Text

Forwards Sqrt Trip Rept & Pump & Valve Operability Review Team Trip Rept.Encl Repts by Inel Evaluate Team Audit Findings
	ML20203A487
Person / Time
Site:	Seabrook
Issue date:	04/04/1986
From:	Nerses V; Office of Nuclear Reactor Regulation
To:	Harrison R; PUBLIC SERVICE CO. OF NEW HAMPSHIRE
References
	NUDOCS 8604160364
	Download: ML20203A487 (4)
	v • d • e

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7 APR 4M Docket No. 50-443 Mr. Robert J. Harrison President & Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105

Dear Mr. Harrison:

SUBJECT:

SEABROOK TECHNICAL EVALUATION REPORTS FOR SQRT AND PVORT AUDITS Enclosed is one copy each of the Seismic Qualification Review Team (SQRT)

Trip Report and the Pump and Valve Operability Review Team (PV0RT) Trip

. Report. These reports were prepared by the Idaho National Engineering Laboratory staff who participated in the audits. The information contained in these reports constitute detailed evaluations of the SQRT and PVORT audit findings and are transmitted for your use.

Victor Nerses, Project Manager Project Directorate #5 Division of PWR Licensing-A

Enclosure:

As Stated cc: See next page DISTRIBUTION:

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Local PDR R. Ballard PD#5 Reading Files G. Bagchi T. Novak N. Romney OELD E. Jordan B. Grimes V. Nerses Office:

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NUCLEAR REGULATORY COMMISSION 3

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m m Docket No. 50-443 Mr. Robert J. Harrison PrisiB5iff T Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105

Dear Mr. Harrison:

SUBJECT:

SEABROOK TECHNICAL EVALUATION REPORTS FOR SQRT AND PV0RT AUDITS Enclosed is one copy each.of the Seismic Qualification Review Team (SQRT)

Trip Report and the Pump and Valve Operability Review Team (PV0RT) Trip

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Report. These reports were prepared by the Idaho National Engineering Laboratory' staff wh,o participated in the audits. The information contained in these reports constitute detailed evaluations of the SQRT and PV0RT audit

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findings and are transmitted for your use.

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l Victor Nerses, Project Manager Project Directorate #5 Division of PWR Licensing-A

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Enclosure:

As Stated cc: See next page l

Mr. Robert J. Harrison Public Service Company of New Hampshire Seabrook Nuclear Power Station cc:

Thomas Dignan, Esq.

E. Tupper Kinder, Esq.

John A. Ritscher, Esq.

G. Dana Bisbee, Esq.

Ropes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 208 State Hosue Annex Concord, New Hampshire 03301 Mr. Bruce B. Beckley, Project Manager Public Service Company of New Hampshire Resident Inspector Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear Regulatory Commission Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshirc 03874 Sun Valley Association 209 Sumer Street Mr. John DeVincentis, Director Haverhill, Massachusetts 01839 Engineering and Licensing Yankee Atomic Electric Company Robert A. Backus, Esq.

1671 Worchester Road O'Neil, Backus and Spielman Framingham, Massachusetts 01701 116 Lowell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manager United Engineers & Constructors William S. Jordan, III 30 South 17th Street Diane Curran Post Office Box 8223 Harmon, Weiss & Jordan Philadelphia, Pennsylvania 19101 20001 S Street, NW Suite 430 Washington, D.C.

20009 Mr. Philip Ahrens, Esq.

Assistant Attorney General State House, Station #6 Augusta, Maine 04333 Jo Ann 3hotwell, Esq.

Office of the Assistant Attorney General Environmental Protection Division Mr. Warren Hall One Ashburton Place Public Service Company of Poston, Massachusetts 02108 New Hampshire Post Office Box 330 D. Pierre G. Cameron, Jr., Esq.

Seabrook, New Hampshire 03874 General Counsel, Public Service Company of New Hampshire Seacoast Anti-Pollution League Post Office Box 330 Ms. Jane Doughty Manchester, New Hampshire 03105 5 Market Street Portsmouth, New Hampshire 03801 Regional Administrator, Region I U.S. Nuclear Regulatory Comission Mr. Diana P. Randall 631 Park Avenue 70 Collins Street King of Prussia, Pennsylvania 19406 Seabrook, New Hampshire 03874 Richard Hampe, Esq.

New Hampshire Civil Defense Agency 107 Pleasant Street Concord, New Hampshire 03301

s Public Service Company of Seabrook Nuclear Power Station New Hampshire cc:

Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent, City Hall Chairman l

126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03801 Town of Salisbury, MA 01950 Ms. Letty Hett Senator Gordon J. Humphrey Town of Brentwood ATTN: Tom Burack RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C.

20510 Ms. Roberta C. Pevear Mr. Owen B. Durgin, Chairman Town of Hampton Falls, New Hampshire Durham Board of Selectmen Drinkwater Road Tcwn of Durham Hampton Falls, New Hampshire 03844 Durham, New Hampshire 03824 Ms. Sandra Gavutis Charles Cross, Esq.

Town of Kensington, New Hampshire Shaines, Mardrigan and RDF 1 McEaschern East Kingston, New Hampshire 03827 25 Maplewood Avenue Post Office Box 366 Portsmouth, New Hampshire 03801 Chairman, Board of Selectmen Town Hall South Hampton, New Hampshire 03827 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention Mr. Angie Machiros, Chairman Conmittee Boardlof Selectmen c/o Rye Town Hall for the Town of Newbury 10 Central Road Newbury, Massachusetts 01950 Rye, New Hampshire 03870 Ms. Cashman, Chairman Jane Spector Board of Selectmen ~

Federal Energy Regulatory Town of Amesbury Commission Town Hall 825 North Capital Street, NE Amesbury, Massachusetts 01913 Room 8105 Washington, D. C.

20426 Honorable Peter J. Matthews Mayor, City of Newburyport Mr. R. Sweeney Office of the Mayor New Hampshire Yankee Division City Hall Public Service of New Hampshire Newburyport, Massachusetts 01950 Company 7910 Woodmont Avenue Mr. Donald E. Chick, Town Manager Bethesda, Maryland 20814 Town of Exeter i

10 Front Street Mr. William B. Derrickson Exeter, New Hampshire 03823 Senior Vice President Public Service Company of New Hampshire Post Office Box 700, Route 1 Seabrook, New Hampshire 03874 NAR J 51956

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1 i-i EGG-NTA-7156 AUDIT OF THE PUMP AND VALVE OPERABILITY ASSURANCE PROGRAM F0R THE SEABROOK GENERATING STATION UNIT 1

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Docket No. 50-443 e

C. Kido H. M. Stromberg

.T. C. Chung Published February 1986 NRR and I&E Support Branch EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 1

Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6415

ABSTRACT-The Seabrook Station Unit 1 was audited November 5 to 8, 1985 to datermine the adequacy of their Pump and Valve Operability Assurance Program. Ten concerns (five specific and five generic), which could not be rcsolved'by the close of the audit, were identified to the applicant; he i

committed'to address these concerns prior to fuel load. The results of this audit indicate that the applicant has established and is implementing I

a program that will track all pumps and valves important to safety from manufacture and in-shop testing through qualification, installation, testing, maintenance, and surveillance for the purpose of assuring continued. operability of these components over the life of the plant.

FOREWORD This report is supplied as part of the " Equipment Qualification Case l

R2 views" project that is being conducted for the U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation, Division of Engineering, Equipment Qualification Branch by the Engineering Analysis Division of EG&G Idaho, Inc.

P The U.S. Nuclear Regulatory Commission funded this work under the authorization, B&R 20-19-40-41-2, FIN Number A6415.

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SUMMARY

The Pump and Valve Operability Assurance Review Team (PVORT),

comprised of one member of the Nuclear Regulatory Commission (NRC) staff h-and three EG&G personnel, conducted an on-site audit of the Seabrook Pump and Valve Operability Assurance Program during the week of November 5 to

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8, 1985. A representative sample of active pumps and valves was selected for review and evaluation. These components are categorized as either Nuclear Steam Supply System (NSSS) or Balance of Plant (B0P), based upon which organization was responsible for the purchase and installation of the component. Westinghouse is Seabrook's NSSS vendor while United Engineers and Constructors Power Corporation, an architectural engineering firm, is responsible for the BOP components.

The process used to evaluate the plant's ov'erall Pump and Valve Operability Assurance Program includes:

(a) becoming familiar with each selected component and the system in which' it is installed, (b) understanding the component's normal and safety function, (c) visually inspecting the component's configuration and mounting, (d) reviewing those d:cuments relating to the operability of each selected component, (e) ensuring the applicant has an adequate document retrieval system, and

- (f) reviewing the applicant's preoperational testing and maintenance / surveillance programs.

The results of the evaluation process are two-fold.

Any component specific deficiencies or concerns are identified and documented.

Of greater importance are any generic concerns, which may be identified, that could affect other components in the plant or possibly even extend to other 3

plants.

During the PVORT review..a number of component specific concerns were raised.

All but five of these specific concerns were satisfactorily resolved during the audit by the applicant supplying additional information or demonstrating that admin-istrative procedures were in place that would iii

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address them. The applicant committed to resol've these f,1ve component specific concerns prior to fuel load.

In addition, the staff also requests that prior to fuel load the applicant confirm that:

(a) all pre-service tcsting that is required to be completed is completed. -(b) all pumps and

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valves important to safety are qualified (c) the maintenance procedures are ccnsistent with manufacturer's recommendations and provide stveral maintenance procedures for review, (d) the FSAR indicates all active BOP valves are ccvered by the Seabrook pump and valve operability assurance program including valves two inches and smaller and (e) all active valves are correctly identified in the FSAR.

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CONTENTS ABSTRACT..............................................................

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FOREWORD..............................................................

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SUMMARY

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1.

INTRODUCTION.....................................................

1 2.

EVALUATION METHODOLOGY...........................................

3 2.1 Nuclear Steam Supply System (NSSS) Components..............

6 2.1.1 Centrifugal Charging Pump, CS-P-2B.................

6 2.1.2 Power Operated Relief Valve (PORV), 456A...........

9 2.1.3 Cold Leg Injection /RHR Return Line Isolation Valve, RH-V-14.....................................

11 2.2 Balance of Plant (BOP) Components..........................

13 2.2.1 Feedwater System Controlled Check Valve, FW-V-331....................................

13 2.2.2 Turbine Driven Emergenc'y Feedwater Pump, FW-P-37A.....................................

14 2.2.3 Primary Component Cooling Water Radiation Monitor Isolation Valve, CC-V-975..........................

18 2.2.4 Feedwater Isolation Valve, FW-V-48.................

19 2.2.5 Primary Component Cooling Automatic Containment Isolation Valve, CC-V-122..........................

20 2.2.6 Cooling Tower Pump, SW-P-110A......................

22 2.3.

Other Equipment Qualification Issues.......................

24 2.3.1 Safety Evaluation Report (SER) Items'...............

25 2.3.2 Long Term Operability of Deep Draft Pumps..........

29 2.3.3 Implementation of the Overall Program 30 3

CONCLUSION.......................................................

34 4.

REFERENCES (NSSS COMPONENTS).....................................

40 5.

REFERENCES (BOP COMPONENTS)......................................

43 y

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. TABLE 1.

Pumps and Valves Selected for the PVORT Audit....................

4 2.

Status of SER Items for Pump and Valve Operability Assurance.....

26 3.

Summary of PVORT Audit...........................................

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AUDIT OF THE PUMP ANO VALVE OPERABILITY ASSURANCE PROGRAM FOR THE SEA 8 ROOK GENERATING STATION UNIT 1 1.

INTR 0' DUCTION The Equipment Qualification Branch (EQB) performed a two-step review of the Pump and Valve Operability Assurance Program being implemented by the Seabrook Station Unit 1.

The purpose of this review was to determine whether Seabrook's program is adequate to ensure that pumps and valves important to safety will operate when required during the life of the plant under normal and accident conditions.

(Seabrook is a 1150-MWe pressurized water reactor (PWR) located in Seabrook, New Hampshire.)

The first step was a review of Section 3.9.3.2 of the applicant's Final Safety Analysis Report (FSAR). This information was general in nature, and, therefore~, by itself was not adequate to properly determine the scope of the applicant's overall equip'm'ent qualification program as it pertains to pump and valve operability.

The results of this FSAR review appeared as input to Seabrook's Safety Evaluation Report (SER).

The resolution of all open SER issues was accomplished prior to or concurrently with the on-site audit.

The second step of the review was an on-site audit to assess the applicant's overall program, as it is implemented. A Pump and Valve Operability Review Team (PVORT) consisting of engineers from the EQB and the Idaho National Engineering Laborator,y (INEL-EG&G) conducted an audit from November 5 to 8, 1985, of a representative sample of installed pump j

and valve assemblies and their supporting qualification documents at the

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applicant's plant site.

Based upon the results of the FSAR review and the on-site audit, the PVORT was able to determine whether the applicant's overall program conforms to the current licensing criteria presented in Section 3.10 of the Standard Review Plan (SRP).

Conformance with SRP 3.10 criteria is required in order to satisfy the applicable portions of General Design Criteria (GDC) 1, 2, 4,14, and 30 of Appendix A to 10 CFR 50 as I

well as Appendix B to 10 CFR 50.

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r Section 2 of this report presents the basic methodology used to cvaluate Seabrook's overall equipment qualification program as well as a discussion of the concerns raised during the evaluation of the' selected components and other qualification issues.

Section 3 presents the staff's cenclusions concerning the audit.

Sections 4 and 5 present the referen'ces

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for the NSSS and BOP components, respectively.

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EVALUATION METHODOLOGY In order to evaluate the adequacy of Seabrook's Pump and Valve Operability Assurance Program and the extent to which it is being

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implemented, the PVORT conducted an audit at the plant site November 5 to 8, 1985. The first phase of the on-site audit consisted of the applicant presenting the major elements of his overall equipment qualification program. The remainder of the audit consisted of determining whether the applicable elements of the program had been (or would be) implemented for the set of selected components.

By performing a detailed review on a diverse set of components, the PVORT is attempting to identify concerns that may be generic to the applicant's overall program. Table 1 presents a list of pumps and valves selected for the PVORT audit.

As the first step of the detailed review of the selected components, the PVORT conducted a plant walkdown of each component accompanied by cognizant itcensee personnel.

One purposeof this walkdown was to obtain information that could later be compared with the evidence of qualification contained in each component's document package.

Some examples of walkdown information that was compared with relevant documents are:- (a) nameplate data versus design and purchase specifications, (b) installed configuration and mounting versus the configuration and type of mounting that was tested (or assumed in an analysis), (c) local equipment environment (including the Gnvironment that could result from an accident) versus the environment enveloped during required testing, (d) system interfaces versus energy or fluid requirements, and (e) installed functional accessories versus actual equipment tested.

In addition, a second purpose of the walkdown was to evaluate each selected component in order to determine whether any operability concerns may have been overlooked.

Examples of such concerns are: (a) the potential for flooding, (b) component misapplication, (c) the potential for pipe whip or missile damage, and (d) the potential for

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personnel interactions that could inadvertently cause a component to become l

inoperable.

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TABLE 1 PUMPS AND VALVES. SELECTED FOR THE PV0RT AUDIT NSSS Components 80P Components CS-P-28 Centrifugal Charging FW-V-331 Feedwater System Control

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Pump Check Valve RC-V-456A Power Operated Relief FW-P-37A Turbine Driven Emergency Valve Feedwater Pump RH-V-14 Cold Leg Injection /RHR CC-V-975 Primary Component Cooling Return Line Isolation Water Radiation Monitor Valve Isolation Valve FW-V-48 Feedwater Isolation Valve CC-V-122b' Primary Component Cooling Automatic Containment Isolation Valve i

SW-P-110Aa Cooling. Tower Pump Note:

The applicant has six weeks to prepare document packages for all but the surprise components; for those he has only a few days.

The contents of the document package for the surprise components is an indicator of:

(a) the applicant's ability to retrieve documents in a timely manner, and (b) the completeness of his central files.

h a.

The applicant provided a separate presentation concerning the deep draft pump issue (refer to IE Bulletin 79-15) for this component.

b.

Surprise component--The applicant is informed of this component only a few days prior to the on-site audit.

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The document review portion of the audit was conducted after the completion of the applicant's program presentation and the walkdown of the salec'ted components.

One purpose of the document review was to verify that the principles established in Seabrook's~ program had been (or would be) uniformly implemented.

Therefore, the document package for each of the audit components was reviewed to ansure that, as a minimum, each' package

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contained the following:

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A purchase specification that reflects design and functional requirements o

Results of applicable in-shop tests o

Evidence that the component was subjected to a qualification

-evaluation that addressed:

Pre-aging 1

Significant aging mechanisms (if applicable)

Normal and accident. loads (including seismic and hydrodynamic loads)

Acceptance criteria requiring operability both during and after an event Identifiable safety margins (difference between design basis parameters and the test parameters used for equipment qualification) o Applicable preoperational test procedures f

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Similarity statements, where the qualification of a similar equipment is used-to qualify the installed equipment (if applicable) 4 i

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o Evidence that maintenance / surveillance practices incorporate qualification and operability concerns.

In addition, a second purpose of the document review was to ensure that an

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auditable link existed between the documents in the package and that all documents had been reviewed and approved by personnel having a working knowledge of equipment qualification issues and concerns.

Those documents

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n;t present in the audit component document package were requested by the PV0RT. Seabrook's timely response to these requests and their. ability to compile a complete package for the surprise components were considered to j

b1 positive indicators of the acceptability of the applicant's central file system.

The remainder of Section 2 is devoted to discussing any concerns j

raised by the PVORT as a result of the equipment and issues reviewed during the on-site audit.

Sections 2.1 and 2.2 present the evaluation of the NSSS l

and 80P components, respectively.

Section 2.3 summarizes the status of other equipment qualification issues relating to pump and valve operability.

2.1 Nuclear Steam Supply System (NSSS) Components

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2.1.1 Centrifuoal Charaina Pumo. CS-P-28 (Audit Status:

Closed) 2.1.1.1 Component Description.

This component is a single speed, l

horizontal, eleven stage centrifugal pump manufactured by Pacific Pump (Model 2-1/2 RL IJ) which is driven by a 600 HP induction motor manufactured by Westinghouse (Model Life Line D).

