IR 05000443/1987026

Safety Insp Rept 50-443/87-26 on 871208-880201.No Violations Noted.Major Areas Inspected:Operational Safety,Licensee Action on Previous Insp Findings,Followup Issues,Usi A-26, Allegation Followup & Control Room Ventilation

ML20147G953
Person / Time
Site:	Seabrook
Issue date:	03/02/1988
From:	Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20147G927	List: IR 05000443/1987026 ML20147G922
References
REF-GTECI-A-26, REF-GTECI-RV, TASK-2.D.3, TASK-2.G.1, TASK-A-26, TASK-OR, TASK-TM 50-443-87-26, NUDOCS 8803080351
Download: ML20147G953 (16)

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. U. S. NUCLEAR REGULATORY COMMISSION REGION I .

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Report N /87-26 Docket N , License N NPF-56 Permit N CPPR-135  : Licensee: Public Service Company of New Hampshire 1000 Elm Street - Manchester,- New Hampshire 03105 Facility Name: Seabrook Station, Unit N Inspection At: Seabrook, New Hampsnire Inspection Conducted: December 8, 1987.- February 1, 1988 Inspectors: A. C. Cerne, Senior Resident Inspector D. G. Ruscitto, Resident Inspector I C. A. Carpenter, Resident Inspector (Yankee Nuclear Power Station)

Approved By

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DonaTd R. Haverkamp, Chief f uw w[ 2/2./#d" f Oate ' Peactor Projects Section NV. 3C '

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il Inspection Summary: Inspection on December 8, 1987 - February 1, 1988

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     (Report No. 50-443/87-26)   '

Areas Inspected: Routine safety inspection during normal and backshift periods  ! by three resident inspectors (98 hours). The areas reviewed included opera-  : tional safety, licensee action on previous inspection findings, follow-up  ! issues, unresolved ~ safety issue A-26, allegation follow-up, control room venti- l 3 lation, design changes / modifications and technical specification l Results: No violations were identified. One issue, involving the temporary .l , loss of control room pressurization, is still under review for reportability  ; j in accordance with 10 CFR 50.73. The actions taken by the licensee to alert  ! l the control room operators of any similar situation are also being evaluated i with respect to the adequacy of corrective measures to preclude reoccurrenc , This issue is discussed in paragraph 8 of this report and remains unresolve i

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In paragraph 5, licensee review of problems occurring at other plants and licensee action to evaluate nonsafety problems at Seabrook are discussed. The  ! licensee responsiveness to these technical issues was effective in providing evidence and assurance of no adverse impact or relationship to safety-related ,

,    equipment at Seabrook Station. Generic implications were assessed and appro-  l 2    priately addressed by the cognizant licensee technical personne I

I 8803080351 880302 i PDR ADOCK 05000t43 l 0 DCD j l

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TABLE OF CONTENTS Page i Persons Contacted............................................ 1 Summary of Facility and NRC Activities....................... 1 Operational Safety........................................... 2 Plant Inspection Tours (71707).......................... 2 Operational and Security Events (93702)................. 3 Licensee Action on Previous Inspection Findings. . . . . . . . . . . . . . 5 Unresolved Item 8 6 - 5 8 - 01 ( 9 2 7 01 ) . . . . . . . . . . . . . . . . . . . . . . . . 5 NRC Compliance Bulletin 87-02 (TI 2515/26).............. 5 Licensee Event Report 87-025 (92700).................... 6 Fo l l ow- u p I s s u e s ( 6 2 7 0 0 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 Equipment Problems at Other Plants...................... 7 U.S. Tool & Die Inspection............................... 7 c, Industrial Crane Failure................................ 8 Caustic Fill Line Leak.................................. 9 , Unresolved Safety Issue (USI) A-26 (TI 2515/19)..............

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9 Allegation Follow-up (92701)................................. 10 Valve Material Traceability Problem..................... 10 Crack in the Core Barrel.... ........................... 11 Control Room Ventilation (92702)............................. 12 9. Design Changes / Modifications (37700)......................... 14 10. Technical Specifications (61700)............................. 14 Administrative Controls................................. 14 Containment Leak Rate Testing........................... 15 11. Unresolved Items............................................. 16 12. Management Meetings (30703).................................. 16

* NOTE: The applicable NRC Inspection  Manual inspection  procedure or !

temporary instruction is listed in parentheses for the appropriate report section . I

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DETAILS 1. Persons Contacted W. A. DiProfio, Assistant Station Manager T. C. Feigenbaum, Vice President, Engineering and Quality Programs W. J. Hall, Regulatory Services Manager

* 0. E. Moody, Station Manager G. S. Thomas, Vice President, Nuclear Produccion J. M. Vargas, Manager of Engineering J. J. Warnock, Nuclear Quality Manager
* Attended exit meeting conducted on February 8,198 Interviews and discussions with other members of licensee and contractor management, and with their staffs, were also conducted relative to the inspection of items documented in this repor . Summary of Facility and NRC Activities During this reporting period, the riant remained in operational Mode 5, cold shutdown, with primary temperature between 110 and 140 degrees F and depressurize On January 28, 1988, Public Service Company of New Hampshire, one of the named licensees of Seabrook, Unit No.1, filed a voluntary petition for relief under Chapter 11 of Title 11 of the United States Code in the United States Bankruptcy Court for the District of New Hampshir By letter (NYN-88013) dated February 2,1988, PSNH notified the NRC Region !

