ML18037A008

Equipment Qualification Program. W/840531 Ltr

ML18037A008
Person / Time
Site:	Nine Mile Point
Issue date:	05/31/1984
From:	Mangan C NIAGARA MOHAWK POWER CORP.
To:	Vassallo D Office of Nuclear Reactor Regulation
References
PROC-840531, NUDOCS 8406070290
Download: ML18037A008 (425)

Text

REGULATOR)INFORMATION DISTRIBUTION S .. Eh1 (RIDS)

ACCESSION NBR:8406070290 DOC,DATE: 8'/05/31 NOTARIZED: NO DOCKET ¹ FACIL:50-220 Nine Mile Point Nuclear Station< Unit ir Niagara Powe 05000220 AUTH, NAME AUTHOR AFFILIATION MANGANrC~ Ve Niagara Mohawk Power Corp.

REC IP ~ NAME RECIPIENT AFFILIATION VASSALLOrD,B, Operating Reactors Branch 2

SUBJECT:

"Nine Mile Point Unit 1 Equipment Qualification Program,"

VJ/840531 I tr, DISTRIBUTION CODE: A048S COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR/Licensing Submittal: Equipment Qualification NOTES; REC IP IENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL HERMANNrR 01 '1 NRR ORB2 BC 12 1 0 1 INTERNAL: ELD/HDS3 12 1 1 GC 13 1 IE FILE 09 1 1 NRR KARSCHiR 1 1 NRR/DE/EQB 07 2 2 NRR/DL DIR 1I1 1 1 NRR B 06 1 1 NRR/DSI/AEB 1 1 FI 00 1 1 RGNl 1 1

.-.EXTERNAL: ACRS 15 8 8 LPDR 03 NRC PDR '02 1 1 NSIC 05 1 1-NTIS 31 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 20

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A ~ I N 4'IAGARA MOHAWK POWER CORPORATION/300 ERIE BOULEVARD WEST. SYRACUSE, N.Y. 13202/TELEPHONE (315) 474"-1511 May 31, 1984 Director of Nuclear Reactor Regulation Attention: Mr. Domenic B. Vassallo, Chief Operating Reactors Br anch No. 2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555 Re: Nine Mile Point Unit 1 Docket No. 50-220 DPR-63

Dear Mr. Vassallo:

Attached is a status of the Environmental gualification Program for Nine Mile Point Unit 1. This information was presented to members of your staff at a meeting on March 15, 1984.

Very truly yours, C. V. Mangan Vice President Nuclear Engineering 5 Licensing CVM/MGM:ja Attachment

NINE MILE POINT UNIT 1 EQUIPMENT QUALIFICATION PROGRAM MAY 1984 840b070290 0500 8400220 PDR ADQCK PDR P

~l y TABLE OF CONTENTS Section Oescri tion Generic Positions Equipment Items Considered Resolved oy the Technical Evaluation Report III Resolution for Specific Equipment Environmental qualification Oeficiencies Identified in the Technical Evaluation Report IV Justification for Continued Operation

Section I Generic Positions Compliance with 10CFR50.49(b)

Safety-Related Equipment Nonsafety-Related Equipment Certain Post Accident Monitoring Equipment Completeness of Safety-Related Systems List, Equipment List and Oisplay Instrument List Qne Hour Minimum Operating Time Margin Aging and gualified Life Maintenance and Surveillance Submergence Chemical Spray Installed TMI Action Plan Items

4 )

A. Compliance With 10CFR50.49 b Paragraph (a) of 10CFR50.49 requires that each licensee establish a program to environmentally qualify electrical equipment. 10CFR50.49(b) groups this equipment into the following three categories:

Safety-related electrical equipment as defined in IEEE Standard 323-1974 and 10CFR50.49.

2. Nonsafety-related electrical equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions of the safety-related equipment.

3. Certain post-accident monitoring equipment.

The following discussions present the methodology used to identify all electrical equipment falling within the above three categories:

10CFR50.49 b 1 Safety-Related Electrical Equi ment Generic position B nerein (Completeness of Safety-Related Systems List, Equipment List and Oisplay Instrument List) describes the methodology used to identify safety-related equipment for Nine Nile Point Unit 1. This methodology is in full compliance with the requirements of Inspection and Enforcement Bulletin 79-018 and 10CFR50.49. System Component Worx Sheets (SCEWS), in accordance with Inspection and Enforcement Bulletin 79-018 format, are computer generated to facilitate nandling and distribution of qualification data within the organization for the ongoing qualification effort. - All postulated design basis accidents wnicn could potentially result in a harsh environment, such as loss of coolant accident inside containment, nigh energy line breaks outside containment and flooding inside and outside containment were considered in the identification of electrical safety related equipment. The master list of electrical equipment requiring environmental qualification addresses all electrical equipment within the scope of 10CFR50.49(b)(1).

10CFR50.49 b 2 Nonsafety-Related Fquipment 10CFR50.49 includes in its scope nonsafety-related electrical equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions of the safety-related equipment. The possibility of failure of nonsafety-related equipment in a manner detrimental to safety-related equipment has been evaluated by a combination of methods wnicn are summarized below.

A master list of equipment was developed in accordance with

.10CFR50.49(b)( 1) and the requirements of Inspection and Enforcement Bulletin 79-01B. This equipment is required to provide safe shutdown and to mitigate the consequences of design oasis accidents such as a 'loss of coolant accident inside containment and hign energy line bream outside containment. Generic position B nerein (Completeness of Safety-Related Systems List, Equipment- List and Oisplay Instrument List) describes the metnodology used to identify the electrical equipment requiring environmental qualification.

Not all the equipment in a particular safety-related system is required to accomplish accident mitigation and safe shutdown. A system failure analysis was performed on each safety-related system to identify the set of electrical equipment which the system requires in order to perform its design basis safety function. This analysis identified the nonsafety related auxiliary systems and equipment which are necessary for the operation of the safety related system or equipment. This analysis included a detailed review of tne system'peration, systems interaction, and operation of equipment witnin each system. This effort included review of the plant safety analyses, emergency operating procedures, piping and instrumentation diagrams, elementary wiring diagrams, electrical one line diagrams and functional logic diagrams. Addition or deletion of equipment from the master equipment list was performed as necessary.

In summary, a master list of equipment was developed for safety related electrical equipment within the scope of 10CFR50.49(b)(l).

This master list was based on review of tne plant safety analyses, emergency operating procedures, tecnnical specification, piping and instrument diagram and electrical drawings. A system failure analysis was performed to identify non-safety related systems or equipment which could potentially degrade the operation of safety related systems or equipment.

Based on the above analysis, we have not classified any electrical equipment as nonsafety-related whose failure under postulated environmental conditions could prevent accomplishment of required safety functions oy the safety related equipment. Therefore, the current master list of electrical equipment requiring environmental qualification and the review methodology adequately addresses electrical equipment within tne scope of 10CFR50.49(b)(2).

10cFR50.49(b 3) certain Post-Accident i~lonitorin Equi ment Paragrapn (b)(3) of 10CFR50.49 includes in its scope "certain post-accident monitoring equipment." Specific guidance concerning parameters to be monitored is provided in Regulatory Guide 1.97, Revision 2. Our generic position with respect to this issue and the methodology that was used to identify equipment tnat falls witnin this category is presented below..

Oisplay instrumentation is included as an integral part of our qualification program in accordance with requirements estaolisned oy Information and Enforcement Bulletin 79-01B. Generic position 8 herein (Completeness of Safety Related Systems List, Equipment List and Oisplay Instrumentation List) describes tne metnodology used to identify display instrumentation. Equipment that currently falls within the category classified as certain post accident monitoring equipment was selected based on the following:

Oisplay instrumentation wnicn is exposed to a narsh environment following a design oasis accident is identified in the plant emergency operating procedures and are used oy the operator to diagnose performance of system. safety functions. This equipment is incorporated into the qualification program in accordance with the requirements of Inspection and Enforcement Bulletin 79-018 and enclosure 2 (Oivision of Operating Reactor guidelines).

Certain installed electrical equipment located in harsh environments required for TNI Lessons Learned Implementation (NUREG-0737) in accordance with I.E. Bulletin 79-01B Supplement 3, Item 2.

At this time, the following activities nave not been completed for Nine Nile Point Unit 1:

The Oetai led Control Room Oesign Review including tne revision to the Plant Emergency Operating Procedures.