The component is part of the chemical volume and control system and is located in the Auxiliary l

Building at the 7-ft level.

During normal operation, the pump maintains a programmed water level in the pressurizer by pumping purified reactor flow

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from the volume control tank reactor coolant to the RCS after heating via I

the regenerative heat exchanger.

Upon receipt of a safety injection signal, the two centrifugal charging pumps are automatically aligned to take suction from the refueling water storage tank and then pump borated water to the RCS.

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.y The charging pump subsystem of the CVCS is an integral part of the ECCS and must be capable.of providing long-term cooling for one year.

There are three charging pumps (one positive displacment and two

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czntrifugal) only one of which is required to handle normal charging flow.

The pump is required to be operable for 1 year post-accident.

2.1.1.2 Component Walkdown. The walkdown of this component revealed fcur anomalies, all of which were resolved prior to the close of the audit.

First, the oil level glass for the gcar box did not indicate the presence of any oil. Also, the vertical oil level glass on the pump sump was 2/3 full but there weren't any markings on the glass casing to indicate sinimum and maximum fill levels. A Westinghouse engineer explained that the pump assembly including gear box is adequately lubricated if any oil is d3tected in the sump oil level glass.

The glass is nermally filled to the halfway position to allow f,or thermal expansion.

The oil level glass on the gear box is located above the normal oil level of the sump.

Consequently, the proper oil level is determined by the sump oil level glass only.

Second, there was a loose wire coming from the rear motor b2aring box.

Documentation was provided to demonstrate that the work to repair the temperature element will be completed before the pump is placed

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in operation.

Third, the boror. injection tank (BIT) was removed completely, which was not shown on the FSAR drawings. The BIT was removed by the applicant because its additional boron concentration was not censidered to be necessary for plant operation.

Documentati'on was reviewed which authorized the modification, specifying those systems and equipment that were affected.

And fourth, the miniflow valves (2"-CS-196 and -197) were shown in parallel in the FSAR drawing, but were shown in series in the Westinghouse drawing. The startup engineer explained that United Engineers took over responsibility for this section of piping from Westinghouse in crder to improve isolation reliability of the miniflow line.

Docun!entation was reviewed which demonstrated that the as-built piping has been qualified by analysis.

In the clarification of these last two items, Seabrook personnel stated that the FSAR is being scrutinized for consistency and that the anomalies in drawings, tables, and text will be resolved in future amendments.

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1 2.1.1.3 Document Review.

The review of the qualification documents revealed that. qualification of this component was addressed by a combination of tests, analyses, and similarity statements.

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It was discovered that the motor was qualified by a generic test program and that the qualified life was given as 5 years.

Documentation was reviewed which listed the various motor types that were covered by 4he generic environmental qualification program. A 450 HP motor stator was used to determine the 5 year qualified life based upon the testing of the thermalastic epoxy insulation.

Draft procedures were already written to maintain qualification of the lubricant and motor per vendor recommendations.

However, a complete set of approved maintenance procedures were unavailable for review. This concern was brought to the applicant's attention as a generic issue.

See Section 2.3.3 for a discussion of the Seabrook maintenance program. The FSAR description was not clear whether LOCA loads were applicable for this pump. Westinghouse l

oxplained that the pump is located outside containment and that the major LOCA effects would be damped out at the penetration.

In addition, the l

generic nozzle loads, used to qualify the pump, are significantly higher than the plant specific loads and provide sufficient margin to envelop residual LOCA effects.

An identical charging pump assembly was mounted to a shake table and operated under full flow but reduced pressure conditions. Operation of the pump was limited to 150 psi to avoid overpressurizing the test loop piping.

Generic nozzle loads and 2.1 g seismic loads were applied while the pump was operated.

The 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months endurance run was comprised of 10 ten-hour runs, each of which included at least one hour of full flow operation. The test results did not detect any degradation in vibration levels or bearing temperature.

Similarly, the motor insulation and lubricants did not degrade from the prolonged pump operation at runout conditions.

Construction-tests have already been performed as required and the hot functional tests were still in progress at the conclusion of the audit.

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2.1.1.4 Findines. No specific operability concerns remained after the evaluation of this component.

2.1.2 Power Operated Relief Valve (PORV). 456A. (Audit Status:

Closed)

>= -

2.1.2.1 Component Description. This component is a solenoid controlled, 3 x 6 in. plug-type, inlet pressure-operated shutoff valve 4

manufactured by Barrett (Model 3750014). The valve is located on the top h:ad of the pressurizer 'and inside containment at the 56.5 ft level. The PORV is designed to flow any combination of air, water, or steam at inlet conditions up to 2600 psig at 700*F.

Valve position indication is achieved through the use of four single-pole, double throw switches. The solenoid ccntrol valve is a continuous-duty, direct-acting, three-way solenoid valve designed and constructed to meet the requirements of the ASME code, S2ction III, Class 1.

The normal function of the PORV is to control pressurizer pressure. The safety function is to prevent the safety valve from lifting, as well as to prevent reactor trip on high pressurizer pressure and cold overpressure mitigation.

~

The PORV is operated automatically or by remote ; control.

Steam from the PORV is discharged into the pressurizer relief tank, where it is cendensed and cooled by mixing with water near ambient temperature.

The valve is required to be capable of operation up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months post accident.

2.1.2.2 Component Walkdown.

At the time of the audit, the hot functional tests were still in progress.

Access to the PORV was determined to be difficult and potentially dangerous.

Instead, recent photographs taken by the utility prior to the hot functional tests were studied. The photographs showed that the valve body was wrapped in insulation, while the valve solenoid and limit switches were exposed.

Westinghouse engineers pointed out that the PORV had been recently modified to eliminate leakage past the body-to-bonnet gasket. The original design used a vent through the bonnet flange connecting the upper pressure chamber of the bonnet with the discharge port of the valve.

The leakage occurred at the point of the brdy-to-bonnet gasket which sealed the vent opening.

The new modification 9

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w.

used a stainless steel p1pe welded to the vent' opening in the valve body.

The-bonnet flange was drilled to allow the pipe to pass entirely through the flange, avoiding the gasket area. Then, a separate piece of threaded' tubing was used to connect the exposed pipe with the upper pressure chamber of the bonnet. This configuration was requested by Seabrook specifically for the hot functional test sequence. Westinghouse presented test results which demonstrated operation of the PORV at the design settings.

No cxternal leakage at the body-to-bonnet joint was detected.

Although the Westinghouse test report had not yet received final approval Seabrook did review the preliminary test results and did accept the PORV in order to conduct the hot functional tests as scheduled.

In general, the modification to the PORY did not invalidate the' original qualification stress analysis of the pipe, demonstrating that the new tubing configuration was acceptable.

Procedures have been written to assure that new pipe will not be damaged whenever the bonnet is pulled.

2.1.2.3 Document Review.

The review of the qualification I

documents revealed that qualification of this component was addressed by a combination of tests and analyses. Minor discrepancies in the PVORT long form were resolved by discussion with Westinghouse personnel

- and substantiated by the appropriate doctmentation.

A static deflection test was conducted without any problems at a generic acceleration of 7.75 g compared to the plant specific load of.06gH1,.43gH2,.10gV.

The solenoid was qualified in accordance with the IEEE 323 test sequence (baseline l

parameters, mechanical and thermal aging, containment pressure, radiation.

vibration aging, seismic, and LOCA conditions).

The qualified life of the solenoid was calculated as 12.7 years, although the vendor recomends replacement after 5 years.

Seabrook personnel indicated that the vendor l

recommendations will be incorporated into the maintenance program once the qualification documents have been transferred from the startup group.

Although the methodology for implementing the maintenance program appears to be established, the specific procedures have not been written.

This concern was found throughout the audit and was elevated to a generic issue at the close of the audit.

See Section 2.3.3 for a discussion of

~

this concern.

10

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2.1.2.4 Findinas. No specific operability concerns remained after the evaluation of this component.

2.1.3 Cold Leo Injection /RHR Return Line Isolation Valve. RH-V-14.

(Audit Status:

Closed) i 2.1.3.1 Component Descriotion. ThiscomponeStis,an8-inchgate valve manufactured by Westinghouse (Model 08002GM88FN8000) powered by a Limitorque motor operator (model S8-1-60). The valve is located in the auxiliary building at the -18.5 ft level. The valve is normally open in the discharge line from the residual heat removal pump downstream of the RHR heat exchanger.

One safety function of the valve is to open for cold leg injection and recirculation. The other safety function of the valve is to close for containment isolation and hot leg recirculation. There are redundant torque switches to prevent the actuator from excceding the specified torque setting.

Likewise, there are redundant limit switches which read the linear travel of the valve' stem to ensure full open and closed positions.

Upon loss of power the valve will fail as is, which is the fail-safe position. The valve is required to be operable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

- ' af ter event initiation.

2.1.3.2 Component Walkdown.

The wal'kd'oyn of this component revealed two anomalies, both of which were resolved prior to the close of the

~

audit.

First, a ventilation system (HVAC) duct was located a foot away from the side of the motor. The PVORT asked what might be the consequences of a cooling water coil in the HVAC system rupturing, ar.d having that water blown directly into the motor. The applicant explained that there were no cooling coils in that portion of the HVAC system.

Second, the valve and P

motor were installed vertically in the pipeline without any additional lateral support.

Documentation'was reviewed which demonstrated that the stress level due to faulted conditions were acceptable for the as-built ccnfiguration.

2.1. 3. 3 Document Review. The review of the qualification documents (

revealed that the qualification of this component was

~

addressed by a combination of tests, analyses, and similarity statements.

11

~.

f The Limitorque SB-1-60 operator was included in the gener,1c design group of

~

operators that were qualified by type testing.

In particular the SMB-00-15 and SMB-1-60 operators net all acceptance criteria associated with the type test sequence in IEEE 323-1974. The various models within the design h

series have identical enclosure assemblies, gasket assemblies, electrical contact assemblies, internal wiring, and materials of construction. The e

only difference between these operators involve the spring mounting configuration, direction and length of travel, and type of travel.

E A stress analysis of the valve assembly met the criteria of the 1974 ASME code,Section III.

Functional tests were successfully completed on 4 and 12 inch valves of a similar design. The cold cyclic tests recorded cycling times within allowable limits.

However, the opening and closing current reading was 13.8 amp compared to the nameplate rating of 12 amp.

The applicant justified the apparent discrepancy by providing a Limitorque letter which indicated that the full load current shown on the motor nameplate represents a current value equivalent to 20 percent of the A

motor's starting torque rating.

Furthermore, the applicant will invoke Limitorque's recommendation to inspect the valve packing and stem lubrication if the current drawn ever exceeds the 120 percent of the

~

[

nameplate rating. The yoke-mounted external limit switches were upgraded by Westinghouse to meet Class 1E requirements. The new switches were shop tested for seismic and environmental conditions, and found to have a qualified life of 10 years.

Seabrook confirmed that the maintenance procedures will include the vendor recommendations for replacement.

The above evidence, as well as completion of the construction tests, provide confidence that the val,vo will operate as required.

2.1.3.4 Findinos.

No specific operability concerns remained after the evaluation of this component.

S 12

o,.

... ~.

s 2.2 Balance of Plant (BOP) Components 2.2.1 Feedwater System Controlled Check Valve (FW-V-331) (Audit Status:

Closed Pendino Resolution by Apolicant) 2.2.1.1 Component Description. This component is an 18-inch, L

Y-pattern, piston-type control check valve manufactured by the flow control division of Rockwell International Corporation (Model 18 x 16 x 18 Fig 2092

/

l (WCC) SJQTY). The valve '.s buttwelded to the 18 inch pipe as a part of the Fesdwater Feed to steam generator "B".

The valve assembly is located'in the main steam and feedwater pipe chase building at the level of 8 ft-3 in.

l (centerline of pipe).

It is a normally open valve.

The safety function is to close at a controlled rate when the emergency 'feedwater pump is activated or in the event of feedwater line break upstream of the check valve.

It is held closed by back pressure to prevent flow in the reverse direction.

Yalve operation is timed open and closed by an internal dashpot formed by the design of the check piston.

2. 2.1. 2 Component Walkdown.

The valve was covered with insulation.

There were no visible anomalies found during the walkdown.

m

2. 2.1. 3 Document Rev'ew. The review of the qualification

' revealed '. sat the operability was demonstrated by

~

documents analysis and limited testing. The documentation review encompassed design and qualification documents.

The review determined that the component had been designed for operating conditions of 1011 psig and 445'F. The qualification documents provided the results of the stress, vibration and operating time analyses. The stress analysis accounted for normal, upset, emergency and faulted conditions stress loading. The stress analysis was s

found acceptable. The vibration analysis determined that the fundamental frequency is greater than 33 Hz.

Therefore, no exploratory vibration analysis or testing was performed.

The operating time analysis determined valve closing time to be 1.26 seconds which is acceptable, compared with the specified minimum closing time of.8 seconds.

However, there was no testing performed to demonstrate timing of the valve as it is installed in 13

the plant. There remains some question as to whether operability can be demonstrated by analysis only. This concern was brought to the applicant's attention as a component specific issue, which must be resolved prior to fuel load. The applicant has committed to evaluating timing requirements and performing testing as required. The applicant has also committed to providing IST training requirements as required.

2.2.1.4 Findinas. The valve was qualified by analysis only. There were no instrumented test results to verify the valve closure time, and there is no reported operational experience at other nuclear plants subject to similar design conditions. Th,erefore, we don't feel the qualification by analysis alone can assure the operability of the valve to mest its I

design requirements.

The applicant is requested 16 address this issue by the following actions:

a.

Demonstrate the valve performance meets the design specifications through means other than the analysis already performed.

.b.

Provide copies of the IST procedures which demonstrate the verification of the valve closure time according to ASME code Section XI requirements to assure its continual operahility.

2.2.2 Turbine Driven Emergency Feedwater Pumo. FW-P-37A (Audit Status:

Closed Pendino Resolution by the Applicant) 2.2.2.1 Component Description.

This component is a ten stage centrifugal pump manufactured by Ingersoll-Rand (Model NH-10) driven by a 770 hp steam turbine manufactured by Terry Corporation (Model GS-2N).

The component is part of the Emergency Feedwater System and is located in the Emergency Feedwater Pump House. There are two pumps in redundant loops; one is steam turbine driven and the other electric driven with both pumps in standby during normal operation.

The pump's safety function is to start and provide feedwater to the steam generators in the event of a line break, loss of main feedwater er reactor-turbine trip.

14

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2.2.2.2 Component Walkdown.

During the walkdown of the auxiliary feedwater pumps, eight operability concerns were identif1ed.

Four concerns were resolved before the close of the audit and.four are items that require rcsolution by the applicant.

{

The items that were resolved during the audit were; operability of the lube oil filter during equipment operation, equipment danger tags on the turbine steam supply isolation valve and the pump feedwater suction and

{

discharge valves, caution tags on the pump pipe hangers, and the pump to turbine coupling and thermocouple had been removed. The lube oil filter concern was addressed by demonstrating that the filter would not have to be cperated during pump operation as the filter would be automatically bypassed if it became clogged,.therefore its operability is not required.

The danger tag installation was resolved by demonstrating that they had been installed to provide a safety boundary for hot functional testing.

Pipe hanger caution tags had been installed to indicate that the hanger turnover was not complete.

The last concer'n resolved during the audit was I

that the coupling and thermoccuple had been removed.

Thus was resolved by i

demonstrating that they were removed to facilitate a cracked pump seal l

replacement.

l There were four concerns that were not resolved; trip and throttle valve operability, pump turbine end seal was cracked, the governor was not qualified and an unqualified modific~ation had been made to the turbine steam piping.

One of the main concerns identified during the component walkdown was that the turbine trip and throttle valve installation was not made in a way that allowed easy operation by a plant operator. The installation had two problems. The first is that the manual operated valve was installed at an elevation that would not allow the operator to deliver a great deal of o

torque to the handwheel.

The second is that the valve was installed in a confined space where only one operator would have access to the handwheel and then in close quarters, this makes operating the valve even more difficult. The applicant could not provide information that demonstrated 15

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,,.,. ~

~

whether or not the valve installation would cause operation to be difficult. The licensee comeitted to perform testing that would demonstrate the ease of valve operation and then evaluate the results and g

requirements.

Another problem that was found during the walkdown was that the a3x111ary feedwater pump turbine end seal had been cracked and was in the i

process of being changed. Upon questioning it was found that the cause of the failure had not been determined. The applicant has committed to d3termining the cause of the failure and taking steps to prevent a recurrence should any be required.

The third concern identified during the walkdown was that the turbine g;vernor had a hold tag attached to it. The questioning determined that the governcr had been qualified, then a modification had been made that made its operation a little easier, however, the modification had removed i

the qualification status.

The hold tag had been installed indicating that the qualification had not been completed. The applicant has committed to

providing confirmation that,the governor qualification is complete prior to i

fuel load.

The fourth concern found during the walkdown was that an unqualified pipe installation had been made at the governor end of the auxiliary feedwater pump turbine. 'The questioning determined that during earlier testing, a great deal of steam leakage had been experienced from steam pipe drains. The installation was made to provide a method of sealing the

~

leakage to prevent it~from being vented to the Emergency Feedwater Pump H:use atmosphere.

The applicant was not sure that the installation would work as designed, it is their intent to demonstrate operability during pump testing. 'If the installation works, the licensee will seismically qualify the installation, however, if the install & tion does not work, it will be removed and another modification tried. The applicant has committed to providing a description of the final installation and confirmation that it meets qualification requirements.

16

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2.2.2.3 Document Review. The review of the qualification documents (32-37) indicated that qualification of this component was addressed by a combination of tests and analyses.

There.were two concerns identified during the document review, one was left open as a confirmatory issue and one resolved during the audit.

During the document review, it was found that the applicant's qualification of the Terry turbine did not address turbine operation when there is moisture in the turbine driving steam.

The applicant has committed to evaluating moisture in the turbine driving steam and providing confirmation that their installation will work as intended.