Office of this Chapter 11 filing in accordance with 10 CFR 50.54(cc).

Subsequent to the filing, the inspector discussed its effects on current and future operation of the Unit No.1 facility with senior New Hampshire Yankee managemen No adverse impact on scheduled plant maintenance and test activities was identified and appropriate contingency planning on the part of plant supervisory personnel was note I On December 15-17, 1987, NRC Region I (NRC:RI) emergency preparedness ' specialist inspectors witnessed an exercise of the onsite portion of the i facility emergency pla During that inspection (Inspection Report N ] 50-443/87-25), several NRC inspector follow-up items were reviewed and i closed. On January 4-5, 1988, an NRC:RI operator licensing examiner, l accompanied by a contract engineer, conducted a re-examination (Report N l 50-443/88-01) of one licensed operator candidat l i i l l l l

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Also, during this inspection period, the resident inspectors attended an NRC:RI Resident Counterpart Meeting on December 15-17, 1987 and partici-pated in a meeting between NRC:RI and the Employee's Legal Project on December 29, 1987, both meetings held in King of Prussia, Pennsylvani The resident inspector was a member of the NRC:RI team that conducted an inspection of the Yankee Nuclear Power Station on January 11-15, 198 The resident inspector also attended a three-week training course in boil-ing water reactor technology at the NRC Technical Training Center, com-mencing on January 25, 1988. Additionally, the senior resident inspector served as team leader for an NRC:RI team inspection at the Oyster Creek alant on January 25-29, 1983 to follow up an operational inciden . Operational Safety Plant Inspection Tours The inspectors observed work activities in progress, completed work and plant status in several areas during general inspections of the plant. The inspectors examined work for any apparent defects or non-compliance with regulatory requirements or license condition Particular note was taken of the presence of quality control inspec-tors and quality control evidence such as inspection records, mate-rial identification, nonconforming material identification, house-keeping and equipment preservation. The inspectors interviewed station staff, craf t, quality inspection and supervisory personnel as such personnel were available in the work area . During control room observation periods, during both normal working hours and on backshif ts, the inspector reviewed control room logs and records including night orders, shift journals, shift turnover sheets, completed repetitive task sheets, the temporary modifications log, weekly surveillance schedules and control board indication Specific note was taken of equipment in "pull-to-lock" conditions, equipment tagged, alarm status and adherence to technical specifica-tion limiting conditions for operation (LCO) and action statement Also, boron samples, taken from the reactor coolant system and con-nected water supplies, were spot-checked for concentration, sample frequency and documentation in accordance with specified zero power license (NPF-56) condition The inspector verified the proper position, in accordance with oper-ational procedure or tag out controls, of specific valves during system walk-downs and cnecked the valve status in the control roo 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

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n i field controls and coordination of the subject work activity with the i" control room and operations personnel on shif t. In certain - cases, the operability of specific components and the applicability of the

;      observed work to the TS requirements were discussed with the

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No violations were identifie b. Operational and Security Events- ' i

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Several events of minor safety significance and/or reportable in accordance with 10 CFR 50 or 10 CFR 73 occurred during this inspec- , tion perio These events are documented belo ,

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, (1) On January 29, 1988, a control building air handling system  ; a (CBA) isolation occurred as a result of planned, modification j

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work to replace certain relays in the CBA system. The subject ' relays ensure the system functions in accordance with the single i failure criterion by implementing an actuation / isolation logic which affects components in both CBA train Thus, the relay replacement work inherently affects both trains and caused a i

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spurious ESF actuation, i.e. , total CBA isolation. Upon actua-tion, the ESF equipment performed as designe I

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Following notification by the control room that a CBA isolation l t had occurred, the original relay was returned to its initial

!      condition, control room remote air intake ventilation restored,  ,

j and work suspended on -the subject relay replacemen An LER  ! d (83-001) on this event is -currently under development by the ( j license ; i  ;

;      Subsequently, the inspector discussed with the shift superin-  !
{      tendent the impact of making both trains of CBA inoperable to  i
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complete the required relay replacement work. Technical Specif-ication (TS) 3.7.6 was reviewed and a Technical Support Group  ; written evaluation was performed to determine the options and j preferred course of action before reinitiating the field work.

j It was determined that although temporary jumpers could be

{      installed to allow a CBA fan to continue to operate during the  l J      relay replacement, both trains of CBA would still be technically  ;
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inoperable because the surveillance requirements for the CBA isolation capability could not be me Also, the use of the i jumpers and associated system modifications would require both  ; j trains of CBA to be inoperable for a significantly longer period  !

of time than merely shutting down and isolating the entire  !