Regulatory Guide 1.97, Revision 2 Review "Instrumentation for Light-Mater-Cooled Nuclear Power Plants .to Assess Plant and Environs Conditions Ouring and Following an Accident."

As these activities are completed, equipment considered to be classified as Regulatory Guide 1.97, Revision 2, Category 1 or Category 2 items will be qualified in accordance with 10CFR50.49 criteria. qualification of these items will be independently scheduled.

Based on the above considerations, electrical equipment within the scope of 10CFR50.49(b)(3) has been adequately addressed and incorporated into the Nine bile Point Unit equipment 1

qualification program.

Completeness of Safety-Related Systems List, Equipment List and Ois lay Instrument L>st The Tecnnical Evaluation Report (TER-C5257-466), Appendix C, addressed the issue of completeness of the lists of safety-related systems, equipment and display instrumentation at Nine Mile Point Unit l. In Section C.2, the Tecnnical Evaluation Report indicated tnat the licensee had responded to the NRC regarding tne list of safety-related display instrumentation. In Sections C.l and C.3, nowever, the Tecnnical Evaluation Report expressed concerns over the lists of safety-related systems and equipment. In summary, the concerns expressed in the Technical Evaluation Report were:

The list of systems was insufficient to verify that all safety functions will be performed The deletion of equipment items or display instrumentation from the lists because they fell within the scope of Regulatory Guide 1.97 is not justified The absence of certain specific equipment items from the list was questioned.

The TER concluded with the following recommendation:

"It is recommended that a thorough review of plant safety analyses and emergency procedures oe performed with regard to the safety functions necessary for loss of coolant accident and nigh energy line brea'ccident mitigation. A complete and comprenensive list of systems to be addressed for environmental qualification snould be submitted to the Nuclear Regulatory Commission for review and approval."

In view of these concerns, we undertook a complete review of the lists of safety related systems, equipment and display instrumentation. This review was intended to provide tne recommended action of the TER and was performed as follows:

A review of all design basis events such as loss of coolant accident inside containment and high energy line breaks outside containment (in reactor building, steam tunnel and turoine building) was conducted.

A list of systems required to mitigate tne consequences of a loss of coolant accident and high energy line breaks was estaolisned.

The list was based upon a review of plant safety analyses, technical specifications and emergency operating procedures, considering the functions that must oe performed for accident mitigation without regard to location of equipment relative to a potentially harsh environment. The six functions considered were:

( 1) emergency reactor shutdown, (2) containment isolation, (3) reactor core cooling, (4) containment heat removal, (5) core residual heat removal, and (6) prevention of a significant release of radioactive material to the surrounding environment.

A list of display instrumentation (including tnose necessary for the operator to monitor plant status) was developed with the systems list. Tne primary objective was to establish a final list of safety-related systems in accordance with Huclear Regulatory Commission criteria.

Not all equipment in a particular safety-related system requires qualification and post-accident active or passive functional capaoi lity in order to accomplish accident mitigation. Oepending on system design, certain motor-operated valves, solenoid-operatea pneumatic valves, temperature switches, limit switches and instrumentation may not be required to perform a safety function or mitigate the consequences of an accident in order for the system to accomplisn its design oasis safety function. Several other systems only require that the containment isolation portion of the system remain functional. A system failure analysis, therefore, was performed to identify the set of electrical equipment which is required in order to perform its design oasis safety function.

Addition or deletion of equipment from the master list was performed as necessary. Oivision of Operating Reactors Guidelines Appendix A and plant emergency operating procedures were used as guides to identify devices and display instrumentation used by the operator. The equipment which must function in these systems was identified by review of system descriptions and appropriate drawings (piping and instrumentation, drawings, elementary wiring diagrams, electrical one line diagrams and functional logic diagrams). Application of system/component failure analyses was performed to identify the electricai equipment wnicn requires environmental qualification.

C Plant areas with environmental parameter s (pressure, temperature, humidity, radiation level, submergence level, etc.) which increase significantly above normal ambient conditions as a result of a design basis event, were considered to be harsh post-accident areas. Containment sprays and radiation dose from recirculating radioactive fluids were included in these considerations.

A review of the location of tne electrical equipment was performed. Equipment items which are required to function but are not located witnin a harsh environment, were deleted from the list. In addition, certain equipment items not exposed to a narsh environment at the same time that tney are required to perform their safety-related function were deleted from the list and justification was provided.

Based on the results of the above tasks, a final safety-related systems list and a final electrical equipment list (including display instrumentation) were developed.

Niagara Mohawk believes that tnis review satisifes each of the concerns expressed in Appendix C of the Tecnnical Evaluation Report. Tne resulting systems list is completely documented to demonstrate that all required safety functions will be performed. Display instrumentation identified in the plant emergency operating procedures are included.

C. One Hour Minimum Operatin Time Mar in In order to account for various uncertainties innerent in equipment qualification test programs, the NRC criteria for qualification incorpor ated a one-hour minimum time margin requirement in addition to the required operability time of equipment. This requirement was established by the Oivision of Operating Reactors Guidelines Section 5.3.1, Inspection and Enforcement Bulletin 79-018, Supplement 2 question/Answer Number,12 and NUREG-0588, Section 3(4). Even thougn some equipment was required by design to perform its safety function within a short time period after tne onset of an accident, the Nuclear Regulatory Commission criteria required that this equipment remain functional in the accident environment for a period of at least one hour in excess of the design operating time for the equipment. The Nuclear Regulatory Commission Safety Evaluation Report/Tecnnical Evaluation Report used this criteria in tne review of the licensee's equipment qualification documentation.

Subsequently, the Nuclear Regulatory Commission issued Generic Letter 82-09 which stated tnat equipment may be qualified using tne required operating time plus an appropriate margin. This criteria is applicable to equipment subject to the requirements of the Oivision of Operating Reactors Guidelines or Category II of NUREG-0588. In addition, tne one hour time margin is not applicaole to equipment wnose safety function is performed prior to significant cnanges in tne environment. For all cases, however, subsequent failure must. be SnOwn nOt to be detrimental to plant safety. Regulatory Guide 1.89, Revision 1, position C-6 also states that equipment wnicn is required oy design to perform its safety function within the first ten nours of tne event snould remain functional in the accident environment for a period of at least one nour in excess of the required equipment operating time unless a time margin of less than one hour can oe justified. This justification must include:

Consideration of a spectrum of line breaks Potential need for use of the equipment )ater in the event Determination that failure of the equipment after the required operating time interval will not degrade safety functions or mislead the operator Determination that margin applied will account for uncertainties in the qualification program.

10CFR50.49(e)(8) states "Margins must be applied to account for unquantified uncertainty, such as the effects of production variations and inaccuracies in test instruments. These margins are in addition to any conservatisms applied during the derivation of local environmental conditions of the equipment unless these conservatisms can be quantified and shown to contain appropriate margins."

Our position with respect to the one hour minimum operating time margin is in accordance with the criteria presented in Generic Letter 82-09, Regulatory Guide 1.89, Revision 1, position C-6 and 10CFR50.49(e)(8).

Test data and analysis used to demonstrate qualification of equipment envelope the required design operating time plus one hour margin or an appropriate margin has been justified.

A in and /uglified Life The Nuclear Regulatory Commission Division of Operating Reactors guidelines, Sections 5.2(4) and 7.0, require that the licensee conduct an assessment of safety- elated equipment to identify materials susceptible to significant age related degradation which could affect performance of design safety functions. A qualified (designated) life should be established for equipment susceptible to significant aging based on engineering evaluations and judgment. Maintenance, surveillance and equipment or component replacement intervals should be based on the established qualified life so that equipment qualification is maintained on a continuing oasis. Specifically, the Division of Operating Reactors guidelines require; identification of materials susceptiole to significant degradation due to tnermal and radiation aging, establishment of maintenance and replacement schedules, and establishment of ongoing programs to review surveillance and maintenance activities to identify equipment exnibiting age related degradation.

Arrhenius techniques are generally considered acceptable for assessment of thermal aging. These requirements are also implicitly established by 10CFR50.49(e)(5), NUREG-0588, Category I, Section 4 and Regulatory Guide 1.89, Revision 1, Section 7; however, for new equipment (replacement equipment), these standards are more rigorous in that the criteria of IEEE-323 ( 1974) must be applied and the equipment must be preconditioned prior to testing. Methods for compliance with established criteria are presented below.