The second concern identified during the document review was that the seismic testing performed by Wyle Laboratories identified some mounting bolts inside of the turbine as being bolts that could work loose during a seismic event. Terry Corporation performed'an evaluation and determined that the bolts of concern required torquing at least once every five years with an application of Loktite 277. The applicant was questioned about

~

whether or not this requirement had been addressed. A draft maintenance package was provided.

Upon review, it was found that the torquing f.'.

requirement had not been identified. The applicant then described the methodology used to develop the maintenance requirements. The methodology demonstrated that the torquing requirement would have been identified during the maintenance package development and sign off, which was found acceptable.

2.2.2.4 Findinas.

During the review of this component, five items were found to be of concern:

1.

Turbine trip and throttle valve operability 2.

Pump seal failure evaluation 3.

The turbine governor was not qualified 17

4.

Steam pipe drain modification was not qualified 5.

Turbine qualificatica did not address moisture in the steam.

Each item is described in detail in Se'ctions 2.2.2.2 or 2.2.2.3.

t 2.2.3 Primary Component Cooline Water Radiation Monitor 1.rolation Valve.

CC-V-975 (Audit Status:

Closed) 2.2.3.1 Component Description.

This component is a one inch plug

~

valve manufactured by Tufline/Xomox (Model CL-150) which has a 6. H. Bettes Ccrporation pneumatic operator (Model CB 4205R60).

The component is a valve in the primary component cooling water system supply line to a radiation monitor and is located in the Primary Auxiliary Building. This valve is normally open to allow water flow.

The valve's safety function is to isolate the radiation monitor from the component cooling water system when non-essential components are not required.

Thus providing more component cooling to the essential equipment when required and helps maintain boundary integrity of the essential portion of the primary component cooling water system.

2.2.3.2 Component Walkdown.

The walkdown of this component identified four minor anomalies, all of which were resolved prior to tne close of the audit. The minor deficiencies involved a hold tag on the i

valve's local control panel, valve position switch that was not wired into service, holes in the bottom of the electrical junction boxes, and covers out of place on a conduit union and main wiring box.

1

~The applicant demonstrated that the hold tag on the control panel was the result of a light cover having been broken and then replaced.

The tag had not been removed as QA had not inspected the replacement. The applicant investigated the switch that had not been wired into service.

They determined and provided evidence that the switch was not required but was provided with the valve-assembly package as a normal part and therefore, left in place.

The applicant researched the holes in the bottom 18 m

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j

.of the junction boxes and found that they had been provided to -allow moisture to be removed. The connections inside are waterproof, therefore, l

the holes did not have a direct impact on operability. The last deficiency.

was investigated and it was found that wires for other installations as

~ i well as this component were run through the conduit and junction box. Work was being performed on the other components, which was the reason the covers had been removed. The applicant indicated that the covers would be replaced when the work war complete.

2.2.3.3 Document Review. The review of the qualification i

I I

d:cuments revealed that operability had been addressed by the combination of tests and analyses.

The document review identified one concern which was addressed before the end of the audit. The concern was that the solenoids on the valve had been replaced. The investigation determined that the solenoids had been replaced as those provided by the vendor were not the correct ones.

2.2.3.4 Findinas.

No, specific operability concerns remained af.ter the evaluation of this component.

2.2.4 Feedwater Isolation Valve. FW-V-48 (Audit Status:

Closed) f,

~

2.2.4.1 Component Description. This component is an eighteen inch gate valve manufactured by Borg-Warner (Model 73890) with a pneumatic-hydraulic operator manufactured by Borg-Warner (Model 37951).

The component is part of the main feedwater system and is located in the j

MSFW Pipe Chase. The valve's normal position is open to allow feedwater b

flow to steam generator RC-E-11C. The safe'ty function of this component is to close and isolate the main feedwater system.

2.2.4.2 Component Walkdown.

During the walkdown cf this component, two minor concerns were identified which were resolved prior to the close of the audit. The concerns involved some grounding cables that were cissing.

During discussions with the applicant it was identified that the 19

~

l valve had been removed, overhauled and replaced. The app 11 cant determined that the grounding cables were not required in accordance with design.

In addition, the applicant indicated that the valve had been removed and overhauled in accordance with manufacturer's five year maintenance

~~

requirements.

2.2.4.3 Document Review. The review of the qualification I

-' I documents

-indicated tr.at qualification of this component was addressed by a combination of analyses and tests.

The document review identified one concern which was resolved befort the end of the audit. The item involved oil-changing requirements.

During t: sting of this valve in a high moisture environment, the valve operation exceeded operability criteria. The problem was researched and it was found that moisture accumulating in the oil caused valve operation to slow down.

To address the problem, the applicant in agreement with the operator 4

manufacturer has initiated maintenance requirements which will monitor the moisture content of the oil.

Review of the maintenance procedure indicated that the applicant intends to change oil every two years and analyze the I

oil for moisture whenever operating time exceeds operating limits.

l 2.2.4.4 Fii.dinos.

No specific operability concerns remained after evaluation of this component.

2.2.5 Primary Component Coolina Automatic Containment Isolation Valve. CC-V-122 ( Audit Status:

Closed) l I

2.2.5.1 Component Description. This component is a 12-inch butterfly

~~

valve manufactured by Posi-Seal International (Model 150) with an air actuator manufactured by Matryx (Model 26072SR60).

The component is located in the mechanical penetration area at the 6 ft level.

The valve',s nsrmal position is open to pass cooling water flow. The safety function is to close and provide system isolation on a "P" containment isolation signal or a low-low head tank water level signal.

20

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..w, The valve position is indicated on the MCB by status light and can be cperated remotely by the manual control from the MCS and the remote safe shutdown panel.

In case of loss of air, the valve closes to its safe pssition.

In the. event of signal loss, the valve fails as is.

2.2.5.2 Component Walkdown. The walkdown of this component revealed three minor concerns; the valve nameplate data tag was unreadable, actuator a

air supply lines were not clamped to the holddown brackets and there was water laaking from the valve flanges.

Upon investigation of the nameplet's data tag, the system test engineer (STE) pointed out that an ideatification number had been stamped into the valve body. This stamped number could be referenced back to the valve's name plate information.

The STE also investigated the lack of attachment of the air supply lines to the hold down brackets.

He explained that the lines were loose-to provide easy access for the on-going system tests during the start-up test program. The lines will be, permanently supported before the compartment

, walkdown program starts.

Investigation of the water leakage problem identified the cause as a faulty gasket.

Reviewing the documentation confirmed that a work request had been issued to correct the problem.

2.2.5.3 Document Review. The review of the qualification

~

documents indicated that this component has been qualified by a combination of tests and analyses.

During the document review, two concerns were identified, both of which were adequately resolved during the audit. The first concern was that the fundamental frequency was found to be 11 hz.

Further review found that the low frequency was the result of the actuator having a large overhanging section. To increase the fundamental frequency above 33 hz, the applicant had installed an additional support to the overhanging 21

l section. This was found to be. acceptable. The second concern was.that the air supply solenoid valves had been replaced.

It was fou'nd that the scienoid valves originally had 120 VAC solenoids installed where 120 VDC was required. The change was made to correct the deficiency.

i The preoperational leak test procedures were reviewed and found to be adequately prepared. There were no concerns identified during this review.

The final area. investigated during the component document review was the aging evaluation'. There were three aging mechanisms identified; thermal, wear and radiation. These mechanisms impacted two portions of the valve ard actuator assembly, the solenoid valves and valve seals.

The replacement of these components was addressed in replacement procedures.

2.2.5.4 Findinos.

No specific operability concerns remained after the evaluation of this component.

2.2.6 Coolina Tower Pumo. SW-P-110A. (Audit Status:

Closed Pending Resolution by the Applicant)

I 2.2.6.1 Component Description.

This component is a two stage, vertical, centrifugal pump manufactured by Johnston Pump Company (Model 33 NLC) and is driven by a vertical, induction, 800 HP motor j

manufactured by General Electric Company (Model SK6339XC179A).

The pump is located in the cooling tower at the 46 ft level.

The pump's normal state is standby for auto-start.

Its safety function is to provide cooling water t

from the ultimate heat sink, a 4 million gallon basin that is completely t

l independent of the circulating water tunnels and Atlantic Ocean.

Transfer of heat loads to the' ultimate heat sink can be performed manually from the main. control board or by a low pressure signal indicative of a low-low service water pumphouse level.

The pump is required to be operable for 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months post accident.

2.2.6.2 Component Walkdown.

The walkdown of this component revealed several equipment tags attached to the motor and pump, all of which were satisfactorily explained by plant personnel.

Numerous are strikes on the 22 L-

1 pump discharge head, as well as some carbon contamination, wirre discovered during the construction phase. The appropriate non-conformance reports were written and work was done to restore the material to its original

. qualified state. A tag attached to the bearing cooling line specified

~ '

inspection of the thermocouples to verify proper resistance of the temperature elements.

Documentation indicated that this work has been completed. The work specified by these tags also involves other plant squipment. All. tags will be removed'when the remaining work has been completed.

2.2.6.3 Document Review. The review of the qualification

~

d:cuments revealed that the qualification of this component was addressed by a combination of tests, analysis, and operating experience.

The motor was qualified for mechanical load conditions by analysis.

Qualification of the motor for aging and environmental conditions was demonstrated by similarity, using generic type test results.

The pump was qualified by stress analysis, shop tests,'a'nd pre-operational tests.

' Review of the stress report revealed that the stress levels were within 7

acceptable-limits, but were,different than the values reported in the

~

FSAR.

UE&C personnel explained that a correction in the computer program as well as revised load conditions accounted for the apparent discrepancy.

During the pre-operational tests the measured flow and head of the installed pump was 13000 gpm at 160 feet compared to the factory test values of 13000 gpm at 170 feet. The corrective action taken by UE&C was to establish limits for the average basin water temperature as well as for the minimum basin water level.

In addition, UE&C engineers had determined that the flow requirements during accident conditions are 9560 gpm through the primary component cooling heat exchanger and 1800 gpm through the 5-diesel generator heat exchanger. With the 7 percent ASME Section XI wear margin included, the cooling tower pump can deliver 9660 gpm and 1800 gpm, respectively.

The FSAR and other.sarvice water system documents will be revised to reflect the reduced flow conditions.

Long-term operability was demonstrated by a 168 hour0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months continuous run.

Vibration readings, taken at the start and end of the run, did not detect any significant change.

The 1

23

F I

vibration readings were taken in 2 horizontal directions at the top and bottom of the motor casing. The applicant's response to IE Bulletin 79-15 CLong Term Operability of Deep Draft Pumps" was reviewed and found to meet the Licensing Review Group-II (LRG-II) quidelines, endorsed by NRC-EQB staff.

Discussion of the deep draft pump operability issue is presented in more detail in Section 2.3.2.

One concern was identified at the conclusion of the audit, which requires resolution by the applicant.

Lateral movement of the pump column is controlled by two brickets approximately 22 feet apart.

The l

circumferential gap between the pump column and each lateral support is L

centro 11ed by identical 0-rings. The maintenance program did not specify the procedures and schedule for replacement of the 0-rings.

In order to rssolve this issue, the applicant must provide the following information

(

prior to fuel load.

1 a.

. Provide a copy of the maintenance procedures which specify the l

replacement of the 0-rings.

Include any special considerations necessary for handling the two 0-rings.

b.

Confirm that the maintenance program for the service water pumps includes similar procedures for replacement of their 0-rings.

l 2.2.6.4 Findinas.

Except for the maintenance of the 0-rings mentioned above, no other specific operability concerns remained after the svaluation of this component.

2.3 Other Eauipment Oualification Issues This section summarizes the status of other issues relating to pump and valve operability tnat were addressed by the PVORT. The following discussions combined with the detailed review of selected equipnient provide additional basis for PVORT's conclusions concerning the applicant's overall program.

~

24

2.3.1 Safety Evaluation Report (SER) Items (Status:

Closed)

The PV0RT reviewed the Seabrook,FSAR and formulated questions and concerns that appeared in the preliminary SER dated October 10,.1982.

Additional comments were presented at the pre-audit meeting held August 7, 1985. At that meeting, the PVORT requested the applicant to provide additional information la order to better clarify his program as well as to detect ap:1 address any major deficiencies. Table 2 summarizes the status of the ten SER items.

Four of these items (1, 4, 5, and 6) were addressed adequately by the app',1c7.nt in a response dated Sep.' nber 24, 1985.*

In this letter, the applicant consnitted to provide the requested information in the form of new or amended tables and expanded discussion in the appropriate sections of the FSAR.

The remaining six items (2, 3, 7, 8, 9,. 10) were addressed during the site audit November 5 to 8, 1985.

Items 2, 3, 7, 8, 9, 10 were resolved'during the on-site audit.

Regarding Item 2, the applicant committed to provide new tables and text in his forthcoming Amendment to the FSAR.

SER Item 3 was not addressed during the audit and, therefore, it appears as generic issue 2 which the applicant has committed to resolve in a future FSAR amendment.

Regarding Item 7, the applicant stated that he did not use the guidelines of the draft standards. The applicant did, however, state that he would evaluate these standards when they were approved.

It is the PVORT's belief that Seabrook's pumps and valves do meet the requirements of the codes and standards that were in effect at the time of purchase and that the applicant's reluctance to review draft standards does not constitute a licensing issue.

Regarding Items 8 and 9, the information requested is 2

a.

Letter from R. Sweeney, Bethesda office manager, Seabrook Station Group Number SBB-85-203 to V. Nerses, NRC/DL/LB3 Seabrook Project Manager,

" Advance Copies of Annotated FSAR pages and System Turnover Status List,

September 24, 1985.

e 25

TABLE 2.

STATUS OF SEABROOK SER ITEMS FOR PUMP AND VALVE OPERABILITY ASSURANCE 4

SER Itess

,Qng$$on a

Status

1. It is not xlear that the applicant has Satisfactory Closedb e

completely qualified the emergency feedwater and fuel oil transfer pumps, based on the summaries provided the appropriate information in each table to demonstrate that these pump 7 are qualified in a manner consistent with Section 3.9(B).3.2a.

(Amendment 53)

2. It is not clear from examining Satisfactory Closede Table 3.9(B)-2 and Section 3.9(B).3.1 (Amendment 48) that LOCA loads have been specified in the design load combinations for B0P class 1 components and supports. The applicant should confirm that LOCA loads have been applied to the appropriate BOP equipment in a. manner similar to Section 3.9(N)1.6 for NSSS equipment.

3. Section 3.9(B).3.2b (Amendment 48)

Satisfactory Closede describes operability assurance for active 80P valves two inches and larger.

The applicant should include all sizes of active 80P valves in his operability assurance program.

4. The applicant should provide specific Satisfactory Closedb information for the BOP pumps and valves in a manner similar to the information provided in Tables 3.9(N)-10 and -11 for NSSS pumps and valves.

5. Table 3.9(B)-2 (Amendment 47)

Satisfactory Closedb su u rizes the load combinations for i

Class 1, 2 and 3 BOP components and

~,

supports. The applicant should identify the stress criteria used to qualify Class 1 BOP valves.

6. Tables 3.9(B)-3 and 3.9(N)-7 provide Satisfactory Closedb the stress criteria for Class 2 and 3, non-active, BOP and NSSS pumps, respectively.

The applicant should identify these non-active pumps.

26

...,.n v.

.. ~

TABL.E 2.

(continued)

Finding /

a SER Items Resolution Status

7. The applicant should clearly show the Satisfactory Closedc extent to which the regulatory positions and guidelines of RG 1.148, ANSI /ASME N551.1 draft standards, and ANSI B16.41 are met.

8. The applicant should clarify the Satisfactory Closede methods used for qualification.

Specific information should be presented in the FSAR, and be available for review at the site.

The applicant should demonstrate a) -The extent to which operational testing is performed at design basis conditions (full flow, pressure, temperature, etc.)

b) The technical basi.s for qualifying equipment by similarity analysis and prototype testing.

c) ~ Qualification of tiie equipment as an assembly rather than individual components.

9. The applicant should clearly show how Satisfactory Closedc implementation of the initial test program, maintenance and surveillance, in-service inspection, and quality assurance programs will maintain equipment operability throughout the 40-year plant life.

Specific criteria should be presented in the FSAR, and be available for review at the site.

3 10.The following actions by the' applicant Satisfactory Closedb would enhance the staff's understanding of the plant.

a) The applicant should define the Terms "DSL" and "LOCA DISPL,"

which are used in. Table 3.9(B)-6 (Amendment 48).

27

~

TA8LE 2.

(continued)

Findina/

a SER Items Resolution Status b) The applicant should specify the seismic accelerations discussed in Section 3.9(B)3.2a and describe how they were used to qualify

' rigid" and " flexible" B0P pumps.

c)

Sections 3.9(B)3.2b and 3.9(N)3.2a(2) describe BOP and NSSS programs for testing valves of various designs and sizes during simulated faulted conditions.

The applicant should describe the criteria used to select the valves for testing and specify the range of sizes that l

are covered.

d) The applicant should confirm that the evaluation of NSSS check valves will include " stress analysis of critical parts which may affect operability, including faulted condition loads," as is the case for 80P check valves.

?;

a.

The Seabrook SER items for pump and valve operability assurance were identified in an earlier SER dated October 10, 1982, and were supplemented by specific comments presented at a pre-audit meeting on August 7, 1985.

l b.

This item was adequately resolved based on information submitted by the applicant in a letter " Advance Copies of Annotated FSAR Pages and System l

Turnover Status", memorandum from R. Sweeney, Bethesda Office Manager, Szabrook Station, to V. Nerses, NRC/DL/LB3, Seabrook Project Manager,

$2abrook Group Number S88-85-203, September 24, 1985.

'~

c.

This item was adequately resolved based on information reviewed by the staff during the site audit November 5-8, 1985. The applicant committed to close out this item in a manner and time frame that is acceptable to the staff.

e e

28

indirectly referenced in F.SAR Chapter 14 (Start-up Testing) and Chapter 16 (Plant Technical S'pecifications). The applicant explained that the extent of full flow tests is difficult to describe in general terms, and must be examined on a component level. The PVORT reviewed the preoperational test procedures for selected components within the context of the audit. The applicant further described inservice test (IST) activities that cover the flow test concerns. Although the IST procedures were unavailable for rcview, the applicant stated that they will comply with the ASME Section XI requirements and will be referenced in the FSAR.