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The license therefore decided to enter TS Action Statement 3.7.6.b, declaring both control room area ventilation systems !

' inoperabl Control room air conditioning would _ not be_ affec- !

ted, so as not to adversely impact the equipment and .instrumen-tation temperature controls or limitations. Inspector review of : the licensee documentation, discussion with both operations and ! technical support personnel, and assessment of the option chosen ! to allow the relay work to continue _ without~ additional spurious

.ESF actuations were conducted. All revealed proper considera-tion of the technical concerns and the appropriate use of both analysis and judgemen ;

While no violations were identified, LER 88-001 will be reviewed , in its written form, after issuance, to inspect and evaluate the , complete licensee analysis of this even j

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With respect to the initial follow-up of the above event Land other- inspection issues, as noted,- no violations were ! identifie , i

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(2) On December 21 and 23,1987 and January 17 and 28,1988, fou ,

separate security incidents occurred for which one-hour notifi- ? cation to the NRC Operations Center was initiated in accordance ; with 10 CFR 73.71. Two Licensee Event Reports (LER) 87-026 and ; 87-027 and one Security Event Report '(SER) 88-001 have subse- 1 quently been issued and a second SER.is being processe .I i

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Two incidents involved station employee access controls while another involved visitor escort controls. In the. fourth ninci- t dent, a security officer was discovered apparently asleep on , pos The inspectors reviewed these_ events following their i occurrence for the immediate consequences and licensee analysis / ,

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action. No adverse safety impact, directly resulting from these incidents, was identifie In the case of the inattentive security officer, the time duration in question was confined to l a period of 17 minutes for which computer- access logs revealed i no unauthorized entries. The subject officer was suspended and j subsequently discharge , For the other incidents, the adequacy of existing programmatic ! controls for plant access and visitor escort were initially ' reviewed for any contributory causal relationship to the event l Based upon past incidents (reference: NRC Inspection Reports . 50-443/87-23 & 24), the effectiveness of licensee corrective l action requires further NRC review, which will be accomplished l 1

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during the course of inspection and closure of the individual LERs and SERs. For the protection of the "Safeguards" nature of certain of the event details and also for a comprehensive, expert review of the generic issues involved with both the cur-rent incidents and the related, previously reported events, the subject LERs and SERs will be reviewed by a Region I security specialist during the next scheduled NRC security inspectio No violations were identifie . Licensee Action on previous Inspection Findings (Closed) Unresolved item 86-58-01: Shunt Trip Relay Spare Stock Items. During an NRC inspection (443/86-58) in December, 1986, the l inspector noted that Class IE automatic shunt trip relays had been ! added as a design modification to the reactor trip breaker circuitry, l as recommended by Westinghouse. However, at the time of the inspec-l tion, no spare relays for the automatic shunt trip feature had yet ' been procured.

' During the current inspection, two Potter & Brumfield, Type MDR-5134 relays were procured, receipt inspected and placed in spare parts storage at Seabrook for issuance when needed. The inspector reviewed the Westinghouse quality documentation, including certificates of conformance and operating characteristics, for the subject relay Licensee QA inspection had noted the need for additional certifica-tion data from Westinghouse with respect to Class 1E applicability l and IEEE standard cornplianc Such quality documentation was subse-quently received, and all material purchase requisition requirements l met by the vendor (Westinghouse).

! The availability of spare parts for the reactor trip switchgear (1-CP-CP-111) is now in conformance with the recommendation of the Westinghouse Reactor Trip Breaker Type DS-416 Maintenance Manual . l This issue is therefore resolved and the item 1., closed.