For installed equipment, we have identified safety-related equipment whose materials are susceptible to significant age related degradation.

A qualified (designated) life nas been estaolisned for each equipment type with requisite replacement or component refurbishment schedules.

Various methods were employed in establishing the qualified life for equipment such as:

Use of available qualification test data on similar or actual components or equipment to support a conservative equivalent life extrapolation of the enveloping temperature test profile using Arrhenius techniques Contact with vendors to obtain bills of materi'al, material information and technical data to identify age sensitive materials Review and engineering evaluation of industry references and technical literature to determine material radiation threshold and thermal-withstand capabilities Engineering analyses to establish a reasonable qualified life and justified replacement schedule.

Calculations, assumptions, technical data and references were incorporated into tne respective equipment qualification documentation.

The results of these evaluations and analyses will 6e incorporated into the existing plant maintenance and surveillance program to ensure that equipment qualification is maintained. Based on these considerations, Niagara Hohawk fully complies with the aging and qualified life criteria presented in the Oivision of Operating Reactors guidelines.

When currently installed equipment (qualified to the Oivision of Operating Reactors guidelines) is replaced, the new equipment wi 11 oe qualified in accordance with the aging and qualified life criteria presented in 10CFR50.49,(e)(5); NUREG-0588, Category I, Section 4 and Regulatory Guide 1.89, Revision 1, Section 7 unless there are sound reasons to the contrary to preclude upgrading. For this equipment, the qualification test plans and test reports are evaluated to ensure that equipment is properly preconditioned (naturally or artificially) prior to testing and a reasonaole qualified (designated) life and component replacement interval is established. The results of the equipment qualification program will be incorporated into the existing plant maintenance and surveillance program to ensure that equipment qualification is maintained. Based on these considerations, Niagara Honawk fully complies with the aging and qualified life criteria for new equipment presented in 10CFR50.49, NUREG-0588 Category I and Regulatory Guide 1.89.

With respect to synergistic effects, Niagara Hohawk recognizes the limitations in the state-of-the-art; therefore, synergisms were not addressed unless known synergisms were identified and were considered to nave significant effect on the equipment's safety function. Based on these considerations, Niagara Honawk fully complies with the synergistic effects criteria presented in 10CFR50.49(e)(7), NUREG-0588, Category 1, Section 4(3) and Regulatory Guide 1.89, Revision 1, Section 7.

Finally, a program will be developed to be used in conjunction with tne established maintenance and surveillance program to identify significant age related degradation trends, characteristics and observations for equipment. Appropriate corrective actions will oe taken on a case-by-case basis.

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E. Maintenance and Surveillance The Division of Operating Reactors guidelines and 10CFR50.49 require that ongoing programs be implemented to establish and perform maintenance, surveillance and equipment (or component) replacement activities for safety-related electrical equipment to ensure that equipment qualification is maintained'n a continuing basis. The program must incorpor ate the established values of designated life for equipment considered to be susceptible to significant aging. Our methodology witn r espect to compliance with Nuclear Regulatory Commission criteria is summarized below.

Me are establishing a maintenance, surveillance and replacement program at Nine Mile Point Unit 1 for electrical equipment requiring environmental qualification. This program will supplement the existing plant maintenance and surveillance program by incorporating the relevant data of the qualification documentation packages so that qualification will be continuously maintained. Although tne program is not completed, many of the maintenance and replacement/refurbishment activities conducted during the 1984 refueling outage at Nine Mile Point Unit 1 is related to equipment environmental qualification.

As part of the equipment qualification program, we have developed environmental qualification documentation wnich provides an auditable basis to substantiate environmental qualification. This documentation contains vendor information, test reports and analyses whicn were conducted to establisn qualification for equipment types sucn as motor operated valves, E/P converters, etc. As part of the qualification review, the documentation identifies components and equipment susceptible to significant age related degradation and estaolisnes a reasonable designated life for replacement or refurbishment, maintenance and surveillance requirements. The equipment qualification packages are currently under review to extract and incorporate the salient information to be used at the plant for maintaining equipment qualification. Once this review and compilation of data is complete, an "equipment qualification maintenance, surveillance and replacement" procedure will be incorpor ated into the existing plant maintenance procedures. As an additional feature of the program, we will conduct periodic equipment qualification training seminars for plant and corporate office personnel to ensure that responsiole personnel are acquainted with the principal aspects of equipment qualification.

Based on the above considerations, we conclude that the activities conducted on environmental qualification and the current development of a maintenance and surveillance program for electrical equipment fully complies with the requirements of the Division of Operating Reactors Guidelines and 10CFR50.49.

Submer ence Section 4.3.4 of the Tecnnical Evaluation Report stated that Niagara Mohawk nas provided a satisfactory response to the Nuclear Regulatory Commission concern regarding submergence. The previous Nuclear Regulatory Commission Safety Evaluation Report. concern identified in Section 3.5 of the June 8, 1981 Safety Evaluation Report was:

"The Licensee has stated that submergence in the drywell is not a concern because of the low resistance flowpath to the torus and the large torus-free volume. It is not evident that the licensee considered the effects of all steamline breaks, both inside and outside containment (for example, a feedwater line break outside

'containment), and their relation to the submergence of Class 1E equipment. The licensee should address this concern."

In our September 8, 1981 response regarding submergence of Class 1E electrical equipment, we stated that submergence of Class lE equipment inside and outside containment is not a concern.

Further investigation, as determined by engineering calculations of the maximum submergence elevations, shows that equipment located inside and outside containment will not be submerged.

Based on these considerations, we consider that this issue is resolved.

~C Section 4.3.5 of the Technical Evaluation Report stated that Niagara Mohawk has provided a satisfactory response to the Nuclear Regulatory Commission concern regarding the effects of demineralized water spray.

The previous Nuclear Regulatory Commission Safety Evaluation Report concern, identified in Section 3.6 of the June 8, 1981 Safety Evaluation Report was:

"The Licensee has stated tnat the plant's primary containment does not have a chemical spray system. If, however, the plant design has a demineralized water spray, its effects should be examined and the Licensee should address this concern."

In our September 8, 1981 response regarding cnemical spray, we stated that various electrical components located inside containment were tested using various mixtures of chemical spray which, in fact, were a more severe and conservative simulation than the demineralized water spray. For example, boric acid and boron were used in the chemical spr ay/loss of coolant accident tests for Limitorque Valve Operators, Namco limit switches, Pyco thermocouples and General Electric caoles.

Further investigation based on engineering evaluations nas determined that deminerlized water spray will not affect performance of safety related electrical equipment located inside containment.

Based on these considerations, we consider that this issue is resolved.

Installed TMI Action Plan Items NUREG-0737 "Clarification of TMI Action Plan Requirements" established actions to be taken by licensees regarding TMI Lessons Learned Implementation. Nuclear Regulatory Commission Inspection and Enforcement Bulletin 79-018, Supplement 3, Item 2, requires environmental qualification of installed safety-related electrical, equipment located in narsn environments required for TMI Lessons Learned Implementation. Equipment items that fall within this category were identified and incorporated into the Nine Mile Point Equipment 1

Jr

(}ualification program. Accordingly, information on TNI lessons learned items was submitted on February 2, 1982 and Marcn 2, 1982. These items were subsequently addressed in the Nuclear Regulatory Commission Safety Evaluation Report/Technical Evaluation Report review (Technical Evaluation Report items 73 thru 99 inclusive). A resolution has been provided for qualification deficiencies identified in the Safety Evaluation Report/Technical Evaluation Report and is included in Section III of this report.

Based on these considerations, we conclude that installed THI action plan .items have been properly incorporated into the qualification program in accordance with Nuclear Regulatory Commission Inspection and Enforcement Bulletin 79-01B, Supplement 3, and Item 2, 10CFR50.49 qualification requirements.