Regarding Item 10, the PVORT reviewed the methods of qualification for various components, and found them to be satisfactory. The applicant's reluctance to present the methodology in extensive details in the FSAR does not constitute a licensing issue.

In sununary, the PV0RT believes that the applicant has, by way of appropriate commitments and clarifications, adequately addressed all ten SER items as they relate to pump and valve operability.

2.3.2 Lona-Term Operability of Deep Draft Pumps. (Status:

Closed) 4 IE Bulletin 79-15 was issued July 11, 1979 as the result of

~

industry-wide problems associated with the long-term operation of deep draft pumps. Plants under construction were required to identify such pumps, provide operating history, and verify the pump's ability to operate without incurring vibration-induced problems.

At the time of the bulletin, S2abrook was in a position only to identify the types of pumps used, since cperating history was unavailable.

As a followup to their original response, the PVORT asked the applicant to review and compare his deep draft pump qualification program to the NRC's suggested guidelines contained in a memorandum regarding the Licensing Review Group-II L

Issue 9-RSB. The applicant stated that long-term operability of the service water and cooling tower pumps is demonstrated by (1) using the v::ndor recommended installation procedures; (2) testing and verifying design fe'atures; (3) an extensive running period (2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months ) of the i

cooling tower pumps and the continuous operation of the service water pumps; and (4) the ability to perform post-accident maintenance and repair of these pumps, Monthly surveillance testing and vibration measurements 29

=

for each pump will be conducted following completion of the preoperational tests. Subsequent to an OL issuance, the surveillance t'esting and vibrationmeasurementswillbeconductedpe}' ASP.E.SectionXIinservicetest requirements. Long term operability (over 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months c'ontinuous run) has already been achieved for both cooling tower pumps without any significant degradation of design parameters. The two service water pumps will be similarly operated for routine system flush tests and start-up activities prior to fuel load. In summary, the PVORT believes that the program

. described by the applicant meets the intent of the NRC's suggested guidelines for long-term operability of deep draft pumps.

2.3.3 Imolementation of the Overall Program. (Status:

Closed Pending 1

Resolution by Apolicant)

The PVORT's evaluation of the applicant's overall qualification program was based on many factors, including the FSAR review, resolution of SER items, pre-audit correspondence, and t'he on-site review of selected squipment.

Another important factor was the follow-up evaluation of the applicant's administrative programs that are linked to equipment qualification. The PV0RT evaluated these programs during the on-site audit. This. evaluation enabled the PVORT to gain a better perspective of the programmatic scope and implementation of the applicant's overall cquipment qualification program.

For example, the PVORT's questions concerning the equipment tags 6bserved during the walkdown, resulted in a brief discussion of the applicant's tag management procedures and system turnover log. Similarly, the PVORT's concern about deep draft pump operability led to discussions of the applicant's in-service test procedures, preventive maintenance procedures, and quality control l

program. Throughout the audit, it was apparent that the applicant's document control system was sufficiently complete and organized to retrieve the documents necessary to support these discussions.

The programs mentioned above enhance the PVORT's confidence that the applicant's overall program can ensure all pumps and valves important to safety will operate as required for the life of the plant.

l 30

The PV0RT's evaluation cf the applicant's overall program was not cntirely absent of qualification issues, however. The PVORT did identify five generic issues that the applicant must resolve prior to fuel load.

All issues were discussed with the applicant at the exit meeting and are

~

presented below.

3 The overall equipment qualification and operability program provides the mechanism for sharing information between various administrative, design,. operations, maintenance, and quality control programs.

Seabrook

~

personnel described the maintenance program as a comprehensive network involving procurement, test control, and design control.

Implementation of the maintenance program is accomplished by a variety of subprograms such as preventive and corrective maintenance, tag management, station staff work requests, and utilization of the incomplete items list (IIL).

However, at i

the time of the audit, the PVORT was unable to review a complete set of maintenance procedures for any of the equipment selected.

Consequently, the first generic issue requires the app 1'icant to provide examples of the complete maintenance procedures for several equipment, at least one of which was reviewed during the PVORT audit. The procedures should clearly describe how limited life compgdertts will be addressed in order to ensure thattheequipmentwillremainlualifiedforthelifeoftheplant.

2 The second generic issue, which the applicant must confirm, is that a11~ active safety-related BOP valves smaller than two inches are included in the overall equipment qualification program in place at the conclusion of the site audit.

Regarding the third generic issue, it was apparent at the conclusion of the audit, that the Seabrook active valve list was not totally up-to-date.

In order to illustrate this concern, a brief discussion of the pre-audit preparation is presented here. Two months prior to the audit, the PVORT reviewed the Seabrook~FSAR, as well as the Master List of safety-related equipment.

Numerous discrepancies were identified, most of which could be attributed to the normal delay in updating FSAR amendments.

31

After some clarification by Seabrook, the PVORT compiled a list of 10 components ~to be audited.

However, the status of thr'ee of these items was changed by Seabrook less than a month before the site audit.

First, 3,

the 3-inch Lonergan relief valve, CC-V-120, was declared to be no longer safety-related. Second, the 18-inch Velan check valve, FW-V-38, had been replaced with a Rockwell control closure check valve in order to reduce the potential for water hammer. Third, the 3-inch air operated globe valve,.

LCV-460, was declared to be no longer safety-related by Seabrook. The PV0RT did not ' review this component or any replacement, but instead investigated the reasons why the status of equipment was not yet complete.

Seabrook explained that an independent consulting firm had been contracted.

to perform a consistency review of the plant equipment.

This review compared the original classification and qualification of all components with the latest industry practices, and recomended any changes such as those mentioned above. All of the mechanical equipment have undergone this consistency review, while the electrical, equipment consistency review will be completed.later this year. Very few changes have been identified by the consistency review to date. However, the PVORT believes that the entire review should have been completed before the site audit was held, since three of the 10 PVORT components were directly affected. Therefore, the

(

third generic issue is that the applicant must provide a complete list of active safety-related valves in the FSAR prior to fuel load.

Regarding the fourth and fifth generic issues, the staff requires that all equipment important to safety be properly qualified prior to fuel load.

However, the PVORT audit was conducted months in advance of the oxpected fuel load date before the applicant had been able to qualify, l -

test, and install all of his equipment. The applicant did provide evidence that the documentation and installation was complete for approximately 85 percent of the Seabrook equipment at the time of the audit.

The remaining 15 percent is scheduled to be completed prior to fuel load.

Similarly, some preoperational tests remained to be completed.

The hot l

functional tests were still in progress at the conclusion of the audit and i

were scheduled for completion in late November.

32

+.,...

Therefore, the fourth generic issue, for which the applicant must provide written confirmation, is that all pre-service tests required to be completed before fuel load'have been performed.

Finally, the fifth generic issue is that all pumps and valves important to safety are properly qualified prior to fuel load.

Complete qualification includes, but is not limited to, confirmation that (a) the j

associated documentation is complete and readily accessible, (b) the

. equipment is properly installed, and (c) the appropriate administrative procedures have been performed as required. The applicant has agreed at the conclusion of the audit to update the FSAR prior to fuel load, which will resolve the remaining SER issues and site audit concerns.

u Section 3 summarizes the five generic issues mentioned above as well as the five specific concerns mentioned in Sections 2.2.1, 2.2.2, and 2.2.6.

e 4

f.

O 33

3.

CONCLUSION The Equipment Qualification personnel for Seabrook are, dealing with the equipment qualification issue in a positive manner. The PVORT has rcached this conclusion because the applicant has:

(a) provided adequate documentation to demonstrate qualification of a representative sample of pumps ar.d valves important to safety, (b) established administrative programs to determine, monitor, and maintain equipment operability for the life of the plant, (c) demonstrated an adequate central file system by the timely retrieval of information requested by the staff, (d) demonstrated that he corresponds closely with the NSSS vendor, architect-engineer, and equipment suppliers concerning details of construction, design, maintenance, utility policy, and plant operation, and (e) demonstrated cverall accountability by committing the appropriate personnel to implement these policies and programs.

/

Based on the results of the on-site audit, the PVORT concludes that an

~

appropriate Pump and Valve Operability Assurance Program has been defined F

and is being implemented at Seabrook. The continued implementation of this program should provide adequate assurance that all pumps and valves 1mportant to safety will perform their safety-related functions as required for the life of the plant.

Table 3 presents a summary of the audit results.

By the close of the on-site audit, all but five specific and five generic concerns have been resolved.

These concerns were identified to the applicant and he committed to resolve them prior to fuel load.

The following is a list of all unresolved pump and valve operability concerns and the applicant's commitments:

Eautoment Specific Confirmatory Issues:

l 1.

The applicant shall confirm that the auxiliary feedwater pump (FW-P-37A) turbine operability is addressed regarding the potential of having moisture in the driving steam.

34

==--~--.----+.,--m,,

,,,,_,,,___.r__

o y

TABLE 3

SUMMARY

of PVORT AUDIT Plant I.D.

Number Description Safety function

_ Findings Resolutions Status Remarks Note,b,c Noted FW-P-37A Turbine driven To provide feedwater a

Open' Turbine operation needs to (80P) auxiliary feed-to the steam generator addressed when there is water ptop in the event normal moisture in the steam.

feedwater is not Turbine trip and throttle d

available.

valve operation after a trip needs to be addressed.

The turbine end pump seal was found to be cracked.

The reason for the failure needs to be investigated and resolved.

FW-V-331 Main feed water to To isolate the feed-NoteI Noted Open*

Operating time of this valve (80P)

, steam generrtor "B" water header in the is laq>ortant to safety isolation check event of loss of Timing requirements were not valve feedwater addressed.

CC-V-975 Primary component Tp isolate the

. Closed Specific concerns were (BOP) cooling water to radiation monitor resolved during the audit.

radiation monitor when full PCCW flow isolation val've is required by saf ety grade equipment.

FW-V-48 SG "C" feedwater closes on containment Closed Specific concerns were (80P) containment isolation signal.

resolved during the audit.

Isolation valve CC-V-122 Primary component Closes on isniation Closed Specific concerns were (BOP) cooling water signal.

resolved during the audit.

return isolation frne non-safety grade components SW-P-110A Cooling tower To provide cooling Note 9 Noted Opene Two 0-rings are used to (80P) pump A water flow when the control lateral support of cooling tower is pump column. The 0-rings used as the ultimate should be maintained for the heat sink.

Ilfe of the plant.

CS-P-28 Centrifugal To provide horated closed Specific concerns were (NSSS) charging pump 8 water and makeup as resolved during the audit.

~

well as high head safety injection.

i E

e I

.-y_

i TA8tE 3 (continued)

Plant I.D.

Number Description Safety Function Findings Resolutions -

Status Remarks RC-V-456A Pressurizer PORV Opens to prevent a closed Specific concerns were (NSSS) reactor trip due to resolved during the audit.

overpressure of pressurizer.

RH-V-14 Cold leg injection Closes for containment Closed Specific concerns were,

(hSSS)

RHR return line isolation and hot resolved during the audit.

1 solation valve leg recirculation.

ALL PUMPS AND VALVES Operate as required NoteheIedekel Noted Open*

None.

IMPORTANT TO SAFETY during the life of the plant under normal and accident conditions.

s a.

(SPECIFIC ISSUE) Turbine operation when moisture is mixed with the steam was not investigated. Turbine operation with moisture in the steam needs to be investigated and addressed.

b.

(SPECIFIC ISSUE) The turbine trip and throttle valve trstallation was not made in a wey that assured easy operation. Easy operation of the trip and throttle valve with a maximum differential pressure across the valve (for example, a turbine overspeed condition) rat not demonstrated. Easy operation of the trip and throttle valve needs to be investigated.

c.

(SPECIFICISSUE) The turbine end pump seal was found cracked. The cause of the cracked pump seal needs to be investigated and resolved.

d.

At the conclusion of the site audit the staff summarized the remaining open issues. The applicant was informed of the appropriate actions necessary to resolve the specific and generic confinsatory issues prior to fuel load.

e.

The qualification status will be " closed" upon resolution of the specific and generic issues, f.

(SPECIFIC ISSUE) This valve was changed from a swing check to a control check that has specific opening and closing times. The Operating times were not addressed in the startup, testing, or operating procedures. The applicant shall confirm that the operating times have been investigated and the timing requirements identified and met.

g.

(SPECIFIC. ISSUE) The maintenance program did not include procedures for replacing the 0-rings per manufacturer's recommendations.-

The maintenance program should include procedures for maintaining the qualification status of the 0-rings for the life of the plant.

h.

(GENERIC ISSUE) Maintenance procedures were in a draf t form and penerally not available for review. The applicant shall confirm that all final maintenance procedures are consistent with manufacturer s requirements. Appilcant shall describe how limited life l

components are identified. The appilcant shall provide examples of maintenarece procedures for review.

I 1.

(GENERIC ISSUE) BOP valves smaller than two inches were not included in the FSAR active valve ilst. The applicant shall confirm that the FSAR BOP list addresses valves less than tuo inches.

e 3'

s

.5

U

.J.

TABLE 3 (continued) e=

J.

(GE ERIC ISSUE) The active valve lists in the FSAR wer2 not comp 1;ts. Thz (pplicant shall confirm that all active pumps and er.lves y

are included in the FSAR active component lists.

k.

(GENERIC ISSUE) All pre-service tests have not been completed. The applicant shall confirm that all pre-service tests that are required before fuel load have been completed.

1.

(GENERIC ISSUE) The applicant has not completed the qualification of all pumps and valtM important to safety. The applicant shall confirm that all pumps and valves important to safety are quellffed prior to fuel load.

d 1

t; O

ep 9

h 8

i:

e L

I e

S l

9 e

s

o 2.

Prior to the audit, the turbine end of the aux 111ary feedwater i-pump-(FW-P-37A) was found to have a cracked seal. The cause of the seal failure had not been determined nor had steps been taken to prevent a recurrence. The applicant shall confirm that this

~'

failure is investigated and resolved.

3.

Operation of the auxiliary feedwater pump (FW-P-37A) turbine trip and throttle valve was not investigated when a maximum differential pressure'e'xisted across the valve such as a turbine overspeed trip' condition. The applicant shall confirm that the trip and throttle valve can be operated easily during an emergency condition.

i 4.

Check valve (FW-V-031) was changed from a swing check to a' control check that has specific opening and closing times. The operating times for this control check valve were not addressed in the startup, testing or operating procedures.

The applicant shall confirm that the operating times have been investigated and the timing requirements identified and met.

5.

The maintenance procedures for the cooling tower pump (1-SW-P-110A) were still in draft form at the time of the audit.

The procedures did not address the two 0-rings located at the-lateral supports for the pump column.

The applicant shall confirm that the final maintenance procedures specify the special handling and replacement of the 0-rings.

Generic Confirmatory Issues:

1.

At the time of the audit, the maintenance procedtres were available for review in draft form only. The applicant shall confirm that the final maintenance procedures will be consistent with the component manufacturer's recommendations.

The applicant shall describe how limited life components are identified, and how the equipment will be maintained in an operable and qualified i

1 38

.......m.'

.. c.

state for the life of the plant. The applicant shall provide several examples (at least 1 pump and 1' valve) of the final maintenance procedures for review.

p.

2.

The applicant shall provide written confirmation in the FSAR that all active BOP valves are covered by the Seabrook pump and valve I

r operability assurance program.

In particular, the applicant shall confirm that BOP valves smaller than two inches have been included.

{

l 3.

At the conclusion of the PVORT audit, it was apparent that a complete list of active valves had not been provided in the 7SAR. The applicant shall confirm that all active valves are correctly identified in the FSAR.

4.

At the time of the audit, most construction tests had already been completed.

However, the hot' functional tests were still in progress. The applicant shall confirm that all pre-service tests that are required before fuel load have been completed.

5.

At the time of the audit, approximately 10 to 15 percent of all pumps and valves important to safety had not been qualified. The applicant shall confirm that all pumps and valves important to safety are properly qualified and installed.

In addition, the applicant shall provide written confirmation that the original loads used in tests or analyses to qualify pumps and valves important to safety are not exceeded by any new loads, such as those imposed by a LOCA (hydrodynamic loads) or as-built

~

5 conditions.

E l

4.

NSSS REFERENCES CENTRIFUGAL CHARGING PUMP. CS-P-28 1

1.

Westinghouse, Lubricant and Bearino Report for Medium. Larce and Chemoump Motors, NS-I&CSL-82146, June 22, 1983.

Crane Packing Company, Seal Performance Test [no for Nuclear Power 2.

Plant Safety Injection Systems, Bulletin 3472.

3.

Westinghouse, Auxiliary Pump Motors, 677474 Revision 0, March 3, 1972.

4.

Westinghouse, E-Spec Class 2 Pumos, 678815 Revision 2 September 6, 1973.

5.

Westinghouse, Class 2 Pumos for NAH, 952470 Revision 2, October 6, 1975.

6.

Westinghouse, Auxiliary Pump Motors for NAH, 952347 Revision 2, April 18,1975.

7.

Westinghouse, Thermal Transient Test, MED-PVE 3840, October 15, 1985.

8.

Westinghouse, 100 Hour Endurance Test, MED-PVE-3840, October 15, 1985.

9.

Pacific Pumps, Nuclear Service Pumo Design Analysis Report, K-435 Revision 2, October 25, 1978.

10.

Westinghouse Large Motor Division, Seismic Analysis of Centrifugal Charoina Pump Motors, Seismic S.O. 76F60176, March 23, 1977. -

40

. e.

a

11. Pacific Pumps, Natural Frecuency Test Report, Pump Serial 51669, July 26, 1977.

12. Westinghouse, Ecuipment Qualification Test Report W LMO Motor Insulation, WCAP 8687 Supplement 2 A02A, July 1981.

s

13. Westinghouse, Eauipment Oualification Data Package-Large Pumo Motors (Outside Containment), WCAP 8587 EQDP-AE-2, March 1982.

14. Westinghouse MM&G, Gear Seismic Analysis Certificate of Applicability, 76261075, November 15, 1977.

15. Westinghouse, Class ? Pumos - Based on ASME BP&V Code Section III Rules for Construction of Nuclear Power Plant Components, 678815, Revision 2, November 12, 1971.