! (Closed) NRC Compliance Bulletin (No.87-02): Fastener Testing to Determine Conformance with Applicable Material Specification In accordance with the prescribed actions to be taken by the licensee, l the NRC inspector participated in the selection process of the sample l of twenty safety class and ten non-safety class fasteners, with I typically appropriate nut This composite sample lot encompassed ! all types of the fasteners listed in Bulletin 87-02, which were available in the Seabrook storage suppl t P

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The method of testing, both mechanical and chemical; the material purchase requisition requirements for the testing services; and QA surveillance over the sample selection and shipping activities were all discussed by the inspector with the cognizant licensee regulatwry services, procurement and QA personnel. On January 21, 1988, the licensee response (NYN-88004) to Bulletin 87-02 with the required Fastener Testing Data Sheets and test results was forwarded to the NRC for review. The timeliness of this response was evaluated by the inspector with regard to the 60-day reporting requirement prescirbed in the Bulletin. Although the licensee response letter was delayed in order to procure and analyze additional testing data from the testing laboratory, this delay had been discussed with the NRC inspecto The inspector concurred with the licensee decision to submit the complete test results in one package, given that the expected delay time was not considered excessiv The inspector reviewed the subject test results and accompanying fastener data for conformance with the bulletin requests for informa- . ' tion. The licensee discussion of the sampling technique, procurement requirements and storage controls was evaluated for consistency with the material control programs in effect, both during construction and under the current operating license. Safety-related fasteners were appropriately referenced to the governing material and fabrication specifications (e.g., ASME Section III or ANSI with IEEE Class 1E controls). The licensee evaluation of test data for the one bolt evidencing hardness results outside the material specification limits was reviewed for both code justification and safety impac The inspector assessed all the actions taken by the licensee in response to Bulletin 87-02 and found them to be properly scoped and in compliance with the bulletin intent and sound engineering judge-ment, No violations were identifie While further analysis and generic review of the test results will be undertaken by the NRC Office of NRR, for inspection purposes at Seabrook, NRC Compliance Bulletin 87-02 is considered close c. (Closed) Licensee Event Report (LER 87-025): Incomplete Surveillance Testing Dat This deficiency involved incorrect selector switch settings on the vibration meter used during the conduct of ASME Section XI Inservice Testing. The inspector was monitoring from the control room the performance of the vibration testing of the primary component cooling water pumps at the time the problem with the read-ings was discovere Subsequent licensee evaluation of this issue determined that although certain frequencies had been filtered out during previous tests, valid vibration data had still been compiled from those tests. Review of that data would have indicated the existence of a potential proble .

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Licensee corrective action was reported to the NRC by letter (NYN-88001), dated January 6,1987. The inspector reviewed this report and evaluated the control room operator activities associated with this surveillance activity. No discrepancies or performance problems were identified. This LER is close . Follow-up Issues Equipment Problems at Other Plants Two specific equipment problems which had occurred at other plants were brought to the attention of the licensee and reviewed for applicability at Seabroo One involved auxiliary feedwater pump turbine trips and oscillations caused by undersized springs in the Woodward governor and excessive condensate trapped in the steam supply lines to the Terry turbine. At Seabrook, based upon extensive redesign and post modification testing of the steam supply to the emergency feedwater pump turbine, both the governor controls and condensate build-up were frequently checked. Analysis of the prob-lems encountered at the other plant by the licensee revealed no similarity to either the governor control or condensate problem The inspector reviewed the licensee evaluat'.on of the events docu-mented in Nuclear Network Operating Event Report No. OE 2259 and concurred with the rationale for a nonapplicability determination at Seabroo The other problem involved failed safety-related pump starts at another plant due to Westinghouse electrical supply breaker malfunc-tions. Further review revealed no relationship between the proble-matic "DB" breakers supplied by Westinghouse to the plant where the failures occurred and the ITE Gould breakers supplied to Seabroo Licensee technical support personnel evaluated the available details of the 03 breaker component clearance problems and confirmed their nonapplicability at Seabroo No violations were identifie U.S. Tool and Die Inspection As documented in NRC Region I inspection report No.443/87-16, the licensee was informed of problems identified by the NRC Vendor Inspection Branch at U.S. Tool and Die, Incorporated. The specific problems were related to the fabrication and testing of spent fuel storage racks which at Seabrook were not fabricated by the subject vendo However, since U.S. Tool and Die had supplied the new fuel storage racks, the licensee initiated an evaluation of the specific deficiencies for their applicability to Seabroo .. , .

The licensee analysis included not only a review of past vendor sur-veillances and audits conducted at U.S. Tool and Die and of new fuel rack receipt weld inspections, but also ~.the conduct of an additional visual examination of a sample of the rack welds, in place in the fuel storage building. . The inspector reviewed the overall licensee e,; i , si s o_f_ this issue, specifically noting that the problems identit led in the NRC vendor inspection report No. 99901082/87-01 were not in evidence at Seabrook. The QA involvement in the.evalua-tion of this issue was noteworthy. While the licensee weld inspec-tion did reveal certain configurations requiring additional engineer-ing analysis, no major quality-related problems were note The insoector had no further questions on the licensee actions taken to follow-up this issu No violations were identified, Indugr,ial Crane Failure As documented in NRC Region I inspection . report No.443/87-24, a structural failure on an overhead crane in the circulating water pump house caused an industrial accident on November 20, 1987 which irJured two workers. An accident investigation committee was ' formed to evaluate this. incident from a personnel safety perspective, : as well as for generic applicability to other cranes throughout the plant. New Hampshire Yankee contracted Teledyne Engineering Services to perform an evaluation of the failed crane equalizer sheave and