Section II Equipment Items Considered Resolved by tne Tecnnical Evaluation Repor t

In Technical Evaluation Report-CS257-466, Franklin Research Center (FRC) assigned equipment items into one of eight qualification categories based upon review of documentation submitted by the licensee. Two of these categories provide a final resolution of the status of equipment items. Tnese categories are:

Nuclear Regulatory Commission Catetory Cate ory Oescri tion Equipment gualified Equipment Exempt From gualification Since the items in these categories have been resolved, they are'not addressed in this submittal. The equipment items in these categories are:

Equi ment /uglified NRC Cate or I.a)

FRC Technical Evaluation Report Plant Identification/

equal Oescri tion Item Number Manufacturer Model Number 61 OZ Gedney XL Electric terminal in steam tunnel Equi ment Exempt From ification NRC Cate or I I I. a)

FRC Tecnnical Evaluation Report Plant Identif ication/

Item Number Manufacturer Model Number Description 36 Rosemount 1151 OP LT 58-05, 58-06

Section III Resolution of Specific Equipment Environmental qualification Oeficiencies Identified in the Tecnnical Evaluation Report

Section 3.1 TRANSDUCER; E/P TER NO. NANUFACTURER/NOBEL NRC CATEGORY DEFICIENCY

2) Fisher Controls I.b Aging degradation evaluated adequately Hode) 546 Criteria regarding radiation satisfied RESOLUTION:

Sumaary: Oetai ls:

Inadequate materials list at time of NRC review. The qualification of these components was not established at the time of the Thermal and radiation analysis inadequate at time of NRC review.

Oua)ified by My)e Report 17655-EPT-1 (Reference ))5) and Fisher NRC review, and the components were to be rep)aced if qualification could not be established. The TER identified the following qualification Report NA-23 (Reference 34). deficiencies: A materials list for non-metallic items was not furnished by Equiiwent is qua)ified. the vendor to identify materials susceptible to radiation and thermal aging; the radiation evaluation for Buna "N" as the limiting material did not consider app)ication characteristics such as compression set, elongation etc.;

and the estimated life was based solely on Buna "N" without evaluation of the other organic materials. Subsequently, Wyle Report 17655-EPT-) compiled a materials list and performed a therma) and radiation aging evaluation for all susceptible non-metallic materials. The Wy)e Analysis considered compression set of the Buna "N" material as the most conservative case. The analysis established a 4.5 year replacement schedule for the Buna "N" o-rings and the upper and lower diaphragms. Based on this documentation, the equipment is qualified for a 40 year life with identified component maintenance and replacement schedule.

0 Section 3.2 TEHPERATURE ELEHENT TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF IC1ENCY 32 Hinco Nickel R-T-D I i.a Documented evidence of qualification adequate RESOLUTiON:

Suranary: Details:

Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time This equipment will be replaced by qualified PYCO Hodel 122-7026 of the NRC review, and an assessment was to be conducted to establish RTOs. qualification. Therefore, documented evidence of qualification was indicated Testing of PYCO RTO's has been completed by the manufacturer. by the TER as inadequate. This equipment will be replaced by qualified PYCO After replacement and plant specific analysis is performed using Hodel 122-7026 RTOs. The manufacturer has completed successful testing of PYCO test report, nunber 16436-82N (Reference )82), the equipment these RTOs and preliminary results indicated that the Nine Nile Point Unit is qualified. No. 1 environmental conditions are enveloped. After replacement and plant specific analysis is performed using PYCO test report number 16436-82N, this equipment is qualified.

Section 3.3 TEHPERATURE ELEHEHT TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY 22 Pall Trinity Cu/6 II.a Documented evidence of qualification adequate RESOLUTION:

Sunmary: Details:

Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time of the NRC This equipment will be replaced by qualified PYCO Hodel 102-3171-07 review, and an assessment was to be conducted to establish qualification.

thermocouples. Consequently, the TER indicated that the documented evidence of qualification Testing has been completed by the manufacturer. was inadequate. These components will be replaced by qua)ified PYCO After replacement and plant specific analysis is perforated using Hodel 102-3171-07 thermocouples. The manufacturer has completed PYCO test report nunber 16436-82N (Reference )82), the equipment successful testing of these thermocouples and preliminary results indicate is qualified. that the Nine Hile Point Unit No. 1 environmental conditions are enveloped.

After replacement and plant unique analysis is performed using PYCO test report number 16436-82N, the equipment is qualified.

l' Section 3.4 TEHPERATURE ELEHENT TER NO. HANUFACTURER/HOOEL PYCO 02-3171-08, Theruocouple ~

NRC CATEGORY Il.a OEF ICIENCY Documented evidence of qualification adequate

~

'I RESOLUTION: h

'uaeary:

Details:

Documented evidence of qualification not provided. Documented evidence of qua'lification was not provided at the time of the Oualif Ied by Patel Report PE I-TR-82-12-4 (Reference 52) and Wyle NRC review, and an assessment was to be conducted to establish Report 17655-TEL-1 (Reference 55). qualification. Consequently, the TER indicated that the documehted evidence of The qualified life of the equipment is 15 year s with gasket qualification was inadequate. Subsequently, Patel Report PEI-TR-82-)2-4 replacement at 3.7 year intervals. established qualification of this equipment for temperature, pressure, Equipment is qualified. humidity and radiation; and Wyle Report 17655-TEL-1 supplemented the Patel report by providing analysis which demonstrates that the thermocouple has a qualified life of 15 years with replacement of the gasket at 3.7 year intervals. Based on the documentation, the equipment is qualified.

Section 3.5 TEMPERATURE ELEMENT TER NO. MANUFACTURER/NOBEL NRC CATEGORY DEFICIENCY PYCO 102-3171-08, Theraocouple II.a Documented evidence of qualification adequate RESOLUTION:

Summary: Details:

Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time of the Qualified by Patel Report PEI-TR-82-12-4 (Reference 52) and Hyle NRC review, and an assessment was to be conducted to establish Report 17655-TEL-1 (Reference 55). qualification. Consequently, the TER indicated that the documented evidence of The qualified life is greater than 40 years. qualification was inadequate. Subsequently, Patel Report PEI-TR-82-12-4 Equipment is qualified. established qualification of the equipment for temperature, pressure, humidity and radiation; and liyle Report 17655-TEL-1 supplemented the Pate) report by providing analysis which demonstrates that the thermocouple has a qualified life of greater than 40 years. Based on the documentation, the equipment is qualified.

Section 3.6 TEHPERATURE SW ITCH TER NO. HANUFACTURER/HODEL NRC CATEGORY DEF I CIENCY 28 Fenwal/)7002-40 I I.a Adequate similarity between equipment and test specimen established Aging degradation evaluated adequately Oua))fied life or replacement schedule established (if required)

Peak pressure adequate Criteria regarding radiation satisfied RESOLUTION:

Suarnary: Deta i ls:

Documentation inadequate at time of NRC review. The TER evaluation concluded that the qua)1fication documentation for this 1)ua)if ication established by Kyle Report 17509-1 equipment was Inadequate with respect to similarity, aging, qualified life, A Myle Analysis will be conducted to justify minor pressure peak pressure and radiation. Subsequently, it has been determined that Fenwal deviation. Hodel 17002-40 temperature switches have been successfully tested by My)e When Kyle Report 17509-1 and Myle pressure analysis is received, Report 17509-1. The Myle test indicates that all environmental parameters the equipment will be qualified. applicable to Nine Hile Point Unit No. 1 have been enveloped except for pressure. A supplemental analysis will be provided by Kyle to justify the small deviation between the required 17.3 psig pressure and the 15.5 psig test condition for pressure. Wyle Report 17509-1 establishes a qualified life of 40 years for this equipment. We are currently in the process of obtaining Wyle Report 17509-1 and the supplemental Kyle analysis. Mhen this documentation is received, this equipment wi )1 be qualified.

Section 3.7 TEHPERATURE SWITCH TER NO. HANUFACTURER/HOOEL NRC CATEGORY DEFICIENCY N/A United Electric/ N/A N/A Hodel 829C RESOLUTION:

Su+nary: Octa i ls:

Iiuali fied for radiation by Hyle Report 17655-TSW-1 These temperature switches have been incorporated into the master list of (Reference )06). equipment based on a detailed systems review conducted subsequent to the NRC The qualified life is 40 years. review. The temperature switches are exposed to a high radiation environment Equipment is qualified. under accident conditions. Wyle Report 17655-TSM-1 establishes qualification for radiation for these devices. A 40 year qualified life has been established for these devices. This equipment is qualified.