16.

Westinghouse, Class 2 Centrifundi and Positive Displacement Pumos (Seabrook 1 & 2), 952470 Attachments to E-Spec 678815 Rev. 2, February 7, 1974.,

1 POWER OPERATED RELIEF VALVE RC-V-456A l

17. Westinghouse, Power Operated Relief Valve E-Spec, 955245 Revis~ ion 0, September 17, 1980.

18. Westinghouse, Ecutoment Oualification Data Packace for PORV Solenoid Pilot Valve and PID, WCAP-8587 and EQDP-HE-9, January 1985.

s

19. Westinghouse, Ecuipment Oualification Test Report for PORV Solenoid Pilot Valve and PID, WCAP-8687 Supplement 2-H09A, January 1985.

20.

Garrett, Design Report Stress Analysis of the PORV, 41-28708, December 9, 1981.

41

21.

Garrett, Seismic-Oualification Operability Tests for PORV, 41-4402A, May 21, 1984.

?

-^

22.

Garrett, Class 1 Desian Report for PORV, 41-3256, December 4,1981.

COLD LEG INJECTION /RHR RETURN LINE ISOLATION VALVE. RH-V-14 x

23. Westinghouse, Project Specification, 952272 Revision 4, February 3, 1984.

24. Westingh.3use, General Specification, G-678852 Rev.ision 2 March 14, 1977.

25. Westinghouse, Stress Report for Westinahouse class 1 6" and Larcer Gate valves, EM-5405 Revision 1 May 27, 1980.

26. Westinghouse, Operability Test Report for Westinahouse Nuclear Gate Valves, EM-4995, January 28, 1977.

27. Westinghouse /Limitorque, Ecuipment Qualification Test Report.

Limitorcue Electric Motor Operator (Environmental and Seismic Te_stino-Outside Cantainment. Non-HELB Environments), WCAP 8687 Supplement 2-H04A Revision 1, March 1983.

28. Westinghouse /Namco, NAMCO Externally V,ounted-Limit Switches-Environmental and Seismic Testina, WCAP 8687 Supplement 2 H03A/H06A, March 1983.

e 42

5.

BOP REFERENCES FEEDWATER SYSTEM CONTROL /CED CHECK VALVE. FW-V-331

21. United Engineers and Constructors, Feedwater System P&I Diaoram, Report No. 9763-F-202079, October 30, 1984.

Y

22. United Engineers and Constructors, MS & FW Pipe Chase-Feedwater System LN4607 Fabrication and Support Isometric, Report No. 9763-F-202397, January 9, 1985.

23. Rockwell International, Control Closure Check Valve 18 x 16 18-FIG 2092 (WCC) 8J0TY, Report No. FP97789-04 DWG No. D84-30371-07, May 29,.1985.

24.

Rockwell International, Desion Report for Size 18 x 16 x 18 Fioure 2092 Controlled Closure Ch'eck Valve, Report No. FP97794-01 and Report No. RAL-4124, March 8, 1985.

25.

RockwellInternatIonal,PerformanceCalculationsfor

~

Size 18 x 16 x 18 Fio. 2092 Controlled Closure Check Valve, Report No. FP97792-01 and Report No. RAL-1071, March 8, 1985.

26.

United Engineers and Constructors, Specification for controlled Closure Check Valves, Report No. 9763-006-248-85, January 8, 1985.

27.

United Engineers and Constructors, Seismic Requirements for PSNH, Report No. 9763-S0-248-85, January 10, 1984.

s

28. Kirk, M. J. & Gradle, R.

J., A Model for Check Valve /Feedwater I

System Water Hammer Analysis, ASME 80-C2/PVP-27, March 17, 1980.

29. Rockwell Internationa?, Hydrostatic and Seat Closure Test Procedure for Size 18 x 16 x 18 Fiqure 2092 (WCC) 8J0TY 9tntro11ed Closure Check Valves, Method Speculation 7098, May 16, 1985.

l

)

43

30. United Engineers and Constructors, Statement on Valve Failure Mode Analysis (by J. J. Parisand), November 7, 1985.

31. United Engineers and Constructors, Temoerature/ Pressure Conditior.s for Valve V331, Design Data Sheet 9763.006.

TUR8INE DRIVE AUXILIARY FEE 0 WATER PUMP. FW-P-37A

32. Polytechnic Desian Co.. Inc./Incersoll-Rand Company Structural Intecrity and Operability Analysis of 4 x 9NH-10 Turbine Driven Feedwater Pumo, FP 22810-03 Report No. EAS-TR-8001 Review, November 19, 1984.

33.

Structural Intearity and Goerability Analysis of 4 x 9NH-10 Turbine Driven Pumo, F.P. 23812-01 Document No. EAS-TR-8001 Revision 0, November 5, 1984.

34.

United Engineers and Constructors, Specification for Emeraency Feedwater Pumos, 9763-006-238-10 Revision 10. September 15,1977.

35. United Engineers and Constructors, Seismic Specification for

!L PSNH, 9763-SD-238-10 Revision 3, September 15, 1977.

36.

Terry Corporation, GS-2N Oualification Report for Ingersoll-Rand-Cameron F-41062/F-41063 and Wyle Test Report 58038, F.P. 22719-02 Report No. TM-105 Revision 0, August 7, 1980.

37.. Woodward Governor Company, Pressure Actuated Shutdown Assembiv, Bulletin 36651A, 1965.

1 e

8 44 i

PRIMARY COMPONENT COOLING WATER RADIATION MONITOR ISOLATION VALVE. CC-V-975 38.

G. H. Bettes Corporation, Seismic Analysis Robotarm Actuator Model CB420-SR60, F.P. 97623-01 Report No. 1062-67 Revision A, January 26, 1984.

e 39.

Xomox Corporation, Valve Operability Procedure, F.P. 97580-02 Procedure Number 399-16000-00 Revision 0, January 23, 1984.

40. Xomox Corporation, Desian and Seismic Report for 1 inch Fiqure 166. CL 150 Plua Valve 66. with Bettis CB420 SR60 Actuator, F.P. 97639-01 Procedure Number 384-18000 Revision 0, January 30, 1984.

41. TuflineDivisionofXomoxCorporption,SeismicAnalysis.

F.P. 91420-02 Report No. 1-1366-78 Revision A, April 9, 1979.

42. United Engineers and Constructors Inc., Specification for Pluo Valves, 9763-006-248-29 Revision 3, November 15, 1977.

43.

G. H. Bettis Corporation, Nuclear Qualification Test Report, F.P. 61940-4 37274 Revision 3, May 10, 1983.

44.

Automatic Switch Company, Qualification Test of Solenoid Valves, F.P. 97647-01 AQS-21678/TR Revision A, February 3, 1984.

45.

Namco Controls, Qualification of EA 740 Series Limit Switches for Use in Nuclear Power Plants, F.P. 93639-01 QTR-111 Revision 0, December li, 1982.

46. Automatic Switch Company, Report on Qualification nf Automatic Switch Company (ASCO) Cataloa NP-1 Solenoid Valve, F.P. 61958-03 AQR-67368 Revision 0, October 10, 1983.

4 45

47. United Engineers and Constructors, Specification for Actuators for Valves and Damoers, 9763-006-248-13 Revision 6. May 23, 1980.

FEEDWATER ISOLATION VALVE. FW-V-48 48.

Borg-Warner Corporation, Seismic Analysis of 18 inch. 900 lb Carbon Steel Pneumati'c-Hydraulic Operator Gate Valve,

.F.P. 2245-07 Document NSR-73890 Revision D. July 22, 1982.

49. Borg Warner Corporation, Operability Test Procedure for 18 inch Feedwater Station Valve, F.P. 21233-03 Document OTP-73890

~ Revision A. December 29, 1980.

50.

United Engineers and Constructors, Specification for Feedwater Isolation Valves, 9763-006-248-36 Revision 2, August 28, 1980.

51.

United Engineers and Constructors, Seismic Reautrements for i

P.S.N.H., 9763-SP-248-36 Revision 0, November 4, 1984.

i

52. 'Borg-Warner Corporation. IEEE Oualifications Report for a

~

Hydraulic Operator BWFC P/N 37951, F.P. 23756-02 Report No. 2064 Revision A, August 12, 1985.

l PRIMARY COMPONENT COOLING AUTOMATIC CONTAINMENT ISOLATION VALVE. CC-V-122 53.

Seabrook Station, Preoperational Test Procedure, 1-PT(I)-37.2, October 16, 1985.

54. Yankee Atomic Electric Company, (YAEC), Nonconformance/LWA Report-Seabrook Station, Report No. NCR 82-284, November 2, 1984.

55.

Seabrook Station, Phase 1 System Test Packace, TPI-51-F01, August 22, 1985. -

46

.... _.. a :

2..._..,..

56.

UE&C, Specification for Butterfly Valves, Report 9763-006-248-45, August 16, 1976.

57.

UE&C, Seismic Recuirements, Report 9763-50-248-45, a

September 22, 1975.

58. Posi-Seal International,' Actuator Mathematical Model, Report F.P. 91742-2 Revision 0 11473, October 27, 1983.

59. NAMC0/ ACME, Limit Switch IEEE Oualification Report, Report-No. F.P. 91908-1 Revision 1 EA-470, May 22, 1980.

60.

Posi-Seal International, Hydrostatic Shell Testinc, Report No. F.P. 90350-1, Revision 1 TP900, April 2, 1976.

61.

Posi-Seal Intern 3tional, Cycle Testina, Report No. F.P. 90349-1 Revision 0 11473 TP-2, April 2, 1976.

62. Posi-Seal International, Leak Testino Bubble Formation Method, Report No. F.P. 90815-1 QCS-004, April 27,1977.

u 63.

Posi-Seal International, Leak Testino. Hydrostatic, Report No. F.P. 90816-1 Revision 0 QCS-005, April 27, 1977.

64.

Posi-Seal International, Hydrostatic Seat Testino, Report No. F.P. 90348-1 Revision 0 11473.TP-1, April 2, 1976.

65.,NORDAN, Seismic Vibration Test Report, Report No. F.P. 90614-01 Revision 0 2060-0001, December 28, 1976.

66. WYLE/POSI-S6AL, Seismic Vibration Test Report, Report No. F.P. 906i5-2 Revision 0 54598, October 23, 1985.

S 47 i

67. Posi-Seal International, Seismic Certificate of Compliance, Report No. F.P. 93535-2 Revision 0, February 23, 1983.

~

68. Posi-Sea 7 International, Nuclear Seismic and LOCA Analysis, Report No. F.P. 91157-8 Revision 8 11473 August 5, 1985.

l

69. ASCO/ISDMEDIX, Qualification Test for Solenoid Valve, Report No. 91904-2 Revision 0 AQS-21678-TR, January 4,1982.

70.

Post-Seal International, Cvele Testino of Valves, Report No. F.P. 9222-1-01 Revision 011473-TP-3 April 17,1982.

COOLING TOWER PUMP. SW-P-110A

71. United Engineers and Constructors, Specification for Service Water Coolino Tower Pumos, 9763-006-238-20 Revision 5 July 7, 1977.

72. United Engineers and Contractors, Seismic Requirements for P.S.N.H., 9763-SD-238-20 Revision 6, March 7, 1980.

73.

Mcdonald Engineering Analysis Company, Seismic Stress Analysis of ASME Section III Class 3 Pumos, F.P. 55828-03 Document ME-772, January 4, 1984.

74.

Mcdonald Engineering Analysis Company Inc., Seismic Stress An_a'ivsis of 800 HP Service Water Coolino Tower Pump Motor, F.P. 51589-04 Document ME-266, March 15, 1977.

.~

75.

General Electric Topical Report on GE Vertical Induction Motor for Class IE Nuclear Service Water Pumo and Coolino Tower Pumo, F.P. 51880-05, February 27, 1970.

b 48

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76.

Johnston Pumps Service Water Coolina Tower-Pumos, F.P. 50628-11 October 10, 1985.

77. Johnston Pumps, SWCTP-Data Sheets and curves, FP 52275 Issue 01, January 27,1977.

o

78. United Engineers and Constructors, letter-Possible 10CFR50.55(e)

Deficiency Service Water Pumos (Coolina Tower Pumos), S8U-94362, June 28, 1985.

79. Johnston Pump, SWCTP Operatina and Maintenance Manual, F.P. 53040 Issue 08, October 27, 1983.

80.

Generai Electric, SWP and SWTCP Motor Instruction Manual, F.P. 51911 Issue 01, May 12, 1976.

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BIBUOGRAPHIC CATA SHEET EGG-NTA,7156

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' -srma rna a u...us Audit of the Pump and Valve Operability Assurance Program for the Seabrook Station i

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C. Kido, H. M. Stromberg, T. C. Chung

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& Mtoa.CrirA.unsons usest mons m EG8G Idaho, Inc.

Idaho Falls, ID 83415 A6415, Project IV i

' at WOpe.OneseG C# Gases 2Ariese seams. ano asasusee a00m.M p.unser4 Casse sIarYP.0Pa.PORr Division of PWR Licensing - A Technical Evaluation Report Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

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Washington, DC 20555 11 SAPPL.33.Mr.4Y PgCL 3

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The Seabrook Station Unit I was audited November 5 to 8, 1985 to determine the adequac'y of.their, Pump and Valve Operability Assurance Program.

Ten concerns (five specific and),1ve generic), which could not be resolved by the close of the audit, were identified to the applicant; he committed I

to address these concerns prior to Tuel load.

The results of this audit indicate that the applicant has established and is implementing a p,rogram that will track all pumps and valves important to safety from manufacture and in-shop testing through qualification, installation, testing, maintenance,'

and surveillance for the purpose of assuring continued operability of these components over the life of the plant.

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Prepared for the U.S. NilCLEAR REGULATORY COMMISSION Work performed under h

DOE Contract

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No. DE-AC07 76fD01570 i.

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J. N. Singh T. L. Bridges B. L. Harris l

i Published February 1986 i

EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 t

Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6415

ABSTRACT EG&G Idaho is assisting 'the Nuclear Regulatory Commission in cvaluating Public Service Company of New Hampshire's program for the dynamic qualification of safety related electrical and mechanical equipment fcr the Seabrook Station Unit 1.

Applicants are required to use test or analysis or a combination of both to qualify equipment, such that its ssfety function will be ensured during and after the dynamic event, and provide documentation.

The review, when completed, will indicate whether an appropriate qualification program has been defined and implemented for scismic Category I mechanical and electrical equipment which will provide reasonable assurance that such equipment will function properly during and after the excitation due to vibratory forces of the dynamic event.

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11

SUMMARY

A seismic qualification review team (SQRT) consisting of engineers from the Equipment Qualification Branch of the Nuclear Regulatory Commission and Idaho National Engineering Laboratory made a site visit to the Seabrook Station, Unit 1 plant of Public Service Company of New Hampshire located near Seabrook, New Hampshire.

They observed the field

. installation and reviewed the qualification reports for twenty-one selected pieces of seismic Category I electrical and mechanical equipment and their supporting structures. ' Three generic, one equipment specific, and two confirmatory concerns were identified for which additional information is needed in order for the SQRT to complete the review. These are referred to as open items The review indicated that the equipment was ade'quately qualified for the dyn3mic environment pending resolution of the open items.

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CONTENTS' ABSTRACT..............................................................

11

SUMMARY

iii 1.

INTRODUCTION.....................................................

1 2.

NUCLEAR STEAM SUPPLY SYSTEM (NSSS) EQUIPMENT.....................

2 2.1 3-inch Globe Valve--Air Operated--(NSSS-2).................

2 2.2 Safety Injection System Accumulator Tank--(NSSS-3).........

3 2.3 Electric Hydrogen Recombiner Power Supplies--(NSSS-4)......

4 2.4 Reactor Water Make-up Valve--(NSSS-5)......................

5 2.5 8-inch Motor Operated Gate Valve--(NSSS-6).................

8 2.6 Reactor Trip Switchgear--(NSSS-7)..........................

9 2.7 Reactor Vessel Level Instrumentation System 8086 Cabinet--(NSSS-8).....................................

10 2.8 Nuclear Instrumentation System Cabinet--(NSSS-11)..........

11 2.9 Safeguards Test Cabinet--(NSSS-12).........................

13 2.10 Instrument Bus Power Supply:

Static Inverter--(NSSS-13)...

14 3.

BALANCE OF P LANT (BOP) EQUIPMENT.................................

16 3.1 36-inch Butterfly Valve--(80P-1)...........................

16 3.2 Control Switch--(BOP-3)....................................

17 3.3 Computing Device--(BOP-4)..................................

19 3.4 Emergency Feedwater Pump and Turbine--(B0P-5)..............

20 3.5 4-inch Motor Operated Globe Valve--(80P-6).................

22 3.6 Neutron Flux Signal ~ P'roce ssor--(BOP-7).....................

23 3.7 Vibration Monitoring Control Panel--(80P-11)...............

25

~

3.8 18-inch Feedwater Isol at ion Val ve--(80P-14)................

26 3.9 6-inch Motor Operated Gate Valve--(80P-15).................

27 iv

1 1

3.10 Diesel Generator Relay Control Panel--(BOP-16).............

28 3.11 Pressure Switch--(80P-17)..................................

29 4.

FINDINGS AND CONCLUSION..........................................

~30 4.1 Generic Issues.............................................

30 4.2 Equi pment Speci fi c I s s ue s..................................

31 4.3 Confirmatory Issues........................................

31 l

4.4 Conclusion.................................................

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TABLE L

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Li s t o f a tte n de e s................................................

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INTRODUCTION The Equipment Qualification Branch (EQB) of the Nuclear Regulatory Commission (NRC) has the lead responsibility in reviewing and evaluating the dynamic qualification of safety related mechanical and electrical equipment. This equipment may be subjected to vibration from earthquakes and/or hydrodynamic forces. Applicants are required to use test or analysis-or a combination of both to qualif> equipment essential to plant i

safety, such that its function will be ensured during and after the dynamic' event.' These pieces of equipment and how they meet the required criteria are described by the applicant in a Final Safety Analysis Report (FSAR).

On completion of the FSAR review, evaluation and approval, the applicant receives an Operating License (OL) for commercial plant operation.