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support assembly, including inspection and metallurgical analysi The NRC safety interest in the follow-up of this accident relates to any potential for adverse impact to safety-related equipment which might be attributed to a common defect in the design or fabrication of other cranes manufactured by the same supplier, Shepard-Nile The inspector reviewed the final accident report, dated December 22, 1987, issued as a result of the committee investigation, chaired by the Assistant Station Manager. The cause of the accident was attributed to a combination of an impact loading with a low safety factor, and marginal material quality in the susceptible area of the crane equalizer sheave and support assembl Because of the potential for generic applicability, the licensee con-ducted an inspection and evaluation of all Shepard-Niles hoist sys-tems at Seabroo The resuits, documented in the final accident report, indicate proper consideration of the evaluation criteria required by NUREG-0612 and the conduct of additional nondestructive examination where equalizer sheave and support design details similar to the failed crane were identifie Only one crane still remains tagged out pending the evaluation of the magnetic particle testin The inspector determined that the licensee investigation of this accident and the resulting report were responsive to NRC concerns regarding the generic safety impact. The inspector had no questions on the planned licensee actions to preclude recurrence of such an incident from a personnel safety standpoint. No violations were identifie . L

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9 l t ( Caustic Fill Line Leak The NRC Region I inspection report No.50-443/87-10 described a -leak  ; on the caustic fill line drain connection, which is part of the steam p generator blowdown recovery system, in April.1987. A request for . !  ; engineering services RES 87-568 was initiated as a result of this occurrenc Design coordination report (DCR) 87-192 was issued to -

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provide seals for the . subject penetrations. -The inspector reviewed ' ! the DCR implementation plan for full service notification and the ! applicable implementing work request No. violations were  ; identifie ; 6; Unresolved Safety Issue (USI) A-26, Reactor Vessel Pressure Transient '

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Protection Background I The technical issue involved with Unresolved Safety Issue A-26 relates to the safety margin-to-failure for reactors .which may, be-subject to severe pressure transients while at relatively low temper-ature The majority of industry events in this area have ' occurred ' during shutdown and startup while the reactor coolant system was in a solid condition. Plants licensed after March 13, 1978 were required 1 to install fully automatic protection system j A review of the NHY actions taken in response to licensing commit- ' l ments concerning this unresolved safety issue was' conducted. Pre-vious inspecticn activity conducted in this area is documented in NRC a Region I inspection report No.50-443/86-12 (TMI Item II.D.3, Direct Indication of Relief and Safety Valve Position and TMI Item II.G.1, Emergency Power for Pressurizer Equipment). Also, licensing committ-ments were made with respect to this issue and are documented in Section 5.2.2 (Overpressure Protection) of the Seabrook Final Safety Analysis Report, the NRC Seabrook Safety Evaluation Report (SER) and subsequent SER Supplement , Inspection Field verification of pressurizer power operated relief valve (PORV)  : position indication and power supplies was conducted during inspec- ' tion 50-443/36-12. During this current inspection, the inspector i performed a detailed review of the PORV ar.d PORV block valve actua-  ; tion circuitry, and discussed the design of the system with the l cognizant NHY and YAEC engineers. The applicability and scope of the requirements to provide safety grade control circuits for the PORV ] and block valves were questioned by the inspector with raspect to the i Seabrook design. It was agreed that in accordance with the intent of the TM1 Action Plan, safety grade controls were required for manual , operation only. The inspector verified that the manual centrols were  ! capable of overriding any failure of the non-safety grade automatic - 1 control I I I I i

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The inspector also confirmed that 'the subject - controls were from  ! independent power supplies and were electrically separated in accord--  ! ance with _the appropriate industry standard No violations were

!  identifie . Allegation Follow-up_       !

c , ) Two allegations involving component quality during the construction phase i j of Seabrook, Unit I were brought to the attention of the NRC. In one case, the alleger indicated that the concern had been reported to and

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evaluated by the licensee. For the other allegation, no specific details

1 of the concern were initially documented, nor was there any indication  ! that the licensee had been informed of the concern. In both cases, subse- '

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quent information was provided by the allegers to clarify- the details of the individual concerns in a manner which would permit meaningful inspec- < tion of each allegation. The NRC inspection follow-up of these two issues ' ! is documented below: , Valve Material Traceability Problem .

Original Allegation

"There was a materials traceability - proble Certain valves on the_ steam generator lacked the engraved manufac- *

!l turer's number." , 1  ! Additional Information: "Relief valves 456A and 4568, on the steam  ! l generator, noted during the first = hydro test-in Unit 1 (RC IT 01A)."