Section 3.8 ELECTRICAL CONTROL" CABLE TER NO. HANUFACTURER/DODEC NRC CATEGORY OEF ICIENCY 95 & 96 Rockbestos/3C f12 AWG; Il.a Adequate similarity between equipment and test specimen established f18 AWG, 6 T/P; 7C f12 AWG; TSP 416 (XLPE Conductor Insulation, Neoprene Jacket)

RESOLurlON:

Suranary: Details Adequate siTnilarlty not established at time of NRC review. The TER evaluation concluded that adequate similarity was not estabalfshed i}ualification established by Patel Report PE I-TR-82-12-2 between the test specimen and the installed cable based on Rockbestos test (Reference 153) which includes Rockbestos test report. report "Oualification of Firewall III Class lE Electric Cables" dated July 7, The qualified life is 40 years. 1977. Subsequent to the NRC review, Patel Report PEI-TR-82-12-2 established Equipment is qualified. qualification for the cable by providing supplemental aging and similarity analysis in conjunction with Rockbestos test report "Oualification of Firewall III Class 1E Electric Cables" dated June 22, 1978. The qualification documentation established a qualified life for the cable greater than 40 years. Based on the documentation, the cable is qualified.

Section 3.9 ELECTR ICAL TERHINAL TER NO. HANUFACTURER/NOBEL NRC CATEGORY DEF ICIENCY 57, 58 Burndy/QAB and QASCB Il.c Adequate degradation evaluated adequately Qualified life or replacement schedule established (if required)

RESOLUTION:

Suaeary: Oetails:

Aging and qualified life inadequate at time of NRC review. The TER concluded that aging and qualified life were not adequately Qualified by Burndy Test Report T079-601A (Reference 13) and llyle addressed in Burndy Test Report T079-601A. Subsequently, lly1e Report Report 17655-TER-2 (Reference 50). 17655-TER-2 established that the QA-B series of electrical terminals contains Terminals are all metal construction. no insulation material and is constructed of high copper alloy metal. Because Equipment is qualified for 40 years. of the all metal construction, these terminals have no time related age d/grading material. Based on this documentation, the electrical terminals are qualified for 40 years.

Section 3.10 ELECTR ICAL TERMINAL TER NO. MANUFACTURER/MODEL NRC CATEGORY 56 Burndy I I.a Adequate similarity between equipment and test specimen established Type GE Ground Connector RESOLUTION:

Sugary: Details:

Adequate similarity between test specimen and equipnent n'ot The TER evaluation concluded that adequate similarity between the test established. specimens of Burndy Test Report TD79-601A and the installed electrical I)ualified by Burndy Test Report TD79-601A (Reference 13) and Wyle terminals was not estabalished. Subsequently, Wyle Report 17655-TER-1 Report 17655-TER-I (Reference 97). established that the type GZ ground connector installed at Nine Mile Point Terminals are all natal construction. Unit 1 and the tested type I)A-B series terminals differ only in mechanical Equipment is qualified for 40 years. configuration of materials. Both types are constructed of high copper alloy metal and contain no insulation material. Because of the all metal construction, these terminals have no time related age degrading material.

Based on this documentation, the Burndy type GL ground connectors are qualified for 40 years.

Section 3.)) CABLE TERHINAL TER NO. HANUFACTURER/HDDEL NRC CATEGORY OEF ICIENCY 60 AHP Pre-insulated, PIDG and Il.a Documented evidence of qualification adequate Plastigrip (Ring Tongue)

RESOLUTION:

Suaeary: Details:

The previous analysis report inadequate at the time of NRC review. The previous report did not provide adequate evaluation for materials.

Ring tongue terminals tested by NP Test Report, ) 10. 1)516 Therefore, the TER concluded that documented evidence of qualification was (Reference 76). inadequate. Subsequently, AHP Test Report, )10.11516 (for type PIDG and AHP Test Report, 110. 115)6 and Wyle Report, 17655-TER-4 PLASTI-GRIP ring tongue terminals) was obtained, which describes the test identified that the test was successful except for one configuration. results for 40-year normal life and loss of coolant accident/high energy line Engineering evaluation is being conducted to resolve the anomaly with break. AHP Test Report, 110.11516, and Wyle Assessment Report, 17655-TER-4 respect to our application. determined that the ring tongue terminals did satisfactorily perform If the anomaly is resolved, the equipment is therefore qualified. throughout the testing except for a configuration in which the insulation If the anomaly is not resolved, the equipment will be qualified by sleeves slipped off the wire barrels on some unenergized "PLASTI-GRIP" an additional engineering analysis or equipment modification/ specimens mounted vertically with the ring tongue end up during loss of replacement. coolant accident/high energy line break simulation. An engineering evaluation When the analysis is completed, this equipment is qualified. is being conducted to resolve the anomaly with respect to our application to determine if "PLASTI-GRIP" type terminals are used and to determine the installation configuration. If the anomaly is not resolved by identification of type and installation configuration, an additional detailed engineering analysis wi)1 be conducted to assess the significance of the sliding of the sleeves with respect to qualification or the equipment will be modified/replaced. Upon completion of this analyses, this equipment is qualified.

Section 3.12 CABLE SPLICE TER NO. MANUFACTURER/MODEL NRC CATEGORY DEP ICIENCY 59 AMP Pre-insulated, Butt Splices Il.a Documented evidence of qualification adequate RESOLUTION:

Suenary: Details:

The previous analysis report was inadequate at the time of NRC review. The previous report did not provide adequate evaluation for materials.

Ring tongue terminals tested by AMP Test Report 110.11516 Therefore, the TER concluded that documented evidence of qualification was (Reference 76). inadequate. Subsequently, AMP Test Report 110.11516 (for type PIDG and Hyle Report, 17655-TER-3, does not identify acceptable PLASTI-GRIP terminals) was obtained, which describes the test results for similarity between the tested terminals and splices. 40-year normal life and loss of coolant accident/high energy line break. Wyle An engineering evaluation will be conducted to establish acceptable Report 17655-TER-3 does not identify acceptable similarity between the tested similarity. terminals and splices. Therefore, an engineering evaluation will be conducted When the analysis is completed, this equipment will be qualified. to establish acceptable similarity. When the analysis is completed, the equipment will be qualified.

Section 3.13 MOTORS TER NO. MANUFACTURER/MODEL NRC CATEGORY DEFICIENCY 42 GE/ I I.a Adequate similarity between equipment and test specimen established Type 5K8288 37C7 I)uallfied life or replacement schedule established (if required)

Peak pressure adequate RESOLUTION:

Surrmar y: Details:

GE Analysis Report (G-EN-0-164) .inadequate with respect to simi- The NRC concluded that adequate similarity, bearing and lubricant larity, replacement schedules and pressure effects. replacement schedules and effects of peak accident pressure were not Oualification addressed by Wyle Report 17655-MOT-l. 1 adequately addressed by G.E. Analysis Report G-EN-0-164. Subsequently, Wyle (Reference 112), G.E. Analysis Report G-EN-0-164 (Reference 47) Report 17655-MOT-1.1 and GE BWR surrmary r eport I)SR-Ill-A-04 were and GE BWR Surrmary Reprt I)SR-Ill-A-04 (Reference 111). provided. This documentation, in conjunction with GE Analysis Report Accident environment is not severe. G-EN-0-164, resolves the deficiencies with respect to similarity of the type The existing qualification documentation does not adequately address 5K motors. The environment that these motors are exposed to is relatively the effects of pressure, aging and identify required lubricants. nonsevere (126F/1 psig/100 percent RN )xlS Rads). An engineering evaluation An additional analysis will be conducted to provide full wi 11 be conducted to adequately assess the qualified life, effects of peak qualification. pressure on the installed motors, identify the required bearings and bearing When. the additional analysis is completed, the equipment is lubricants and requisite replacement schedule, and provide qualification for qualified. the motor pigtail lead cable and splices. Upon completion of this analysis, this equipment is qualified.