A Seismic Qualification Review Team (SQRT) consisting of engineers from the EQB of NRC and Idaho National Engineering Laboratory (INEL), made a site visit to the Seabrook 1 nuclear power plant of Public Service Company of New Hampshire near Seabrook, New Hampshire, from November 5 through November 8, 1985.

The purpose of the visit was to observe the field installation, review the equipment qualification methods, procedures (including modeling technique and adequacy), and documented results for a 1

list of selected seismic Category I mechanical and electrical equipment and their supporting structures.

This report, containing the review findings,-

indicates which of the items are qualified and require no additional documentation.

It also identifies some equipment and certain general concerns for which additional information is needed in order for the SQRT to complete the review.

These are referred to as open items. The applicant is to further investigate and provide additional documentation to resolve these issues.

Table I contains a list of personnel who attended the site visit.

Subsequent sections of this report give a brief overview and identify the concerns, followed by the findings, for the selected seismic Category I equipment.

1

2.

NUCLEAR STEAM SUPPLY SYSTEM (NSSS) EQUIPMENT 4

2.1 3-inch Globe Valve-Air Operated (NSSS-2) i The 3-in. air operated globe valve, supplied by Copes-Vulcan (model 3I88RG) was used for reactor coolant system pressure boundary

~

isolation. This function has now been moved to a Westinghouse EMD supplied I

3 in. motor operated gate valve.

The decision to move the safety function was made in a meeting with personnel from Westinghouse, UE and C, and Yankee Atomic. The paperwork from the meeting was not finalized at the -

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time of the audit.

Therefore, the qualification documentation for th'e l

Westingh'ouse valve was not available at the audit. Therefore the d:cumentation for the Copes-Vulcan valve were reviewed at the audit.

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The qualification of the valve was done using a static analysis to l

d:monstrate structural integrity of the valve since no natural frequencies balow 33 Hz were identified from hand calculations.

(See Copes-Vulcan report 10.3.030, Rev. 5, Seismic Analysis Air Operated Control Valve, i

l January 27,1984.)

l Stresses in the valve body and bolts were shown to be below ASME j

Szction III allowable values in Copes-Vulcan Design Report 10.2.117, Rev. 2, 1 Inch Class 1500, 2 Inch Class 1500, 2 Inch Class 2500, 3 Inch ~

i

. Class 1500 Air Operated Control Valves, December 28, 1979.

I Operability of the valve was demonstrated using a static deflection test in Copes Vulcan Procedure No. 4.4.496, Rev. 2, Conducting Seismic Tests on Diaphraam Actuated Valves, October 5,1981.

The deflection used simulated simultaneous accelerations of 4 g vertical and 5.66 g

~

hsrizontal. The valve leakage rate in the deflected position of 1.8 cc/hr compared to acceptable rate of 9 cc/hr. The opening and closing times were i

2.4 sec and 1.0 sec, respectively, compared to a required time of 10 sec for both opening and closing.

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9 Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the 3-in. air operated globe valve is adequately qualified for the prescribed i

loads.

2.2 Safety Injection System Accumulator Tank (NSSS-3)

The safety injection system (SIS) accumulator tanks (MPL SI-TK-9A) were supplied by Southwest Fabrication and Wolding Co.

These four tanks were purchased to specification E-spec. 679065 Rev. 3.

Each tank provides 1350 cubic feet of storage for emergency core cooling water.

The tanks are vertical with base anchorage consisting of 28 2 in, diameter embedded

' bolts. They.are located at elevation -26 ft of the containment building.

These tanks were qualified by analysis performed by Basic Technology, Inc., documented by reports BTI-PJ 75015 Rev. 2, dated September 29, 1975, and BTI-76057 dated June 4, 1976.

The stress analysis of these tanks was performed in accordance with the requirements of ASME Code Section III for Class 2 vessels.

The tanks were evaluated for dead weight, pressure, nozzle, and seismic inertia loading.

The natural frequencies of the tanks were calculatad to be 22 Hz horizontal and 71 Hz vertical.

The peak horizontal design spectra for SSE is.85 g at 2.5 Hz for a damping value of 4%.

The corresponding vertical value is.81 g at 3.5 Hz.

The tanks were analyzed statically using a horizontal acceleration of 1.5 g and a vertical acceleration of 1.G g thus providing a very conservative analysis for seismic loading.

The areas of critical stress was determined to be the anchor bolts and support skirt.

For the faulted load combination (SSE plus normal loads) the calculated stresses were 7,674 psi for the anchor bolts and 4,868 psi for the support skirt compared to allowables at 26,600 psi and 14,500 psi respectively.

It was noted that the tank anchor bolts were 2 inch diameter rather than 2 1/4 diameter listed on the seismic and dynamic qualification summary long form.

This discrepancy was also noted by Westinghouse personnel during a walk down inspection prior to the NRC SQRT visit.

United Engineers and Constructors, Inc. personnel have perfor:aed an independent analysis of the tank and tank anchor bolts using the actual bolt size and determined that anchor bolt strength is adequate.

3

9 Even though the bolts are smaller in size than originally specified by the tank vendor, the bolt material (ASTM A193, Grade B7) yield strength is i

higher than specified, therefore, the bolts used had a net strength in oxcess to the specified bolts.

Based on our inspection of the field installation, review of the qualification documents, and the applicant's responses to questions, the SIS accumulators are adequately qualified for the prescribed loads.

2.3 Electric Hydrogen Recombiner Power Supplies (NSSS-4)

The electric hydrogen recombiner (EHR) system is designed to fulfill the requirement of regulatory guide 1.7: Control of Combustible Gas Concentrations in Containment Following a loss of Coolant Accident.

This EHR power supply (tag no. 1-CGC-CP-246A, B) was supplied by Westinghouse par purchase spec. no. 679042.

It was manufactured by Halmar Electronics Inc. with serial number 8110453 according to Westinghouse's drawing I

no. 9556078.

There are a total of two units in the safety system. The unit is located in the control building at the 21 feet 6 inches ele *vation.

The area has mild environment.

Seismic loads are considered in the qualification.

j This is a vertically floor mounted item.

The mounting consists of oight 5/8 inch bolts with a three inch long spacer provided for each of the four corner. bolts.

It was qualified by tests performed by Westinghouse.

The results are in the reports:

Electric Hydrogen Recombiner for Water Reactor Containments, no. WCAP-7709L, revision 0, dated July 1971 and l

Electric Hydrogen Recombiner LWR Containments Supplements Test Number 2, no. WCAP-7709L SUPP. 7, revision 0, dated August 1977.

The laboratory mounting consisted of eight 1/2 inch-13UNC4 socket head cap screws. A l~

three inch long spacer was provided for each of the four corner bolts.

The difference between the field and laboratory mounting is reconciled by the fact that the field mounting has equal or greater strength than the laboratory mounting.

i l

4

There were a number of series of tests performed.

The first series was resonance search.

This was a pseudotriaxial sine sweep with a 0.2 g input. The range of sweep was from 1 to 42 Hz at one octave per minute i

rate.

Resonances detected were 3.5 and 11.'0 (front /back), 16 (side / side) and 33 Hz in the vertical direction.

There were two other kinds of qualification tests performed. A series of five pseudotriaxial with random inputs tests were done.

These were done with generic OBE level spectra.

Another series of eleven single frequency sine beat tests with inputs of five beats having ten cycles per beat at several frequencies were performed. The frequencies in hertz were 1.25, 1.75, 2.5, 3.5, 5.7, 9.5, i

13, 18, 24.5 and 33.5 plus resonances. These were performed in four orientations; each 90 degrees apart.

It was done with generic SSE level requirements. A 2.5 percent of critical damping was used in each s

-case of the TRS generated. TRS in each case adequately envelope the RRS

.for the location.

Structural integrity and operability were monitored in each case.

Only one anomaly, a loose wiring harness, was detected during the qualification test. A strap was added and four subsequent tests were done without further loosening.

This modification has been provided to the shipped equipment.

Another modification (FDR No. NAHM-10055,10/27/82) provided alternative terminal blocks to allow for minimum cable bend radius with no significant seismic consequence.

The qualified life of the equipment is 40 years based on adequate and proper maintenance which is in the process of being addressed by the utility.

Based on observation of the field installation, review of the qualification documentations and responses of the applicant to our inquiries the electric hydrogen recombiner power supply is adequately qualified for Seabrook Unit 1 location.

2.4 Reactor Water Make-up Valve (NSSS-5)

The reactor make-up water valve (ID no. RMWV-30; model no. N-226-B-ACC-SP) was manufactured and supplied by Walworth according to purchase spec no. 248-41.

It was located in Mech. Pen. Area at the -3 ft 5

. o 10.5 in. elevation. This 3 in. valve was butt welded to the pipe in the reactor make-up water system.

Its function is to provide containment isolation.. Seismic loads are considered in its qualification.

Its

~

qualificat1on is based'on a combination of test and analysis.

The details Gf the tests are contained in the report:

Report of Test for Nozzle Load--Seismic Testing of One (1) 3 Inch 150 lb Gate Valve With Pneumatic Operator, no. 17062-82N-2, rev. O, dated February 28, 1983.

Acton Environmental Testing Corp. wrote this report and United Engineers & Constructors reviewed it.

The analysis portion of the qualification is documented in the Walworth's reports:

Seismic Analysis and Natural Frequency Determination Calculations, no. ASF-13, rev. O, dated February 8,1983 and Design Stress Report for 3 Inch-150 lb Gate Valve, no. ADSR-16, rev. O, dated June 21, 1983.

The lowest-frequencies from the analysis were calculated to be i

s/s = 188.94 Hz, f/b = 54.34 Hz and v > 33 Hz.

However, the resonance search test indicated the following frequencies z c osed) f0.5

, f/b = 12.2 Hz and v > 33 Hz s/s =

On inquiry about the very significant Idifference between the test and analytically calculated frequencies the applicant submitted that the qualification of the valve was mainly based on testing.

The analytical model was mainly used for stress calculation where significant conservatism as well as margins exist.

This is a satisfactory explanation.

The stress results are as follows.

S 6

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Total calculated A1.lowable stresses stresses Source of Identification Location Loads (psi)

(psi) allowable ~s Bonnet Neck

Seismic, 12,724 25,500 ASME III stem thrust Bolting Bonnet / yoke Seismic, 12,749 21,200 ASME III stem thrust Bonnet Flange

Seismic, 21,022 25,500 ASME III stem thrust press.

Body Crotch

Nozzle, 14,757 25,500 ASME III press.

These stresses were calculated with 3 g applied at the C.G. in each of the three directions simultaneously.

The required g-levels were:

s/s f/b v

.(.9.L

.UO _

_(.9.1 s

0.4G 0.99 OBE 0.46 SSE 0.92 0.92 1.98 The tests performed on the valve as'sembly were a resonance search and a qualification test.

Mounting consisted of welding the valve pipe stub to a pipe end and the pipe flange, in turn, being welded to a test fixture.

The fixture was then welded to the vibration table.

Results of the resonance have already been noted earlier.

Qualification was performed with pseudobiaxial sine beat inputs. The specimen was mounted at an angle of 45 degrees.

Sine beat tests were performed at the frequencies of 1.0, 1.25, 1.6, 2.0, 2.5, 3.2, 4.0, 5.0, 6.3, 8.0, 10.0, 12.5, 16.0, 20.0, 25.0 and 33.0 Hz plus the resonances.

Operability and leakage were verified using nitrogen as a pressurizing agent.

During the test, the bolts between operator and bonnet were tightened as required.

This phenomenon indicated that the bolts would need periodic tightening.

On inquiry, the applicant indicated that the vendor installation and maintenance manual F.P. 97601-01 recommends bolt tightening sequence at regular intervals of approximately 12 months.

7

The reconciliation of the assumed g-loading with as built condition remains open.

This should be confirmed when completed.

Another issue of concern is the evaluation of life-span of nonmetallic parts.

~

Based upon the observation of the field installation, review of qualification documents, and applicant.'s response to questions, this item is adequately qualified pending confirmation of the generic issues tentioned above.

2.5 8-inch Motor Operated Gate Valve (NSSS-6)

The 8-in. motor operated gate valve (MPL No. RHR-8716A, B; model No. 8GM74FE.B) was supplied by Westinghouse. These two valves were purchased to specification G-678852.

The motor operators for these valves cre Limitorque model SMB-00.

The valves are line mounted using full penetration nozzle welds at elevation -21'8" of the auxiliary building.

The operstors are bolted to the valve bodies with eight 5/8 in. diameter bolts.

This active valve is in the RHR piping system.

Westinghouse qualified the valve by a combination of testing and analysis. The valve analysis, Westinghouse report No. 5405, rev 1, dated May 5, 1980, was performed in accordance with the requirements of ASME code section III for class 2 valves.

Loading in this analysis included pressure, nozzle loads, and operator seismic inertia loading.' The maximum calculated stresses were determined to be 7,770 psi compared to an allowable of 31,425 psi.

Demonstration of operability of t.he valve was accomplished by static deflection tests of 4 inch and 12 inch similar valves with operators.

Test loading consisted of 4.5 g seismic inertia loading of the operator, 2,500 psi pressure. and nozzle loads which resulted in attached piping bending stres-3/4 of material yield strength. This testing is documented by Wes;.1nghouse report No. 4995, dated January 28, 1977.

8

(

Qualification of the Limitorque motor operators was accomplished by single frequency, single axis testing documented by We'stinghouse report No. WCAP-8687 supplement 2-H04A dated September 1983.

The operator was tested to 7.75 g in each d.irection over the frequency range of 2 to 35 Hz.

Operability of the operator was demonstrated with no observed failures or anomalies.

5 Based on our inspection of the field installation, review of the qualification documents, and the applicant's responses to questions, the 8 in. motor operated gate valves are adequately qualified for the prescribed loads.

1 2.6 Reactor Trip Switchgear (NSSS-7)

The reactor trip switchgear, supplied by Westinghouse (model no.05-416) is located in the control building at elevation 21 ft 6 in.

The switchgear provides the reactor trip function.

The reactor trip switchgear circuit breakers were qualified by test as reported in Westinghouse report EQDP-ESE-20, Rev. 5, Reactor Trip Switchgear (OS-416 Circuit Breakers), March 1983.

Other parts of the switchgear were qualified by test in Westinghouse report EQDP-ESE-62A, Rev. O, Auto Shunt Trip panel and Shunt Trip Attachment for Reactor Trip Switchgear (05-416 Circuit Breakers), December 1984. A 0.2 g pseudotriaxial sine sweep from 1 to 50 Hz with a sweep rate of 1 octave / minute was performed and the following natural frequencies were identified:

S/S--between 5 to 6.8 Hz; F/b--between 10 and 12.5 Hz; V--33 Hz.

Multifrequency pseudotriaxial input from 1 to 33 Hz was used in.

the qualification testing of the switchgear.

Five OBE tests and 4 SSE tests were run with the'TRSs enveloping the Seabrook specific RRSs for all tests.

The damping was 5 percent.

Thermal aging was performed and a qualified life of 5 years was establishe'd.

The equipment functioned before, during, and after all tests.

9

The switchgear was bolted to the test table.

The bolt holes on the switchgear are slotted. The' actual plant mounting was done by welding the switchgear to a channel embedded in the concrete floor.

More sine sweep i

testing was performed and reported in Westinghouse report EQ&T-EQA-502, Siismic Confirmation of Welded Base Reactor Trip Switchgear and Static Inverter for Seabrook Units 1 and 2 App 1tcation, October 1983.

The following natural frequeacies were found:

S/S--10,Hz; f/b--15 Hz; V > 33 Hz. The welded base was shown to be stronger than the bolted base.

Some slippage wa's also expected in the slotted bolt holes.

(

Operability was not monitored during the welded base-tests.

Therefore the l

TRS was derated in the region from 9 to 20 Hz.

Spot derating of a TRS is not acceptable. However, if the maximum derating factor of 2.32, as reported in Westinghouse Calculation RTS-D, S.O. no. NAH-149, RTS Welded Base, October 13, 1983, is applied to the whole TRS, the TRS envelops the Seabrook RRS at all frequencies. Therefore, the testing is considered adequate.

Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the reactor trip switchgear is adequately qualified for the prescribed loads.

2.7 Reactor Vessel Level Instrumentation System 8086 Cabinet (NSSS-8)

The reactor vessel. level instrumentation system (RVLIS) cabinets and associated devices,(MPL No. 1-MM-CP-486A) were supplied by Westinghouse with model No. 1726E11. Two of these cabinets will be located at elevation 75 ft of the control building.

At the time of the site visit these I

cabinets were not available for inspection, however, the floor embedments where these cabinets are to be placed was inspected and the cabinet.

installation drawing was provided for review.

The purpose of the RVLIS is to provide reactor vessel liquid level information after a seismic event, These systems are passive during the' seismic event.

f i

10

These cabinets and associated devices were qualified by testing performed by Westinghouse, documented by report WCAP 8687 supplement 2, E53A,. dated March 1983.

Testing of the cabinet and devices cow $isted of a i

low level resonance search test and 5 OBE level and 8 SSE level pseudo triaxial tests. 'The input motion for the OBE and SSE tests was sine beat over the frequency range of 1-33 Hz.

The cabinet natural frequencies were determined to be 12-13 Hz side ~to side, 14-15 Hz front to back, and rigid vertical.

Peak SSE RRS values for elevation 75 of the control building are 3.7 g at 5-6 Hz in the N-S direction, 3.1 g at 8-10 Hz in the E-W direction, and 6.0 g at 8-10 Hz vertical.

These RRS values correspond to 4% damping.

For 5% damping the SSE peak TRS values enveloped 9.3 g in the frequency range of 5-10 Hz for all three directions.

Anomalies were observed at the beginning of the tests. An electrical malfunction occurred, due to frayed wires and a cold solder joint and a-loose fuse holder resulted in intermittent loss of 15 volt power.

These items were repaired and the tests were started over and completed without

'any additional electrical anomalies.

Assembly and inspection procedures were modified and implemented to provide assurance that the anomalies would not reoccur in this equipment in the plants. An additional anomaly observed was a cracked weld in a horizontal door jam. This had no effect on operability or structural adequacy of the cabinet and therefore required no corrective action.

Based on our review of the installation drawing and the qualification documents and the applicant's responses to questions, the RVLIS is adequately qualified for the prescribed loads.

2.8 Nuclear Instrumentation System Cabinet (NSSS-11)

Nuclear instrumentation system (NIS) cabinet (model no.1062E37G03, tag no. IN1-CP-0016) was manufactured and supplied by Westinghouse according to purchase specification no. 953266.

It was located in the control butiding at the 75 feet elevation.

The mounting consisted of 3/16 inch welds in the front and rear (minimum 3 inches at corner columns 11

4 and 6 inches at the center columns). This vertically floor mounted four bay cabinet has common top and bottem rails.

They are also bolted together en its sides.

It provides alarm function, secondary control function of indicating reactor status during startup, power operation and overpower trip protection.

Seismic loads are considered in qualification.

Its qualification is based on tests performed by Westinghouse and documented in reports:

Seismic Testing of Electrical and s

Control Equipment (PG&E Plants), no. WCAP-8021 rev.- 0, dated May 1973 and Equipment Qualification Data Package--Nuclear Instrumentation System (NIS)

Console, no. WCAP-8587, EQDP-ESE-10, rev. 5, dated March 1983.

i The laboratory mounting was different than the field in that it was bolted to the test with eight 3/4 inch diameter A307 bolts (4 bolts per bay). However, this field mounting was shown to be at least as strong and rigid as the laboratory mounting.

It was only a two bay configuration which was tested. Two series of tests were performed on this unit. The first was a resonance series with single axis 0.2 g sine sweep in each of the equipment principal axis. The sweep was from 1 to 35 Hz at a rate of one octave per minute.

The following resonances were indicated:

i s/s f/b v

5.0-7.7 Hz 5.0-7.7 Hz

>33 Hz Subsequently, single axis sine beat tests were performed in the three axis directions. The sine beat consisted of five beats with ten cycles per beat end performed at frequencies of 1.25, 1.75, 2.5, 3.5, 5,.7, 9, 11, 13, 15, 17, 19, 21, 23, 25, 27, 29, 31, 33, 35 Hz plus resonances.

These tests had

~

adequate intensities to envelope the required response spectra.

A damping of five percent was used in the generation of TRS against a four percent RRS.

The operability was demonstrated through another test sequence done by Westinghouse documented in the report:

Seismic Operability Demonstration Testing of the Nuclear Instrumentation System Bistable Ampiifier no. WCAP 8830 dated October 1976.

During this series of tuts, the cabinet fell off the table due to abnormally high input, however, the equipment in it still functioned satisfactorily.

i i

12 t

1

In the sine beat series, the drawer latch mechanism failed which was modified for satisfactory performance in the shipped unit.

This was done by~ bolting the drawer face plate to the cabinet.

Based upon our observation of the field installation, review of qualification reports and applicant's responses to our questions, the NIS cabinet is adequately qualified for Seabrook application having a qualified life of five years.

2.9 Safeguards Test cabinet (NSSS-12)

The safeguards test cabinet, supplied by Westinghouse (model i

no. 1065E21)_.is located in the control building at elevation 75 ft.

The safeguards test cabinet supplies power to the control panel.

~

The safeguards test cabin'et wa,s, qualified by test as discussed in l

Westinghouse report WCAP-8687, supp. 2E16C, Rev. O, Equipment Qualification Test Report Two-Bay Safeguards Test Cabinet (Seismic Desian Verification Testing), April 1982. A 0.2 g resonance search was performed from 1 to 50 Hz with a sweep rate of 1 octave / minute.

The following natural frequencies were found:

S/S--between 19 and 21 Hz; f/b--between 17.5 and 21.5 Hz; V--33 Hz.

The qualification testing, as reported in Westinghouse report EQDP-ESE-16, Rev. 5, Solid State Protection System

~

(SpSS) Two Train (Three and Four Bay) and Safeguard Test Cabinet, March 1983, was done using pseudotriaxial multifrequency input to the test table.

Five OBE tests were performed followed by 1 SSE.. The TRSs used in all tests enveloped the Seabrook specific RRSs.

The damping was 4 percent. The safeguards test cabinet was operable before, during, and after all tests.

Environmental aging has been performed and a qualified life of 5 years has been established.

The tests were performed with the cabinet bolted to the test table.

The cabinet is welded to a channel which is embedded in the concrete in the plant.

UE and C found stresses in the welds to be satisfactory for the 13 f

l p:ak SSE and OBE accelerations taken'from the floor spectra at that lecation. (See UE and C computation sheet--SBMAG-CO-01, Issue no. 1, Data Sheet NSS.387-1E, Safeguard Test Cabinets.)

6 4

l Based on our observation of the field installation, review of the~

qualification reports and the applicant's response to our questions, the safeguards test cabinet is considered adequately qualified for the prescribed loading.

2.10 Instrument Bus Power Supply:

Static Inverter (NSSS-13) i The static inverter (drawing no. 4950C70) was' supplied by Westinghouse according to purchase spec. no. G676573, rev. 5.

It is located in the i

control building at the 21 foot 6 inch elevation.

It is mounted vertically en the floor with 1/4 inch fillet welds 3 inches long at four places (two in the front and two in the rear).

It is a part of the reactor protection l

~

system. The input can be either a 480 V ac 3 phase or 125 V de from battery.

Its output is 118 V ac 60 Hz power to an instrument bus distribution panel that provides power to instrumentation monitoring and l

indicating various plant parameters.

Seismic loads are considered in the qualification.

The static inverter was qualified through tests performed on a similar unit. Tests were performed by Westinghouse and documented in their j

reports:

Seismic Testing of Electrical and Control Eautoment (High l

Saismic Plants), no. WCAP 7821 rev. O dated December 1971 with its i

supplement 5 dated September 1976 and Equipment Qualification Data i

Package--Instrument Bus Power Supply (Static Inverter),

WCAP 8587/EQOP-ESE-18, rev. 5 dated March 1983.

The laboratory mounting consisted of four 5/8 inch A307 bolts.

There were two series of tests I

performed. The first was a resonance series.

It was a single axis sine sweep from 1 to 35 Hz at a sweep rate of one octave.per minute in nach of l

the equipment principal axes.

It indicated the following frequencies:

h-l s/s f/b v

i l

6.8 Hz 6.8 Hz 33 Hz I

14 l

_ __. __ _ _ _,- L _ _ __._.._.__.._... _.~. _. _ -.. ~ - _,._ __,_

The second series was single axis sine beat tests. The sine beat consisted of 5 beats with ten cycles per beat.

These tests were performed at a minimum of eleven frequencies per test direction.

Structural integri.ty and operability were verified with two anomalies.

Large capacitors mounted on a V-shaped bracket vibrated loose and there were broken welds in the frame base.

However, the unit maintained its operability regardless of the mechanical failures.

Mild environment tests per WCAP-8587, supp. 1, EQDP-ESE-18 were also done.

The qualified life based on the aging of the weakest link of the unit is five years.

The tests performed are adequate.

The results are acceptable. An au'ditable link between the field and tested unit which was lacking at the time of review was subsequently provided in document:

Auditable Link Document for Seabrook 1, no. EQAL-NAH, Rev. 4 dated December 1985.

Based upon observation of the field installation, review cf qualification documents, and responses of the applicant to our questions, the unit is adequately qualified for Seabrook 1 site.

1 1

9 i

15

3.

BALANCE OF PLANT (BOP) EQUIPMENT 4

.3.1 36-inch Butterfly Valve (BOP-1)

The 36-in, butterfly valve, supplied by Post-Seal is located in.the etntainment building at elevation 21 ft 6 in.

The model number was not shown on the valve itself and, therefore, could not be compared against the i

number given on the long form. -The valve is used to isolate the

!1 containment. A large Matryx air actuator is mounted on top of the valve with the center line of the actuator horizontal and perpendicular to the j

center line of the pipe. A bracket constructed from 3 x 3 x 3/8 in. angle connects the long end of the actuator to the pipe.

I j

The valve and actuator were qualifted by test in Foreign Print 90615 Issue 1, Butterfly Valve Seismic Vibration Test Report, January 12, 1976.

i.

Wyle Laboratories (see Wyle test report 54598) tested a 30 in. valve with j-the actuator mounted on top with the centerline of the actuator parallel to i

the centerline of the pipe.

No bracket was installed on the valve that was i

tested.

The testing showed many frequencies below 33 Hz:

s/s--31 Hz, f/b--12, 14, 20, 30, 36 Hz and V-14, 29 Hz.

Single frequency biaxial 3 g sine dwell tests were run for 30 seconds at 12, 14, 20, 29, 30, 32, 36, and 40 Hz both in phase and out of phase. The valve was operated during some of the tests.

No natural frequencies below 33 Hz were identified for the valve itself.

(See Foreign Print 91723 Issue 1, Nuclear Seismic Analysis, i

report no. 11473 Rev. B, Post-Seal International, Inc., June 29,1983.)

i The bracket was added to make the valve / actuator assembly stiffer and to remove all natural frequencies less than 33 Hz.

l In Foreign Print 91157 Issue 8, Butterfly Valves Nuclear Seismic Analysis and Addendum A,,10/31/85, Posi-Seal performed hand calculations l

to show that the lowest natural frequency for the field mounting is i

33.9 Hz.

No other qualification testing was performed since no natural I

frequencies were identified below 33 Hz.

l i

16 w

The reviewer accepted the single frequency qualification testing based on the following discussion.

The 36" pipe section is very short and was therefore considered rigid.

The actuator then sees the floor spectra with no amplification.

The peaks of the horizontal and vertical spectra for 2%

damping are 2.03 g at 3-4 Hz and 1.75 g at. 10-12 Hz respectively.

The lowest natural frequencies vertical and front /back are probably the movement of the long end of-the actuator.

These movements have been restrained by the bracket.

Assuming that the next lowest frequency is 20 Hz f/b the acceleration is only about 0.6 g.

For higher frequencies the acceleration decreases.

Therefore not much multimodal contribution is expected and the single frequency qualification testing is adequate.

The qualified life of the valve seal was found to be 5 years based on thermal aging of the seal.

It was also specified that the seal must be replaced if the seal is exposed to radiation.

The qualified Iffe stated on the long form was 24 years.

This comment was relayed to the utility as a request to verify the maintenance procedures, j

Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the 36 in. butterfly valve is adequately qualified for the prescribed loads.

3.2 Control Switch (80P-3)

The control switch (Model No. GE-Shi,I Cat. No. 184B8816G1X2; MPL no. CP-CS-6601-1) was manufactured by General Electric and supplied by York Electro Panel Control Co. according to the purchase spec. no. 170-1.

It was located in the control room at the 75 feet elevation of the control building.

This switch was mounted on a panel with three screws, its standard mounting.

It t!i a part of the rod control and position system.

j It is intended to be used in tripping the reactor manually from the main I

control board.

1 1

1 I

1 1

17 j

a b

The qualification of this switch is based en tests performed on a similar switch in a similar configuration. Details of the tests are in the

' reports:

Seismic Simulation Test Program on Electrical Control Panel, 4

no. 45657-1, rev. O, dated July 23, 1981 by Wyle Laboratories and Final l

Report--Seismic Qualification Test of Main Control Board Zone E, ns. 80127-407, rev. O, dated August 28, 1981 by Analytical Engineering i

Associates, Inc. These two reports were reviewed by-United Engineers and l

Constructors.

Seismic loads were considered in the qualification.

i The tests were performed with the switch mounted on the panel with screws. The panel in turri was welded to the shaker table with 1/4 inch j

fillet weld five inches long both inside and outside.

There were two

{

series of tests performed.

The first series was resonance search.

It was j

a single axis, sine sweep from 1 to 50 Hz in horizontal and 1 to 87 Hz in-the vertical direction with a 0.1 g magnitude.

The sweep rate was one cctave per minute.

The following resonances were indicated for the

]

nounting locations:

l i

s/s f/b v

18 Hz 12 Hz 21 Hz

.l Subsequently a series of qualification tests were performed.

The inputs w2re random, independent biaxial.

TRS were generated for each test.

Two I

and three percent damping values, respectively, were used in TRS generation l

for OBE and SSE.

The same values of damping were used for the RRS.

There ware five OBE and one SSE level tests performed.

The TRS envelop the RRS f

satisfactorily in each case.

Structural adequacy and operability were j,

verified.

No anomaly was detected.

The specimen was aged before the i

seismic tests.

1 i

There were additional tests performed on the switches by General l-Electric Company.

The details are in the report:

IEEE Qualification j

Report for SBI Switches, no. F.P. 73385, rev. O, dated July 26, 1981.

In these tests, the switches were mounted in its standard mounting and placed en the table.

The inputs were random and independent biaxial.

There were i

I 18 i

l five 08E and one SSE level tests performed.

The damping values were five and three percent, respectively, for the TRS and RRS. Aging was also performed on the specimen prior to seismic tests.

Structural adequacy and operability were verified.

No anomaly was detected.

Based on the observation of the field installation and review of the

. qualification documents the control switch is adequately qualified for the Seabrook location.

3.3 Computing Device (BOP-4)

The computing device (model no. NLP; MPL No. EDE-AY-9700) was manufactured and supplied by Westinghouse according to the purchase spec.

no. 174-2.

This card was mounted in a cabinet with two 10-32 " wiz-lock" bolts. The cabinet is located in the control building at the 75 feet elevation.

It is a part of the electrical distribution emergency system.

Its function is signal conversion for monitoring diesel generator output current.

The seismic qualification of the computing device is based on tests performed on a similar unit.

Details of the tests are in the report:

Equipment Qualification Test Report Process Protection System (Supplemental Testing of Printed Circuit Cards), no. WCAP-8687 supp. 2-E13C, rev. O, dated August 1984 by Westinghouse.

It was reviewed by United Engineers & Constructors.

The mounting configuration was identical to that used in the 7300 series two bay cabinet test, i.e., 10-32 " Wiz-lock" bolts. There were resonance tests performed but the resonances were not noted in the Westinghouse's report.

For the qualification, pseudotriaxial, phase coherent, random for OBE and pseudobtaxial, phase coherent, random for SSE inputs were utilized.

There were five OBE and one SSE level tests performed.

A four percent damping was used in the RRS in comparison to a five percent in the TRS. The TRS were compared to the Westinghouse's generic RRS.

The TRS enveloped this generic RRS adequately.

However, Seabrook specific RRS exceeds the Westinghouse's generic RRS.

In response to a question on this apparent drawback, the applicant (through Westinghouse) indicated that a subsequent test had since been performed which showed the adequacy of the tests with respect to the Seabrook 19

specific RRS.

Further that the report was under revision reflecting the qualification of'the device to be adequate.

During the series of tests p3rformed, there were two PC cards which did not function properly during the first and second cycles of the test.

They were replaced and passed subsequent tests.

j t

I Based upon the observation of the field installation, review of the qualification documents and particularly the responses to the questions, the computing device is adequately qualified for Seabrook site.

3.4 Emergency Feedwater Pump and Turbine (B0P-5)

Turbine The emergency feedwater water pump turbine (model no. GS-2N; MPL no. FW-TD-2) was manufactured by Terry Turbine and supplied by Ingersoll-Rand Company according to purchase spec. no. 238-10.

It was located in the emergency feedwater pumphouse at the 27 feet elevation.

In

~

the field the turbine will be attached to a skid with taper pins and guide blocks.

The guide blocks were to be welded with 1/4 inch continuous weld.

They were not complete yet.

The skid to floor mounting consisted of eight 3/4 inch anchor bolts.

This turbine and pump is a part of the emergency feedwater system which provides feedwater to the steam generator during

' loss of heat sink, due to line break, loss of normal feedwater and reactor / turbine trip.

The redundancy is a motor driven pump (MPL No. 1-FW-P-378).

The seismic qualification of this turbine is based on a combination of tests and analysis.

The testing part of the qualification was done by Wyle Laboratories as documented in.the report:

Seismic Testing On One GS-2

~

Turbine, no. 58038, rev. O, dated April 21, 1976.

The report was reviewed by United Engineeis and Constructors (UE&C).

The analysis part was done by Ingersoll-Rand documented in the report:

GS-2N Qualification Report for Ingersoll-Rand (Supplementary Analysis), no. TM-105, rev. O, dated 1

l S:ptember 24, 1979.

It was also reviewed by United Engineers and Constructors.

20

,,_ x - -

r For the test, the turbine was bolted with six 1 inch bolts to a 1 inch plate test fixture. There were two series of tests performed.

The first was a. resonance sear'ch.

It was done with 0.2 g magnitude sinusoidal frequency sweep at one octave per minute.

The following resonances were indicated below 35 Hz:

s/s f/b v

15, 21 & 27 Hz 22, 32 Hz 22.32 Hz The qualification tests consisted of pseudobiaxial random inputs.

The required ZPA were:

s/s f/b v

OBE 0.47 g 0.39 g 1.39 g SSE 0.74 g 0.64 g 1.60 g The tests ZPA were s/s f/b v

OBE 2.0 g 2.0 g 3.0 g SSE 2.0 g 3.0 g 3.2 g TRS were generated for each case.

Two percent damping was used for both TRS and RRS.

For OBE and SSE, TRS do not envelop the Seabrook RRS for the vertical direction between about the 16 to 26 Hz region.

The equipment has resonant frequency in this region.

On inquiry, it was revealed that UE&C was aware of the situation and in the process of resolving it.

However, there are a number of anomalies (too many to enumerate) as reported in the Wyle Laboratory report.

Some of them appear to indicate modifications and changes. A satisfactory disposition of all of them is to be confirmed.

The analysis part consists of static coefficient method with a two dimensional model for stresses.

The calculated stress in the #9 taper pins at the base of the turbine with seismic, nozzle and deadweight is 48127 psi egainst an allowable of 52200 psi.

21

Pump j

The pump (model no. NH, MPL no.1-FW-P-37A) was manufactured and i

supplied by Ingersoll-Rand Company (IRC) according to the purchase spec.

no. 238-10.

It is also skid mounted as the turbine but has #10 taper pins.

l.

The qualification of the pump is based on analysis performed by (PDCI)

Poly'ecnic Design Co. Inc.