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!  Inspection Follow-up: The inspector noted that the two valves spec-i  ified by the alleger represent the power operated relief valves    ,

4 (PORVs) for the pressurizer, not a steam generator. The inspector  ! ! visually examined the current condition and status, including valve j identification and tagging, of these reactor coolant (RC) system I

valves, RC-PCV-456A & B, and identified no marking or traceability I i deficiencies, j 1  ;

Additionally, conduct of the reactor coolant system hydrostatic test l

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 (RC-IT-01) in April,1985, was witnessed by the resident inspectors    !

l (reference: inspection report 50-443/85-01) and Region I specialist 3 inspectors (reference: inspection report 50-443/85-08). The RC-PCV-  !

456A & B represented hydrostatic -test boundary valves which were  ; } spot-checked by the inspectors during the conduct of RC-IT-01. Sub- , ! sequently, during a Region I resident inspection (reference: inspec-  ! I tion report 50-443/85-25), the inspector witnessed PORV modification l [ in accordance with an engineering change authorization and under the i , direction of Westinghouse and Crosby Valve engineers. This work was ' ! verifed to have been properly conducted in accordance with ASME Section XI requirements and additional hydrostatic test requirements  ! j for the PORV pressure boundary were delineated. The hydrostatic test i of RC-PCV-456A & B was later witnessed by an NRC resident inspector,

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as is documented in inspection report 50-443/85-3 l

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. At no time during the conduct of previous NRC inspections, as docu-mented above, of the subject valves (RC-PCV-456A & B) were tracea-bility or component identification problems identified. QA coverage of the PORV modification work, documented in inspection report 50-443/85-25, programmatically checked material identification / traceability prior to any ASME welding or part replacemen Based upon previous NRC inspection of the valves in question and the conduct of a recent inspection to ensure that the current valve identification is both proper and traceable, the allegation was not substantiate b. Crack in the Core Barrel Original Allegation: "A quality assurance man in Pullman-Higgins believed the core barrel was cracked. The QA person reported this to the company, and engineers responded to his concern, but the QA inspector was never satisfied that the core barrel was not cracked."

Additional Information: "About 2 1/2 to 3 years ago, while he was with a New Hampshire Yankee inspector and another inspector, they were there while the core barrel was being moved. He saw a crack about 18 inches long which changed direction about four or five times. Both of the other inspectors saw it as well, so it was reported at the time to NHY."

"The crack was located in an area where there is an upper and lower flange protruding out from the core barre Sketches to follow later."

A sketch, marked up to designate the alleged crack, was transmitted to the NRC with a letter noting the following applicable information:

"He drew the zig-zag line on the enclosed drawing of the core barrel to indicate where he saw the crack and drill holes at the junction of each crack."

Inspection Follow up: The inspector reviewed the Employee Allegation i Resolution (EAR) program files documenting the concern, interviewed l the QA inspector that accompanied the alleger on an inspection of the l core barrel and discussed with other NRC pet sonnel the independent I NRC inspections of the core barrel while in its storage position in l the refueling cavity and during insertion and removal from the l reactor pressure vessel for testin Historically, NRC inspectors j have conducted examinations of the reactor pressure vessel internals, l including the core barrel, both in storage and during the installa- l tion process (e.g. , Region I inspection reports, 50-443/81-01, 82-08, 82-12 and 83-05). In addition to the licensee programmatic quality I

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? , I assurance controls and inspections of the core barrel, Westinghouse ; engineering personnel were on site to observe and provide direction !

to the initial internals package installation proces Subsequently, ! during the' preoperational testing phase ~of Seabrook, Unit No.1, the ' , core. barrel was visually examined by an NRC inspector for evidence of (

physical damage'or any abnormalities.

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Separate from any routine NRC inspactions of the core barrel, the EAR file review revealed that the alleger, accompanied by a certified QA

' i inspector, conducted a reexamination of the core barrel in an attempt - to find the "crack" he had seen. An NRC interview with this QA - - inspector indicated that no evidence of a "crack" was discovered and , j that the alleger had expressed apparent satisfaction that no problem ,

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existed at that time. In July,1986 upon completion of his employ-

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ment at Seabrook, the alleger again raised the same concern to EAR : program personnel. An EAR investigator conducted .an additional , - examination of the ' core barrel and discovered no ' crack. The EAR investigation also reviewed the relevant quality records associated j with both the Unit No.1 and- No.'2_ core barrels and . identified no , evidence of cracking indication , l' Given the extent that the core barrel was examined, both prior to and' ' d af ter this allegatic, was raised, and given that a zig-zag crack with ,

,  drill holes at the junction of each crack would be expected to be ;

clearly visible upon examination, the inspectoe determined that this "

] allegation was not substantiated. The inspector also noted that the ! j EAR investigation had included checks for evidcnce of unauthorized i j repair welding on the core barrel, to include a review for any design ! modifications on structural attachment which could mask an unauth- l

orized weld repai None were identified.