Section 3.14 TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF I CIENCY 43, 44 GE Ii.a Documented evidence of qualification adequate Type 5K6328 XC136A, 5K6336 XC166A RESOLUTION:

Surnnary: Details:

Qualification addressed by Wyle Report 17655-HOT-1.1 The NRC concluded that the documented evidence of qualification was (Reference 112), GE Analysis Report G-EN-0-164 (Reference 47) and not provided because GE Analysis Report G-EN-0-164 was not referenced as the GE BWR summary report QSR-lll-A-04 (Reference 111). evidence of qualification. Subsequently, Hyle Report 17655-HOT-l.l Acc.ident environment is not severe. and GE BWR sumnary report QSR-111-A-04 were provided. This documentation, in The existing qualification documentation. does not adequately address conjunction with GE Analysis Report G-EN-0-164, resolves the deficiencies with the effects of pressure, aging and identify required lubricants. respect to similarity of the type 5K motors. The environment that these An additional analysis will be conducted to provide full motors are exposed to is not severe (110F/I psig/100 percent RH Ix106 Rads).

qualification. An engineering evaluation will be conducted to adequately assess the qualified Mhen the additional analysis is completed, the equipment Is life, effects of peak pressure"on the installed motors, identify the required quali f i ed. bearings and bearing lubricants and requisite replacement schedule and provide qualification for the motor pigtail lead cable and splices. Upon completion of this analysis, this equipment is qualified.

Section 3.15 MOTORS TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY N/A GE N/A N/A Types 5K184AL 218, 5K8143 16A73 RESOLUTION:

Suranary: Detai ls:

Items added to master list after NRC review. These motors have been incorporated into the master list of equipment Qualification addressed by Wyle Report 17655-HOT-1.1 based on a detailed systems review conducted subsequent to the NRC review.

(Reference 112), GE Analysis Report G-EN-0-164 (Reference 47) These motors drive the reactor building emergency exhaust fans and the control and GE BWR Suaeary Report QSR-Ill-A-04 (Reference 111). rod drive pumps. Wyle Report 17655-HOT-l.l, GE Analysis Report Accident environaent is not severe. G-EN-0-164 and GE BWR suamary report QSR-ill-A-04 have been provided.

Existing documentation does not adequately address effects of The environment that these motors are exposed to is relatively nonsevere peak pressure, aging, bearings, lubricants, splices and similarity (133F/1 psig/100 percent RN/2.8 x 106 Rads). An engineering analysis will be between random wound motor and form wound motor. conducted to adequately assess aging, effects of peak pressure, identification An additional analysis will be conducted to provide full of bearings and lubricants and requisite replacement intervals, pigtail lead qualification. cable and splices and the correlation between the form wound motor insulation When the additional analysis is completed, the equilxnent is system and the installed random wound motor insulation system. " Upon completion qualif ied. of this analysis, this equipment is qualified.

Section 3.16 MOTORS TER NO. MANUFACTURER/NODEL NRC CATEGORY OEF ICIENCY GE Il.a Documented evidence of qualification adequate Type SK445AK-249A RESOLUTION:

Surnnary: Oetails:

O ocumented evidence of qualification not provided at the time of ~ Documented evidence of qualification was not provided at the time of the NRC NRC review. review, and an assessment was to be conducted to establish qualification.

ualification addressed by Wyle Report 17655-MOT-l. I Consequently, the TER indicated that the documented evidence of qualification Reference 112), GE Analysis Report, G-EN-0-164 (Reference 47), was inadequate. These motors drive reactor building closed loop c'ooling pumps.

and GE BWR Suranary Report I)SR-Ill-A-04 (Reference 111). Wyle Report 17655-MOT-I.I, GE Analysis Report G-EN-0-164 and GE BWR Existing documentation does not adequately address effects of peak suaeary report I)SR-III-A-04 have been provided. Additional engineering pressure, aging, bearings, bearing lubricants, splices and analysis will be conducted to adequately assess aging, effects of peak pres-similarity between random wound motor and form wound motor. sure, identification of bearings and lubricants and requisite replacement in-An additional engineering analysis wi)1 be conducted to provide tervals, pigtail lead cable and splices and the correlation between the form full qualification. wound motor insulation system and the installed random wound motor insulation When the additional analysis resolves the above deficiencies, the system. Upon completion of this analysis, the equipment is qualified.

equipment is qualified.

Section 3.17 MOTORS TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF ICIENCY 91 *Howard 48-59214-005 II.a Documented evidence of qualification adequate RESOLUTION:

Sumnary: Details:

Documented evidence of qualification not provided at the time of Documented evidence of qualification was not provided at the time of the NRC review. NRC review and an assessment was to be conducted to establish Iiuaiification by Wyle Report 17655-HOT-2. 1 (Reference qualification. Consequently, the TER indicated that the documented evidence 114). bf qualification was inadequate. Subsequently, Wyle Report, The equipment is qualified for 40-year normal life plus the 100-day 17655-MOT-2.1, demonstrates that the equipment has a accident requirement. qualified life of 40 years plus the 100-day post-accident sampling requirement for the radiation only environment. Based on the documentation, the equipment is qualified.

a The NRC TER/SER identified this motor as being manufactured by Franklin; however, field verification has indicated that this motor was manufactured by Howard.

Section 3.18 SOLENOID ACTUATOR FOR RELIEF VALVE TER NO. HANUFACTURER/HODEL NRC CATEGORY DEF ICIEHCY 14 GE Solenoid I.b Documented evidence of qualification adequate CR9503-213C and UHINAX Switch WHB-5 RESOLUTION:

Sumnary: Details: ~

Iiualification documentation was inadequate during the HRC No qualification documentation was provided at the time of the NRC review; we evaluation. stated that the equipment would be replaced if qualification could not be GE Test Report PEP 42963 (Reference 118) shows that similar established. Subsequent to the TER , we obtained GE Test Report PEP 42963 equipment was tested to loss of coolant accident conditions. which described simulated loss of coolant accident steam tests performed on Wyle Report 17655-SVRV-1. 1 (Reference 119) describes thermal, Dresser Electromatic actuators. These tests were considered satisfactory in qualified life and radiation analyses of component materials. NRC/SER for other BWR plants except for thermal aging, qualified life and rad-Some additional analysis will be conducted to show full qualifi- iation testing. Wyle Report 17655-SVRV-l.l addressed these deficiencies and cation. determined on the basis of a materials evaluation and testing reported in PEP Documentation will be obtained to establish similarity between 42963 that the actuator is qualified for 40 years plus the required tested and installed equipment. operating time under accident conditions. An engineering evaluation will be conducted to access equipment materials and establish the similarity of installed to tested equipment, for which additional documentation is being

~

obtained. This equipment will be qualified when the additional materials evaluation and similarity documentation is obtained.

Section 3.19 INSULATING TAPE {ELECTRICAL)

TER NO. MANUFACTURER/HODEL NRC CATEGORY OEF I CI ENCY 63 Scotch Brand 3M I I.a Documented evidence of qualification adequate 88 RESOLUTION:

Sumnary: Details:

Documented evidence of qualification not provided at the time Documented evidence of qualification was not provided at the time of the NRC of NRC review. review. Iluallfication will be conducted by a detailed engineering Iluali fication will be conducted by a detailed engineering analysis analysis and by obtaining applicable test reports for the equipment.

and by obtaining applicable test reports. Upon completion of this analysis, this equipment is qualified.

When the analysis is completed, this equipment is qualified.

Section 3.20 TERHINAL BLOCK TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY 53 General Electric I I. a Documented evidence of qualification adequate EB-5, EB-25 Adequate similarity between equipment and test specimen established RESOLUTION:

Sumnary: Details:

The TER concluded that NUS Report 1961-6080-001-Rl (Reference 18) The NRC concluded in the TER that the previous NUS Report 1961-6080-001-Rl did not provide documented evidence of qualification. (Reference 18) did not establish adequate similarity and documented evidence Hyle Report 17655-TB-1.1 (Reference 57) addresses of qualification. Subsequently, Hyle Report 17655-TB-l.l was qualification. However, similarity has not been adequately obtained which addressed qualification. An additional detailed engineering substantiated. analysis will be conducted to assess equipment materials and establish the A detailed engineering analysis wi ll be conducted to establish similarity of installed to tested equipment for which additional documen-qualification. tation is being obtained to qualify the equipment. Upon completion of this Hhen the analysis is completed, this equipment is qualified. analysis, this equipment is qualified.