The details are in the report:

Seismic j

t Qualification. 4 x 9 NH-10 Emergency Feedwater Pump by PDCI and Structural l

Integrity and Operability Analysis of 4 x 9 NH-10 Turbine Driven Emergency l

Feedwater Pump, No. EAS-TR-8001, re'v.1,' dated June 25, 1984.

The i

frequency calculations show that the pump is relatively rigid. Therefore, i

a static analysis has been used to calculate the stresses and deflections.

l A finite element technique with isoparametric 3-D solid elements with the ANSYS computer code is used for analysis.

Seismic loads in conjunction

]

with other loads are considered in the analysis.

The calculated stress in

!l the pump shaft is 7430 psi against an allowable of 7500 psi (Ingersoll-Rand j

Design Criteria).

The total shaft deflection at the seal is calculated to be 0.004 inch against manufacturer allowable of 0.005 inch.

This indicates I

that the pump is structurally adequate and can operate.

f Based on the observation of the field installation, review of the I,

qualification reports and responses to the questions provided by the applicant, the turbine and pump assembly is qualified for Seabrook 1 application pending confirmation the following:

l a.

The anomalies as reported in the Wyle report are satisfactorily resolved, and j

b.

Seismic qualification of the temporary 3-inch drain line (which may be a permanent fix) is performed and found adequate.

3.5 4-inch Motor Operated Globe Valve (BOP-6)

The 4-in, motor operated globe valves (MPL No. MS-V-204, 205, 206, i

207) were supplied by Rockwell International with vendor serial 22 i

.No. BE 728.

The four valves were purchased to specification No. 248-65.

The valve's motor operators are Limitorque model no.- SMB-00-10.

The valves ar,e line mounted (full penetration welds) in 4 inch bypass lines around the main steam isolation valves.

The operators are bolted to the valve bodies with eight 5/8 in. diameter bolts.

The valve was qualified by a combination of test and analysis performed by Rockwell International.

The valve was. analyzed in accordance with the requirements of ASME Code Section III for class 2 valves.- This analysis is documented by report no. RAL-3142. A three dimensional computer model of the 4 inch bypass piping, valve and operator was developed to determine its natural frequency.

It's natural frequency was determined to be 39.5 Hz from this analysis.

A static deflection test was-performed on a similar valve and operator assembly to demonstrate operability of the valve.

This test consisted of imposing nozzle loads on the valve and displacements on the operator corresponding to seismic inertia loading.

This testing was d'ocumented by test report no. RAL-1073.

The operator was qualified separately testing documented by Limitorque report No. B0058.

The operator was tested using single axis sine dwell test for the frequency. range of 5 to 35 Hz.

The operator was bolted directly to a rigid test table.

No natural frequencies were found below 33 Hz.

The test acceleration levels were all above 3 g's and up to 6 g's horizontal and 3.2 g vertical at 35 Hz.

Operability of the operator was demonstrated, with no observed failures or anomalies.

Based on our inspection of the field insta11aticn, review of the I

qualification documents, and the applicant's responses to questions, the 4 in globe valve and operator are adequately qualified for the prescribed loads.

o 3.6 Neutron Flux Stonal Processor (80p-7)

The neutron flux signal processor, supplied by Gamma-Metrics (model 900043-101), is located in the control butiding at elevation 23

r 21 ft-6 in. The model number shown on the long form did not match that shown on the actual piece of equipment.

The processor gives the operator an indication of neutron flux and shutdown margin.

The processor was qualified by test in Foreign Print 73327 Issue 2, W'le Laboratories test report 58826, Rev. A, RCS Series Seismic Test y

Report, 5/23/83. A 0.2 g sine sweep test from 1 to 40 Hz with a sweep rate of one octave per minute was performed and two natural frequencies below 33 Hz were identified.

They were 20 Hz S/S and 26 Hz vertical.

No f/b resonances below 33 Hz were found.

Qualification testing was done using a biaxial random motion input containing frequencies from 1 to 50 Hz 1

t with incoherent phasing.

Three different mounting orientations of the processor with 90 degree angle rotations from each other were used to simulate motion in each direction.

Five OBE tests and 1 SSE test were performed in each orientation.

The damping values used for OBE and SSE ware 2 and 3 percent, respectively.

Each TRS for all tests enveloped the S:abrook specific RRSs for both OBE and SSE.

One anomaly occurred during one of the SSE tests.

The shaker table reached its maximum stroke capability in the vertical direction resulting in a shock impact to the test specimen.

However, no structural damage cccurred during any of the tests and the processor functioned before, during, and after all tests.

The test mounting adequately simulated the field mounting.

A shutdown margin isolator and 4 other isolators are mounted to a circuit board inside the processor.

The shutdown margin isolator was icosely mounted to the circuit board but the other isolators appeared to be tighter. All of the isolators must be operational for the processor to perform its function during testing.

The processor did function during testing and, furthermore, the vendor stated that the unit tested was id:ntical to the one in the plant.

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24

r Irradiation tests and extreme temperature and humidity tests have been performed for aging considerations.

However, a qualified life has not yet been established for the processor.

Impe11 Corporation performed the-environmental testing but the report is still under review by UE and C.

The environmental testing is reported in Foreign Print 73326 Issue 2, RCS Series Qualification Test Report, April 1983.

Based on the~ observation of the field installation, review of the qualification documents, and the applicant's response to questions, the neutron flux signal processor is adequately qualified for the prescribed loads.

3.7 Vibration Monitoring Control Panel (80P-11)

The vibration monitoring control panel, supplied by Technology for Energy Corp..is located in the control building at elevation 75 ft. The model number was not shown on the panel itself and could only be found by referring to the drawing. The model number shown on the drawing was the number included on the long form.

This panel indicates the position of_the pressurizer relief valves to the operator.

The TEC model 1414 valve flow monitor is a bank of centrols mounted on the TEC model 158 cabinet.

The TEC model 158 cabinet at Seabrook was built before November 1983 and the cabinet tested was slightly different.

Foreign Print 73784 Issue 1, Addendum A to Seismic Qualification Test Report 30080-TR-02 discusses the differences and concludes that there is no reduction in seismic integrity for the 158 cabinet, compared to the one i

tested.

Foreign Print 72975 Issue 1, Test Report TEC 1430-4 Valve Flow Monitor System, September 1981 discusses the testing program.

Five OBE and 1 SSE tests were performed. 'The TRS exceeded the RRS at'all frequencies for all tests. The valve flow monitor system was functional before, during, and after the tests.

No abnormalities occurred during testing.

25

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The Model 158 cabinet was bolted in the test and, welded in the plant.

UE and C has a program in progress to evaluate the differences in bolted and welded mounting conditions.

It is, felt that the evaluation program is adequate.

Based on our observation of the field installation, review of the qualification reports and the applicant's response to our questions,'the vibration monitoring control panel is considered adequately qualified for the prescribed loading.

3.8 18-inch Feedwater Isolation Valve (BOP-14) i The 18-in. feedwater isolation valve (MPL no. FW-V-30) was supplied by Borg-Warner Co with model no. 73890.

Four of th'ese valves were purchased to specification No. 248-36.

These valves are pneumatic-hydraulic operated 18 inch, 900 lb rated gate valves. They are line mounted with full I

ptnetration welds at elevation 8'-3" of the mainstream feedwater pipe 1

j chise. The valves are horizontal with vertical actuators.

Qualification of these valves was performed by a combination of tasting and analysis.

The valves were analyzed in accordance with the r:quirements for ASME code section III for class 2 valves.

The valve analysis is documented by Borg Warner Report no. 73890 dated July 15, 1982. The valve body maximum calculated stresses were 24,610 psi compared to an allowable stress of 42,000 psi.

The valve and operator were tested for operability by a static deflection test which included nozzle loads and operator loads simulating 3.0 g seismic inertia loading.

This test is dscumented by Borg Warner report no. OTP73890 datcd Dec 1, 1980.

The valve operator was qua'lified for seismic loading by separate biaxial random motion tests.

The SSE test motion was in excess of 10 g and j

the TRS enveloped the plant generic RRS for line mounted equipment for all l

frequencies above 4 Hz.

The operator was subjected to 5 SSE tests and 1 OBE test in each test direction (horizontal 1/ vertical and (horizontal 2/ vertical).

Operability of the operator was demonstrated.

A total of four anomalies were observed during the qualification of the 26

e i

operator.

They were 1) "0" ring seals were nicked, 2) hydraulic fluid absorbed moisture, 3) "0" ring seal took a set, and 4) a solenoid valve rupture disc ruptured.

Design and maintenance procedure changes for the operator were incorporated to prevent reoccurrence of these anomalies.

These were 1) the nicked o-ring was determined to have occurred at installation or upon removal and required no design change, 2) the hydraulic fluid was changed every 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months during environmental testing and the maintenance procedures were modified to require the oil to be changed every 2 years which corresponds to the accelerated aging test duration of 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months , 3) the o ring seal design was modified to provide a better seal,

4) the solenoid valve rupture disc was determined not to be required and the design was changed to remove the rupture disc.

l Based on our inspection of the field installation, review of the qualification documents, and the applicant's responses to questions, this valve and operator are adequately qualified for the prescribed loads.

3.9 6-inch Motor Operated Gate Valve (80p-15)

The 6-in. motor operated gate valve (Mpl No. CBS-V-38) was supplied by Walworth Co. with serial no. A 3699.

It was purchased to specification q

No. 248-41.

The motor operator for this valve is a Limitorque model No. SMB-000-5.

The valve is line mounted using full penetration welds at elevation 22'-3" of the tank farm.

The motor operator is bolted to the valve body with sixteen 5/8 1.

diameter bolts.

This valve is an active valve located in the containment spray system.

The valve was qualified by a combination of analysis and testing.

The valve was analyzed in accordance to the requirements of ASME code section III for class 2 valves.

The valve body stress was calculated to be 20,292 psi compared to a code allowable of 26,250 psi. The valve and operator assembly was tested by Action Environmental Testing' corporation.

The tests consisted low level resonance search and pseudobiaxial sine beat tests.

The resonances found were 27.5 Hz side to side, 30.5 Hz, front to back, and rigid vertical.

27 1

S

r The pseudo biaxial tests were performed for the frequency range from 4 Hz to 33 Hz with an acceleration level of 3.0 g in all test directions.

The valve was mounted with an imposed nozzle load of 33,810 in-lbs for these-tests.

The valve was tested in 4 positions to account for the use of coherent test motion. Operability of the valve and operator was demonstrated with no observed failures or anomalies.

The operator was also qualified by a separate test report by Limitorque (Test Report B 0058).

See BOP-6 for the details of that testing.

Based on our inspection of the field installation, review of the qualification documents, and the applicant's responses to questions this valve and operator are adequately qualified for the prescribed loads.

3.10 Diesel Generator Relay Control Panel (80P-16)

The diesel generator relay control cabinets (MPL no. DG-CP-36, 37) were supplied by Colt Industries with model no. 11-871-638.

These two cabinets and associated devices were purchased to specification no. 201-1.

The cabinets a supported off the diesel generator skids at elevation 21'-6" of the diesel generator building.

The cabinet is supported to the diesel generator skip by four gusset supports which are welded to the skid and angle frame around the exterior back o'f the cabinet.

The relay control cabinet and associated devices were qualified by testing performed by Wyle Laboratories and documented by report No. 44011-1 dated May 24, 1978. Test mounting was more flexible than field mounting of the cabinet in that the angle framework was not included.

Gussets were welded directly to the cabinet for testing.

Testing consisted of 5 OBE level and 1 SSE level test in each test direction.

The tests were biaxial phase incoherent random motion tests.

The TRS enveloped the RRS for each test.

Operability of the devices was demonstrated however two anomalies were observed. A cracked insulator and a broken bolt were found.

Adequate justification for these anomalies was not provided.

28

r i

' Based on our inspection of the field installation, review of the qualification documents, and the applicant's response to questions, the diesel generator relay control is adequately qualified for the prescribed loads pending adequate justification is provided for the test anomalies.

3.11 Pressure Switch (80p-17)

The pressure switch, supplied by Detroit Switch (model 222-10) is mounted on the diesel generator at elevation 21 ft-6 in.

This switch regulates lube oil flow to the diesel engine.

The pressure switch was a surprise item and the long form and qualification documentation were examined only for completeness.

A thorough review of the documentation was not performed. The qualification package appeared complete but it is still under review by UE and C.

l l

0 29

T

)

u.

4.

FINDINGS AND CONCLUSION The review of the Seabrook Station, Unit I will be completed when the following open items are closed.

4.1 Generic Issues 4.1.1 During the field observation of the nuclear instrumentation system cabinet it appeared that the clearance between this unit and the adjacent solid state protection system train B was not adequate. On inquiry, it was further learned that this probled was associated with many other cabinets as well.

However, the applicant was aware of the problem and indicated that the analysis and resolution of the problem was being actively pursued.

A final and satisfactory resolution of'this problem, on g'eneric basis, shculd be confirmed to the NRC.

4.1.2 During the documentation review of the reactor make-up water valve (RAWV-30: NSSS-5), it was discovered that the g-loading assumed fcr the valve qualification has not been reconciled with the as-built condition.

This is true for other valves as well.

However, the applicant indicated that a program was already in place and in progress to do the r: conciliation.

NRC should be informed when the program is completed.

4.1.3 The evaluation of the life-span of nonmetallic parts for the 3-inch valve-air operated has not been performed.

This problem exists with many other equipment items.

1 30

4 v

It should be performed and confirmed to the NRC as to its completion.

4.2 Equipment Specific Issues 4.2.1 The review of the report prepared by Wyle Laboratories of the tests performed on the diesel generator relay control cabinet (80P-16), revealed a number of anomalies.

These are detailed in the report.

In order for the item to be adequately qualified, these anomalies should be satisfactorily resolved.

4.3 Confirmatory Issues 4.3.1 The Wyle Laboratories tests on the emergency feedwater pump turbine (Terry Turbine:

BOP-5) reported a substantial number of anomalies (too many to enumerate here).

Some of them appear to need field modifications.

A confirmation of satisfactory disposition of all the anomalies is required.

4.3.2 During the field observation of the emergency feedwater pump turbine (BOP-5), it was found that a temporary 3-inch drain line was installed.

It was also reported that the line might become a permanent fix.

Seismic adequacy of the line should be established if it became a permanent fix.

This should be confirmed to the NRC.

l

)

4.4 Conclusion Based on our review, we conclude that, pending resolution of all open items, an appropriate qualification program has been defined and implemented for the seismic Category I mechanical and electrical equipment 31

which will provide reasonable assurance that such equipment will function properly during and after the excitation due to the vibratory forces iCposed by safe shutdown earthquake in combination with normal operating loads.

e

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TABLE 1.

LIST OF ATTENDEES V. Nerses NRC/NRR/DL/LB#3 C. D. Greiman UE&C/EQTF K. C. Robertson UE&C H. Flora UE&C D. A. Mehta UE&C Phila.

S. N. Caruso UE&C Seabrook J. P. Namburd UE&C Phila.

J. J. Parisano UE&C Phila.

K. M. Kalawadia UE&C Phila.

H. J. McGiuley UE&C Phila.

J. A. Herrera UE&C Phila.

J. D. Brundt UE&C Phila.

E. Pilhuj UE&C Phila.

~

V. W. Sanchez NHY SB J. W. Stacey YNSD Feam.

L. Cerra CP&L N. Carolina R. E. White YNSD R. K. Tucker YNSD Fram.

N. D. Romney NRC/DE/EQB G. Bagchi NRC/DE/E08 J. N. Singh EG&G Idaho Inc.

T. L. Bridges EG&G Idaho Inc.

B. L. Harris EG&G Idaho Inc.

R. E. Cyr NHY Seabrook J. L. Cad'e NHY Seabrook S. H. Dunphy UE&C Seabrook H. M. Stronberg EG&G Idaho Inc.

Clark Kido EG&G Idaho Tony Chung EG&G Idaho George O'Connor YSND D. Ruscitto NRC Resident Inspector NRC j

A. Cerne Sr. Res. Insp.

NRC G. Sessler Engineering Services NHY W. Dickson Engineering Services NHY J. Martin Tech Services NHY W. F. Gurein Westinghouse. Nuclear Safety C. G. Draughon W. Mgr., I&C Syst. Lic., N. Safety Dept, l

R. M. Span W. Nuclear Safety G. Rigamonti UE&C D. Maidrand YNSD Asst. Pro. Mgr.

D. McLain NHY Startup Mgr.

R. P. Neustadter UE&C Phila. I&C SDE E. Skolnick UE&C Phila. Mag.

T. C. Feignebaum NHY IRT Leader R. C. Julian YAEC/QA Review Supervisor R. E. Gillette Asst. Constr. QA Mgr.

NHY G. F. Mcdonald Constr. QA Mgr.

NHY N. J. DeLoach YNSD Proj. Mgr.

YNSD 33

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.a., oar a w.,ra TJ;">'#,'-

situOGRAPHIC DATA SHEET EGG-EA-7163

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i Seabrook 1 SQRT Visit Report

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February 1986 u.or o...

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.o a.O f e.

T0 4 J. N. Singh, T. L. Bridges B. L. Harris February 1986 l

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e a'*ecma*=== va" ""*' a EG&G Idaho, Inc.

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P.O. Box 1625 i

Idaho Falls, ID 83415 A6415

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... e,*eo,asaoar Division of Pressurized Water Reactor l

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

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1 EG&G Idaho, Inc., is assisting the Nuclear Regulatory Commission in evaluating Pubite Service Company of New Hampshire's program for the dynamic qualification cf safety related electrical and mechanical equipment for Seabrook 1 nuclear power plant. Applicants are required to use test or analysis or a combination

~

of both to qualify equipment, such that its safety function will be insured during and after the dynamic event, and provide documentation. The review.

when completed, will indicate whether an appropriate qualification program has been defined and implemented for seismic Category I mechanical and electrical cquipment which will provide reasonable assurance that such equipment will function properly during and after the excitation due to vibratory forces j

of the dynamic event.

i

...................x....o.

...,,.d'?i'Only as specifica11) approvedy de WtW.Nt'6.W.'ichhh

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Unclassified nr Unclassified

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ML20203A487

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ML20203A487

Text

Dear Mr. Harrison:

SUBJECT:

Enclosure:

Dear Mr. Harrison:

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Enclosure:

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