- i With respect to the follow-up of both of the noted allegations, no viola- l tions were identified. The inspector considers both allegations close l No further follow-up is planned, pending Region I management revtew of the q satisfactory closure of these allegation ' l 8. C_ontrol Room Ventilation  ! l On September 20, 1987, licensee personnel discovered that the centrol room

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differential pressure was zero inches (water gauge, WG) with the control !

! room exhaust fan, CBA-FN-15, running and its exhaust damper full ope ; ! Seabrook Technical Specification 3.7.6 requires that a minimum of (+)1/8" ;

WG differential pressure be maintained between the control room and ; j adjacent areas tu preclude the entry of contaminant I i

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        'l Subsequent licensee investigation determined that the cable spreading room
. supply fan CBA-FN-17, had tripped due to a battery test on the local fire
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,'  ! protection panel on the previous day. This fan is not designed to auto-

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I matically restart following a trip. Therefore,:the pressure in the cable j

spreading room became slightly negative with respect .to atmosphere since ,

'- the exhaust fan CBA-FN-18 continued to run. The control room pressure  ! > contrciler senses differential pressures between the control room and  ! either atmosphere or the cable spreading room below. _ It was noted that  ! ! this is an auctioneered high signal and therefore provides control func- l i tions based upon the highest differential signa In -this particular' > j case, the differential pressure between the control room and cable spread- l , ing room was greater than that between the control room and atmospher t

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Therefors, the pressure controller was attempting to reduce the control - j room / cable spreading room differential pressure to (+)1/8" WG by maximiz- l ' ing the exhaust from the control room. However, CBA-FN-15 is much smaller r j than CBA-FN-18. Even with the control room exhaust damper fully open and  ! j the control room depressurized, CBA-FN-18 operation continued to cause a l j differential pressure greater than (+) 1/8" WG between the. control room t and cable spreading room. This resulted in a continued erroneous signal  ! , to the pressure controller which main +-ained the control room. exhaust Ian  ; i operating with the supply fan of ;

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Subsequently, by taking manual control of the pressure controller, the  !

control room operator was able to reestablish pressure in the control room by closing the exhaust damper. The inspector verified that no limiting i condition for operation (LCO) was violated since TS 3.7.6 requires action  ;

be taken to place the control room in the recirculation. mode after seven i days of component inoperability. The specific condition discussed above l existed only for about fourteen hours.

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The inspector reviewed the sequence of events and system design which l l resulted in a depressurized control room for the 14-hour period. Further  ;

i discussion with licensee personnel resulted in two issues which require additional evaluation or action by the licensee, as follows:

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j 1 (1) The potential reportability of this event based upon a loss of an engineered safety featur '

 (2) Short-term actions requires of the licensee to ensure that a similar j  situation cannot exist undetected for any length of time until a   -

final design modification to the subject pressure control functions i is implemented.

J l pending the presentation by the licensee of evidence that both of the

above issues have been adequately addressed. NRC concerns regarding the ,

j corrective measures remain unresolved (50-443/87"26-01). * ! ! l J l } i _ . , . _ , , - . _ - . , . . _ . _ _ _ _ _ _ _ _ _ - - - . - - - - - _ .

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. 9. Design Chinges/ Modifications The inspector reviewed the following design coordination reports (DCR's) and spot-checked work in progress in the field relative to work control, change authorization and engineering overview:

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DCR 87-082, "Modify Thermal Barrier Cooling System Head Pipe" I

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OCR 87-315. "Add Restricting Orifice to OG-E-42A and B Outlet" q The observed work was discussed with the cognizant engineering personne In the case of DCR 87-082, corrosion in the drain lines o'f the thermal barrier heat exchangers was chemically analyzed. The piping upstream of , valve CC-V-1090 was visually inspected by the licensee and also examined t by the inspecto Subsequent analysis revealed that the corrosion was specific to the stagnant conditions in the drain piping and not represen-tative of the conditions of the thermal barrier cooling system in genera The DCR 87-082 specifies the installation of rupture discs on the large vent openings on the thermal barrier head tank piping and the addition of a smaller expansion vent. These modifications, along with the installa-tion of a new chemical addition connection, will minimize corrosion due to excess oxygenation of the system and will allow for the more ef ficient addition of corrosion inhibitors to the syste With respect to DCR 87-315, the inspector raised specific cuestions regarding orifice sizing, weld relocation and modifications to tN service water valves, SW-V-16 & 18, previously designated as throttling valve Existing DCR change authorizations were found to provide the answers to certain NRC questions and justification for physical configuration changes in the desig For the valve rework, specific stroking criteria vere added to in process change authorization, CA-08, to clarify the need for

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full-open position adjustment The inspector also verified thail tha fail-safe testing of these valves to their 100*; open position would be routinely accomplished by an engineering surveillance, EX 1804.029, in accordance with in-service testing requirements. The inspector also verified the conduct of a weekly operational surveillance, 05 1426.12t for stroking the subject valve The inspector had no unresolved safety concerns regarding implementation of the above design coordination reports. No violttions were identifie . Technical Specifications a, Administrative Controls The inspector reviewed Saction 6.0, Administrative Controls of the j Seabrook Technical Specifications to determine the consistency in the l descriptions of the licensee off-site organization and station staff

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with the current r;rganizational structure of New Hampshire Yanke ,

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i The inspector noted that recent organizational changes have affected ! j the official titles and both the= function and responsibility of cer- , i tain personnel. Also, the organizational charts included in tho l Technical Specifications been superseded by the NHY reorganizations !

which have taken place since the license was issued in October, 198 ;

i The insp0ctor discussed the noted inconsistencies with licensee : !

't 4 regulatory service personnel and was provided a draft copy of Section -!
'.. 6.0 revisions to the Technical Specifications which will be submitted i

? to the NRC Office of Nuclear Reactor Regulation _(NRR) for. proposed ! future amendment to the Technical Specifications. The NRR' staff -. is . - expected to either (1) approve the changes as a separate ' licensing l action, or (2) incorporate such administrative changes, along with j any required technical revisions, into the revised Technical Specifi- ! cations which will serve as conditions to the low power operating _

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license, when i_ssued. The inspector had no further questions on the- ' way such orgarizational changes would be processe t l The inspector also examined the licensee controls over the Technical ; { Specification Improvement Program, as discussed in Section 6.7 of the 1 i Technical Specifications and Chapter 16.3 of the FSAR. Revision 3, i i effective December 29,1937, to the Technical Requirements Manual was :

' reviewed to verify the requisite Station' 0peration Review Committee !

  (SORC) and 10 CFR 50.59 evaluation The inspector checkeds the NHY {

Programs and Procedure Manual, noting a provision in procedure H210 * , 9 that prescAbes safety evaluations for all procedural cha'nges that

 ' require 50RC review. The Technical Specification Improvement Pro- l gram, as implemented by the Technical Requirements Manual, is admin-

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istratively controlled by both the Technical Specification conditions 1 i and the SCPC review process. A Nuclear Safety Audit and Review .

Committed . review of any change is also required. . Therefore, even l l though prior NRC approval to the Technical Specification Improvement l J

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Program revisions is not mandated by the existing facility license 1 j and Technical Specifications requirements adequate prendural control j i " and post-implementation reporting to the NRC appears evident in the , : NHY' progra l

, No violations were identifie t

l  ! [ Containment Leak Rate Testing I

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The inspector noted an apparent inconsistency between the surveil- j

lance requirements, Technical Specification (TS) 4.6.1.2.a for pri- 1 l mary containment leakage and a commitment in FSAR Chapter 6.2.6.4 for i the scheduling of Type A tests (Overall Integrated Contalement Leak-

! age Rate). The inspector verified -that the TS provisions were in conformance with 10 CFR 50, Appendix J criteria and discussed the i FSAR write-up with the appropriate technical support and regulatory . service personnel. The licensee subsequently provided the inspector I with a FSAR change request, revising Chapter 6.2.6.4 to bring it into ] conformance with the Technical Specification l

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.. Additionally,. the inspectcyl reviewed .an ' engineering -surveillance

'^g . procedure, EX 1803.004, effective, August 14,1987,-which discusses the containment interior steel liner and exterior ' concrete inspections '

conducted during each plant shutdown prior to. the Type. A containment , leakage rate tests. Con;rliance with ' the requirements of TS 4.6.1.6, -

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with reference to TS 4.6.'.2 schedular - provisions, was confirme . No violations were identifie . Unresolved Items An unresolved ite;n is a matter about w.iich more (nformation is required to ascertain whether it is an acceptable item, a> deviation, or a viola- -

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tion. An unre' solved item is discussed in7secteor. D of this repor . Management Meetings , . L At periodic intervals during the course ot/this irt:pection, ceetings were

  . held with plant' management to discuss t$e scopeJand findings of this inspectica. - An exit meeting was conducted on February 9,' 1988 to discuss the inspection findings during the period. During this' inspection, the NRC hrgpUtors received no comments from tha licensee thd. any of their inspe4( y/ttens or issues contained proprietary information. No written materialwjsgrovidedtothelicenseeduringthisinspectic '
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