K, Section 3.21 CABLE TER NO. HANUFACTURER/MODEL NRC CATEGORY DEFICIENCY 65 General Electric/ I I.a Adequate similarity between equipment and test specimen established Vulkene RESOLUTION:

Sunanary: Detai ls:

Similarity between tested and installed cable was not established. Sufficient information to identify the similarity of cable installed with

• Myle Report 17655-CBL-I (Reference 157), NUS Analysis respect to that tested in FIRL Test Report No. F-C4497-2 was not provided at NUS-1961-G080-002 (Reference 10), and FIRL Test Report F-C4497-2 the time of the NRC review. Subsequently, Myle performed preliminary cable (Reference 11) address qualification. qualification assessments described in Myle Report 17655-CBL-1. We have iden-We have identified installed cable and is attempting to establish tified six safety related Vulkene-insulated cables installed in our plant by similarity to cable tested in FIRL Report F-C4497-2. GE cable specification SI index number (58073, 57275, 58745, 58743, 53032, A qualified life evaluation will be determined based on thermal 58146). A qualified life evaluation will be determined based on thermal aging aging in FIRL Report F-C4497-2. conducted in FIRL Report F-C4497-2. Me are also developing documentation The equi@rent will be qua)ified if similarity is established. which establishes the SI indices of cable tests reported in FIRL Test Report Testing will be conducted if adequate similarity cannot be No. F-C4497-2 in order to establish similarity. If similarity is established, established. the cable will be qualified. If similarity cannot be established, then qualification testing wi 11 be conducted.

4, Section 3.22 CABLE TER NO. HANUFACTURER/NOBEL NRC CATEGORY OEF ICIENCY 94 Rockbestos/RSS 6104 IV Documentation not made available RESOLUTION:

Surnnary: Detai ls:

No qualification docmnentation was provided for the TER/SER review. Rockbestos test report provides evidence of qualification, however, this This equipment is being replaced with qualified Rockbestos test report was not provided to NRC for review. The Rockbestos coax cable RSS 6104-1081. is used with General Atomics radiation detectors. The TER/SER discussed certain anomalies involving Rockbestos coax which were observed during qualification tests of the radiation detectors reported in a General Atomics Test Report (Reference 54). Based on vendor contact, we have determined that the Rockbestos coax installed in our plant is similar to the cable which experienced testing anomalies. These anomalies were applicable to the loss of coolant accident/high energy line break test, not radiation which is the only harsh environment for these detectors/cable. We will replace with Rockbestos RSS 6104-1081. Following the installation of the replacement cable, this Rockbestos cable item will be deleted from the master list.

Section 3.23 INSULATING TAPE FOR ~ 5KV TERHINAL INSULATION TER NO. NARUFACTURER/HODEL NRC CATEGORY DEFICIENCY 70 Kerite Ii.a Documented evidence of qual ification adequate Splicing Compound Tape RESOLUTION:

Suaeary: Details:

Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time of the NRC Hyle Report 17655-TPE-4.1 (Reference IOO) addresses review. Subsequently, Hyle Report 17655-TPE-4.1 addressed qual ification quali f ication. of the Kerite insulating tape. However, the Hyle Report addresses Isomedix Test Report I-R975-Ol will be used to establish aging qualification; however, aging qualification has not been established at this qualification by analysis. time. Isomedix Test Report, I-R975-01 has been obtained and is being used Hhen the additional analysis is completed, the equipment is for establishing aging evaluation. Upon completion of this analysis, this qual I f led. equipment is qualified.

Section 3.24 CABLE TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF ICIENCY 71 Kerite/I)uadr aplex I I. a Documented evidence of qualification adequate SkV Power Cable RESOLUTION:

Surnnary: Details:

Documentation was not provided at the time of NRC review. No qualification reports were available at the time of the NRC if liua) ication could not be established by ongoing assessments. review. However, we cennitted to performing an ongoing qualification Hyle', Test Report 47)76-) (Reference )8)) has been received. assessment of this cable. This assessment included a search for Upon review of this test report, this cable will be qualified. possible test reports and an attempt by Kerite to determine if our cables were similar to other qua)ified Kerite power cables. We were unable to obtain adequate documentation, and consequently contracted Wyle Labs to perform qualification testing of the cable. The testing program has. been completed. Hyle Test Report 47176-1 demonstrated that the cable has successfully withstood exposure to our steam tunnel conditions. Upon completion of review of the report and other pertinent vendor data, this cable will be qualified.

Section 3.25 INSTRUNENT CABLE TER NO. NANUFACTURER/NOBEL NRC CATEGORY OEF ICIENCY 54 Rayc hem/RG59B/U Il.a Adequate similarity between equipment and test specimen established RESOLUTION:

Suranary: Details:

This coaxial cable is not used in a safety related application Similarity between the installed cable and test documentation could not be requiring environmental qualification. established for the NRC review. Subsequent to the TER, we have This cable has been deleted from the list because it is outside determined that the subject cable is not used for any safety related appli-of the scope of 10CFR50.49.. cations requiring environmental qualification and can therefore be deleted from the list of equipment requiring qualification because it is outside of the scope of 10CFR50.49.

Section 3.26 PENETRATION CONNECTOR ASSEHBLIES TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF ICIENCY 52 0. G. O'rien II .a Aging degr adation evaluated adequately 5, 19 and 28 Pin f16 and 4 Pin f8 Oualified life or replacement schedule established (if required)

Criteria regarding spray satisfied Criteria regarding radiation satisifed RESOLUTION:

Sumnary: Details:

TER deficiencies included aging, replacement interval, FIRL Test Report F-C4879-I and NUS Analysis 1961-005-001 were reviewed radiation and spray effects. by the NRC in the TER. TER-identified deficiencies included: inadequate FIRL Test Report F-C4879-1 (Reference 8), Hyle Report 17655-PC-1. 1 evaluation of thermal aging degradation and replacement interval; no testing or (Reference 160) and Patel Report PE I-TR-82-12-101 (Reference 70) evaluation of spray effects; and test irradiation exposure (26 Hrad) less than support qualification of this equipment. the plant drywell TIO envelope conditions (50 Mrad). Subsequent evaluations Additional analyses and documentation are required to demonstrate reported in Hyle Report 17655-PC-l.l and Patel Report PEI-TR-82-12-101 full qualification. concluded that connector thermal and radiation materials degradation due to plant service conditions were insufficient to adversely affect qualification.

The evaluations are currently being revised to include additional materials data and Incorporate vendor test information on spray effects, thereby resolving all TER concerns. The Wyle and Patel technical evaluations, in conjunction with loss of coolant accident tests described in FIRL Test Report F-C4879-1, will demonstrate full qualification of the equipment. When the additional analyses is completed, this equipment is qualified.

Section 3.27 CONNECTOR TER NO. HANUFACTURER/NOBEL NRC CATEGORY OEF ICIENCY 92, 93 0. G. O'rien Il.a Documented evidence of qualification adequate C10C0001G14, C10C1001G21 RESOLUTION:

Su+nary: Details:

I)ualification references were not provided for the NRC No qualification references were provided for the NRC review. Sub-review. sequent to the TER, Wyle Report 17655-PC-l.l demonstrated, through a com-Wyle Report 17655-PC-l.l (Reference 160), FIRL Test Report bination of testing and analysis, that these connectors were qualified. This F-C4879-1 (Reference 8) and Patel Technical Report PEI-TR-82-)2-101 report documents similarity of these connectors to the D. G. O'Brien connector (Reference 70) establish qualification. assemblies tested in FIRL Test Report F-C4879-1 (see Section 3.26). Haterial This equipment is qualif led. evaluations for thermal and radiation degradation were performed in Wyle Report 17655-PC-I.I and Patel Report PEI-TR-82-12-101. Although these reports are currently being revised to incorporate additional materials data, the revisions are not applicable to this equipment. This equipment is qualified.

~

Section 3.28 FILLER FOR" 5KV TERHINAL TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY 64 GE/227 11.a Documented evidence of qualification adequate RESOLUT ION:

Su+nary: Details:

Docenented evidence of qualification not provided at th'e time of Documented evidence of qualification was not provided at the time of the NRC review. NRC review. Qualification wi 11 be conducted by a detailed engineering Qualification will be conducted by a detailed engineering analysis analysis and by obtaining applicable test reports for the equipment.

and by obtaining applicable test reports.

Section 3.29 FILLER FOR 5KV TERMINAL TER NO. MANUFACTURERjNOOEL NRC CATEGORY DEFICIENCY 55 J-M Duxseal Ii.a Documented evidence of qualification adequate RESOLUTION:

Suaeary: Details:

Documented evidence of qualification not provided at the time of Documented evidence of qualification was not provided at the time of the NRC review. NRC review. Iiualification will be conducted by a detailed engineering Iiuallfication will be conducted by a detailed engineering analysis analysis and by obtaining applicable test reports for the equipment.

and by obtaining applicable test reports.

Section 3.30 INSULATING TAPE {ELECTRICAL TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY 67 GE/8380 Il.a Documented evidence of qualification adequate RESOLUTION:

Surunary: Details:

Documented evidence of qualification not provided at the time of Documented evidence of qualification was not provided at the time of the NRC review. NRC review. Ijualification will be conducted by a detailed engineering Iiualification will be conducted by a detailed engineering analysis analysis and by obtaining applicable test reports for the equipment.

and by obtaining applicable test reports.

1 Section 3.31 SPLICE KIT VARNISH TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF I CI EHCY 66 'Westinghouse 11. a Documented evidence of qualification adequate 3109'Black Insulating Varnish RESOLUTION:

Sumaary: Details:

Documented evidence of qualification not provided at the time of Documented evidence of qualification was not provided at the time of the HRC review. NRC review. Ilualification will be conducted by a detailed engineering llualification will be conducted by a detailed engineering analysis analysis and by obtaining applicable test reports for the equipment.

and by obtaining applicable tes't reports.

The NRC TER/SER identified this varnish as being GE 1309. Review of original plant specifications indicated that Westinghouse 3109 was used.

Section 3.32 CABLE TRR NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY N/A ITT N/A N/A RG-59/U Coaxial Cable RESOLUTION:

Surrmar y: Detai ls:

This is a replacement item for the Rockbestos RSS 6104 cable which The ITT coax cable is a replacement item for the Rockbestos RSS 6'i04 cable will be deleted. (See Section 3.22.) which will be deleted (see Section 3.22). This equipment is qualified by FIRL FIRL Test Report F-A5550-8 establishes qualification Test Report F-A5550-8. The 40-year qualified life for our normal (Reference 154). temperature and maximum conductor operating temperature wi 11 be established by 4 A 40-year qualified life will be established for our specific an additional analysis. This cable is qualified.

service conditions.

The equipment is qualified.

Section 3.33 SEALANT FOR 5KV TERHINAL INSULATION TER tS. MANUFACTURER/HODEL NRC CATEGORY DEFICIENCY Kerite Cement II.a Documented evidence of qualification adequate RESOLUTION:

Surnnary: Details Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time of the NRC Myle Report 17655-TPE-4.1 (Reference 100) addresses review. Subsequently, Myle Report 17655-TPE-4.1 was obtained which qualification. addresses qualification of the Kerite sealant type cement. The A plant specific analysis will be performed using Isenedix Myle Report addresses qualification; however, aging qualification Test Report, I-R975-01 (Reference 73) to establish qualification. has not been established at this time. Therefore, a plant specific analysis Supplemental analysis will be provided as necessary. will be performed using the Isomedix Test Report, I-R975-01. Supplemental Upon completion of this analysis, this equipment will be qualified. analysis will be provided as necessary. Upon completion of this analysis, this equipment will be qualified.

Section 3.34 UNDERCOAT FOR 5KV TERHINAL FILLER TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF ICIENCY 69 Kerite Friction Tape II .a Documented evidence of qualification adequate RESOLUTION:

Sumaary: Details:

Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time of the NRC Wyle Report 17655-TPE-4.1 (Reference 100) addresses review. Subsequently, Wyle Report 17655-TPE-'4.1 was obtained which qualification. addresses qualification of the Kerite friction tape. The Wyle Report A plant specific analysis will be performed using Isomedix Test addresses qualification; however, aging qualification has not been Report I-R975-Ol (Reference 73) to establish qualification. established at this time. Therefore, a plant specific analysis will Supplemental analysis will be provided as necessary. be performed using the Isomedix Test Report, I-R975-01. Supplemental analysis When the additional analysis is completed, the equipment is qualified. will be provided as necessary. Upon completion of this analysis, this equipment will be qualified.

Section 3.35 VACUUM SNITCH TER ND. MANUFACTURER/MODEL NRC CATEGORY DEFICIENCY 37 Hercoid I).a Documented evidence of qualification adequate CP4122 RESOLUTION:

Sugary: Details:

Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time of the NRC Oua) if ication documentation could not be found. review; therefore, a qualification assessment was to be conducted.

This equipaent will be replaced with qualified SOR mode) 52TA-8116- Subsequently, an engineering evaluation and literature search revealed that the NX-ClA vacua switches. existing equipment contained materials susceptible to radiation. This equip-The SOR vacuum switches are qualified by Acton Test Report ment also has mercury switches which are susceptible to transient phenomena 17344-82N-C, Revision 1 (Reference )56). caused by temperature var iation and radiation exposure. In addition, qualifi-The replacement equipment is qualified. cation documentation could not be found. Therefore, the existing mercoid vacuum switches will be replaced with environmentally qualified vacuum switches manufactured by SOR, Inc., Hodel 52TA-8116-NX-C)A. The SOR vacuum switches are qualified by Acton Test Report )7344-82N-C, Revision 1. To obtain similarity between the test specimen and the model 52 TA-8116-NX-C)A vacuum switch, we are procuring the SOR, Inc. 52TA-Bl)6-NX-C)A switches to be fabricated with the same materials (o-rings, gaskets, insulators, etc.) which were installed in the model )2TA-84-NX-C)A-JDTTX6 switches tested by Acton and described in Report )7344-82-NC. These replacement vacuum switches are qua)if ied.

Section 3.36 PRESSURE: SWITCliES TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF I Cl ENCY N/A Hercoid N/A N/A DA23-156 and XONI-43 RESOLUTION:

Suamary: Details:

This equipment was added to the master list based on systems review. This equipment was added to the master list subsequent to the NRC re-Ilualification documentation could not be found. view based on detailed systems review. An engineering evaluation and This equipment will be replaced with qualified SOR model literature search revealed that the existing equipment contained materials 12TA-BB4-NX-ClA-TTJJX6 pressure switches. susceptible to radiation. This equipment also has mercury switches which are The SOR pressure switches are qualified by Acton Test Report susceptible to transient phenomena caused by temperature variation and 18878-84N-2 (Reference 155). radiation exposure. In addition, qualification documentation could not be A plant specific aging evaluation will be conducted to extend the found. The existing mercoid pressure switches will be replaced with qualified qualified life. pressure switches manufactured by SOR Inc., model 12TA-BB4-NX-CIA-TTJJX6. The The replacement equipment Is qualified. SOR pressure switches are qualified by Acton Test Report 18878-84N-2. A plant specific aging evaluation will be conducted to extend the present 20 year qualified life. The replacement pressure switches are qualified.

Section 3.37 PRESSURE SWITCH TER NO. MANUFACTURE~R MODEL NRC CATEGORY DEFICIENCY Static-0-Ring Il.a Documented evidence of qualification adequate.

5NNKK351CIA RESOLUTION:

Suamary: Details:

Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time of the NRC Iiualification is established by Wyle Report 17655-PSW-2 review; therefore, a qualification assessment was to be conducted.

(Reference 121) and Patel Report PEI-TR-82-12-18 (Reference 133). Subsequently, Wyle Report 17655-PSW-2 and Patel Report PEI-TR The only harsh environment is radiation. 12-18 were obtained which establishes qualification for this equipment. This This equipaent is qualified for a 40 year life. pressure switch is used in the reactor vessel post-accident sampling system and is required to remain functional after a loss of coolant accident. The pressure switch is located in the reactor building at the 237 foot elevation and is required to be qualified for a radiation environment. This equipment is qualified for a 40 year life.

Section 3.38 PRESSURE SW I TCII TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY N/A N/A Barksdale 825-H12SS t(/A

~

RESOLUTION:

Details:

'u+nary:

This equip~rent was added to the master list subsequent to the NRC This equipment was added to the master list subsequent to the NRC review. review based on our detailed systems review. The equipment was furnished Wyle Report 17655-PSW-6 (Reference 139) addresses quali- by Dresser Industries and is located in the reactor building east or west in-fication; however, qualification is not established. strument rooms. This equipment actuates the automatic depressurization system The equipment is in an enclosure and consists of a terminal block, electromatic pressure relief valves for reactor vessel depressurization. Based wiring, a relay and a pressure. switch. on actual field verification, we have determined that this equipment consists A detailed engineering analysis will be conducted to establish of an enclosure which contains the following control circuitry: terminal qualification after which this equipment will be qualified. block, wiring, Barksdale model B25-iil2SS pressure switch and a relay. