field ,

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1 marks to vendor documents which verify the quality and further

traceability of th3 installed component The licensee evaluation concluded that ASME code compliance had been confirn e The inspector checked the licensee's Thickness Data Sheet resulting r l from the ultrasonic testing field examinations and reviewed a sample j i of Dravo pipe fabrication sketches, establishing traceability of '

weldolet/elbolet field scribe marks to the heat number codes docu-mented in the manufacturers' mill test reports.

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! The acceptability of field conditions for a number of components, which might represent the subject of the stated allegation, was  !

i verified by independent NRC and li:ensee inspections. The inspector l l concluded that this allegation could not be substantiate '

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i (e) Qualification of an Authorized Nuclear Inspector (ANI) trainee2 i The inspector reviewed EAR records documenting licensee investigation of an allegation regarding the qualification of an ANI trainee and

.. , authority to conduct independent inspection As discussed in NRC:RI Inspection report 50-443/88-07, NRC inspection of a similar concern resulted in substantiation of certain of the facts, but in a conclusion that neither a noncompliance with the ASME Code, nor evidence of wrongdoing was identifie The EAR records confirmed that the allegation previously reviewed by the licensee involved the same ANI trainee that was the subject of the allegation raised to the NR The licensee investigation concluded that during the period of time from May to December,1935 when the subject ANI trainee was assigned to Seabrook, he performed assignments in accordance with his assigned training program. NRC inspector review of documents dating back to the 1985 time frame veri fied that qualified ANI's had evaluated and monitored the ANI trainee's training, progress and inspection wor While the facts surrounding this allegation may be true, both NRC and licensee reviews of the stated concerns have identified no impropriety with respect to the certification or conduct of work on the subject ANI trainee while at Seabrook Statio The five allegations listed as open in NRC:RI inspection report 50-443/

88-07 were addressed by the licensee in the response letter, NYN-8311 Independent NRC inspection of these issues prior to raising the questions with the licensee had identified no hardware problems or quality concern Subsequent licensee EAR investigation of the allegations concluded that the allegations had no substantive meri This inspection has included a review of those EAR investigation results and the process by which they were achieved. The inspector verified that licensee actions were compre-hensive relative to the information provided in the allegation The allegations generally either could not be substantiated, or represented issues with some factual basis, but with no adverse safety impac These five allegation issues are considered close . Training General Empoloyee Training NRC:RI Inspection Report 50-443/S7-16 discussed the topic of cheating on general employee training (GET) exams and the lack of written policy on cheatin During this inspection period this issue was re visite The inspector reviewed the GET examiration cover sheet which listed instructions to be read aloud by the instructor prior to

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the examination. These instructions specifically addressed the steps to be taken should suspected cheating occu Additionally, the f i inspector reviewed the draft of training procedure NT-7010. "Examina- !

tion Administration and Integrity" which also formalized the station >

j policy on cheating. The inspector determined that licencee follow-up i j actions this issue have been appropriate and had no further l

{ question l 3 i 1 Ojerator TraininJ  !

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l On July 20,193S, the inspector discussed the recent Nuclear Manage- f rnent and Resources Council rneeting on operator requalification ttst-

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4 ing, and the status of Institute of Nuclear power Operations (INPO)

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accreditation with the Training Manager. In the area of INPO accred- !

! itation, the licensee stated that an INPO programmatic inspection is !

I due to be performed in November of this yea ;

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) 11. Electrical Confiouration Control i s .

As documented in NRC:RI inspection report 50-443/8S-06, several engineer- l

! ing discrepancies and configuration control problems identified in the ;

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electrical area were resolved with the issuance of licensee engineering !

l evaluation 88-01 NRC open item 87-24-01 was therefore close l i

.' During this inspection, the inspector identifiec certain field conditions l

1 for which questions of electrical detail and adequacy were raised. Spec- ;

} ifically, electrical fire wrap requirements in ar.cordance with engineering i j change authorization 03/11295G. the protection of spared cable termina- l l tiens, the conformance of Sf6 switching station breaker alignment to the j

plant technical specifications, and the status of missing condolet covers j

, were all checked and found to be either acceptable or under work request i

] control. Additionsily, the inspector reviewed a quality assurance (CA) I

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J assessment (reference: CAIR E8-0597) of electrical design changes where l the potenLial for interface prCblems from engineering to constructien to j startup/ operational control appeared to be high. Only minor discrepancies i were identified as a result of this assessmen !

j i Another QA surveillance report 87-00%3 was reviewed with regard to the ;

ieplementation of work request activities in the cannibal 1:ation or Unit 2 '

! equipment and spare part components, including electrical iten The i Station procurement and Materials Manual (Chapter 5.5) delineates criteria

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for the control and docutent tracking of the cannibalication crociss. lhe j subject surveillance activity resulted in no adverse finding l

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With respect to the licensee's programs of control for electrical work

activities and its efforts to ensure ele:trical field configurations meet l design requirements, the inspector noted comprehenshe QA/QC cepartment {

i nv ol v e'ne n t . Based upon internal licensee assessments and NRC inspector ;

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j spot-check and review, no generic prcblems or violations were identified, .

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12. Management Meetings _

On August 17, 1988 a meeting was held in King of Prussia, Pennsylvania l with NHY senior managers at the request of the NRC. The purpose of the meeting was to discuss licensee plans for heatup, initial criticality and '

low power testin In addition, the current status of NRC Bulletin 88-05 !

was prosente Both parties agreed to meet again prior to Initial !

criticalit A copy of the meeting handouts and attendance sheet is !

appended to this report as Attachments A and B, respectivel j At periodic intervals during the course of this inspection, meetings were held with plant managment to discuss the scope and findings of this -

inspection. An exit meeting was conducted on September 9, 1988 to discuss the inspection findings during the period. An additional meeting as held '

on September 22, 1988 between the Assistant Station Manager and the Senior Resident inspector to discuss item status not covered in the previous exit meeting. During this inspection, the NRC inspector received no comments .

from the licensee that any of their inspect'on items or issues contained proprietary information. No written material was provided to the licensee i curing this inspection other than a listing of minor inspection i deficiencies summarized in paragraph 3.a of this repor l

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J ATTACHMENT A NHY/NRC MEETING ON AUGUST 17, 1933 NRC REG'ON I, KING OF PRUSSIA, PENNSYLVANIA Name Title i Organi z ajt_icLn W. Russell Regional Administrator NRC/RI W. Kane Director, Division of Reactor Projects NRC/RI

W. Johnston Director (Acting), Division Reactor NRC/RI Safety J. Wiggins Chief, Projects Branch 3 NRC/RI

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R. Gallo Chief, Operations Branch NRC/RI 0. Haverkamp Chief, Reactor Projects Section 3C NRC/RI M. Shanbaky Chief, Radiation Safety Section NRC/RI a

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A. Cerne Senior Resident Inspector NRC/R1 O. Ruscitto Resident Inspector NRC/RI D. Brinkean Project Manager NRC/NRR

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R. Wessman Director, Project Directorate I-3 NRC/NRR y F. Brown President NHY G. Thomas Vice President, Nuclear Production NHY

T. Feigenbaut Vice President, Engineering, Licensing NHY and Quality Programs 0. Moody Station Manager NHY J. Vargas Manager of Engineering NHY J. Warnock Nuclear Quality Manager NHY R. Sweeney Washington Lt +nsing Representative NHY i

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! New Hampshire Yan<ee  !

) Presentation  ;

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AGENDA  :

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Introduction G. S. Thomas :

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NHY Organization E. A. Brown !

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J Low Power Test Program G. S. Thomas :

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Self-Assessment T. C. Felgenbaum

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Status of Bulletin 88-05 J. J. Warneck

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Conclusion G. S. Thomas F

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NEW HAMPSilRE YAN (E E O 3GAN ZATION E. A. Brown

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NEW HAMPSHIRE YANKEE ORGANIZATION Posident & CEO PSNH President & CEO NHY

I Executive Director Vce President Vce President Comptroller Emergency Pfarnry Nuclear Production Engmeering. Lhasirg E

& & Quality Programs CAO Commuruty Relatons Regulatorv 'h- Independent Resnew Team (IRT)

I I I I I I I Station Training Production Engmeering Nuclear bcensiry Reliability / Safety Manager Manager Sennces Manager Quality Manager Engmeenng Manager Manager Manager

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0W POWER TEST PROGRAM G.S. Thomas

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LOW POWER TEST PROGRAM STARTUP ORG.ANIZATION Startie Manager fleactw Startup Supervism

, SNft Test Shift Test Shift Test i Directs Directw Directw 1 I i

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3 Test Directas 3 Test Directws 3 Test Directas 3 Startup Engineers 3 Startup Engineers 3 Startup Engineers

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LOW POWER TEST PROGRAM STARTUP ORGANIZATION .

e Shift test directors and test directors (directors of test activities) !

will be qualified in accordance with the requirements of Reg Guide 1.8 as specified in the FSA e All shift test directors and test directors have previously worked in the Seabrook preoperational and startup test program ,

o Test personnel will be formed from the following organizations: l

- Technical Support

-- Engineering ,

- Operator Training  :

- Regulatory Services  :

- Yankee Atomic Electric Company l

- Westinghouse Electric Company l l

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TECHNICAL SPECIFICATION SURVEILLANCE TESTS e All local leak rate tests (Type B & C) have been reperformed e Emergency diesel generator and engineered safety features actuation  ;

testing scheduled for the last two weeks in August l

e Other surveillance testing has been incorporated into the schedule l

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i PREVENTIVE MAINTENANCE Data Date 8/2/88

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ACTIVITIES PERFORMED DEPARTMENT 1987 1988

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Mechanical 1868 1144 Electrical 2834 1558 '

l&C 2634 1435 Utilities 536 207 TOTAL 7872 4344 ,

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f MAN-HOURS CONSUMED l 1987 1988

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TOTAL 47232 20267

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PREVENTIVE MAINTENANCE  :

Data Date 8/2/88 i i

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1987 1988 1  ;

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Activity 56 % 51 % l i  !

! Man-hours 32 % 22 %  !

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e RADIATION PROTECTION TIME PRIOR TO CRITICALITY ACTIVITY 4 weeks Start reissue of dosimetry to qualified rad workers 1 week Establish Radiological Control Area for training Just prior Establish full Radiological Controlled Area (RCA)

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Just prior implement full radiation protection program i

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OPERATIONS

Licensed Operators 23 SR0*

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32 Total !

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Staff Licenses 5 SRO-Operations

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9 SRO-Training

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14 Total Includes 16 STA - Qualified Operators

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EMERGENCY PREPAREDNESS ON-SITE EMERGENCY RESPONSE ORGANIZATION (ERO)

e Fully staffed and trained e Demonstrated during 1986,1987 and 1988 Graded Exercises

, e Fully implemented since receipt of Zero Power License e Meets requirements of proposed change to 10CFR 50.47(d)

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S EL:-ASS ESS V E \

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_0W POWER TESTING EVOLLTION T.C. Feigenbaum l

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PURPOSE:

To perform a self-assessment of the preparation for and the conduct of activities associated with the Seabrook Station low power testing evolution in order to assess the readiness and effectiveness of personnel, programs and equipment and to identify areas requiring immediate or long term management attention.

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SCOPE:

l The scope of the self-assessment effort will include, as a minimum, the following topical areas:

1. Plant operations 2. Radiological controls 3. Maintenance 4. Surveillance and testing 5. Safety assessment / Quality verification 6. Control room operations

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7. Effectiveness of internal problem identification and resolution

8. Plant chemistry and health physics

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Within the above topical areas, the self-assessment effort will focus on the following organizational conduct and activities:

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1. Orgenizationalinterfaces and management effectiveness 2. Plant configuration control 3. Program / procedural adequacy and compliance l 4. Communications and teamwork 5. Operational Quality Assurance effectiveness

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6. Timeliness and adequacy of support of Station activities

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7. Training program adequacy and effectiveness 8. Timeliness and adequacy of corrective action reporting and

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follow-through 9. Adequacy of design based on Low Power Test Program elements

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SELF-ASSESSMENT TEAM ORGANIZATION:

MANAGEMENT OVERSIGHT COMMITTEE e E. A. Brown - President and CEO e G.S. Thomas - V.P. Nuclear Production e T.C. Feigenbaum - V.P. Engineering, Licensing and Quality Programs e D.E. Moody - Station Manager SELF-ASSESSMENT TEAM MANAGER e N. A. Pillsbury - Independent Review Team Manager SELF-ASSESSMENT TEAM MEMBERS *

AREAS of EXPERIENCE:

e Operations e Maintenance e Chemistry / Health Physics e Training e Engineering / Technical Support e QA/QC e Independent Safety Engineering Group

• Approximately 30% of each work week to be dedicated to evaluat;on activities

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NRC INTERFACE:

l e Periodic updates by Team Manager and members of the Management Oversight Committee (bi-weekly suggested)

e Final report available to NRC Resident and Region 1 office (approximately 6 weeks after completion of Low Power Testing) -

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• Normal daily contact with NRC Resident as required e NHY/NRC critique of performance following completion of majpr activities

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SCHEDULE of ACTIVITIES:

WEEK LOW POWER TESTING SELF-ASSESSMENT MGMT. OVERSIGHT COMMITTEE 1 Preparation 2 Preparation Team Preparation Team & Mgmt Briefing 3 Preparation Self Assessment Team Start t 4 Preparation Self Assessment Team Status Report 5 Preparation Self Assessment .

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6 Heatup Self Assessment Team Status Report 7 Heatup Self-Assessment Team Status Report

- Precritical Concurrences Status -

8 Low Power Tests Self Assessment Team Status Report 9 Low Power Tests /Cooldown Self Assessment I

10 Layitp Self Assessment Team Status Report 11 Layup Self Assessment 12 Layup Self Assessment Team Status Report 13 Layup Self Assessment End 14 layup Draft Report D/R Internal Distribution 15 Layup 16 Layup Issue Final Report Team & Mgmt Debriefing i

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l STA"L S 0" B J _LE l \ 88-05 J.J. Warnock

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BULLETIN 88-05 SUMMARY e Falsified CMTRs - WJM/ PSI / Chews Landing e identify Installed Fittings and Flanges (F/F) and other material and test e Engineering evaluation for F/F as required e Written report to NRC

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'O SEABROOK APPROACH e Documentation review e Field walkdowns e Procedures developed e Testing of installed F/Fs e Laboratory testing of selected F/Fs e NUMARC/EPRI support e Engineering Evaluation e Additional confirmations

- DRAVO

- Radnor Alloys

-- Other suppliers

- Continued NUMARC support

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O BULLETIN 88-05 RESULTS DOCUMENTATION REVIEW e Complete e 358 WJM flanges / fittings installed; 12 F/F vendor markings not available (B31.1 only)

e 13 S/R ASME systems affected; 1 S/R B3 systems affected e Predominently carbon steel (5 stainless steel flanges)

TEST RESULTS e 368 tested e 30 requiring engineering evaluation e No replacement anticipated 0THER e DRAVO review consistent l e Supplier responses

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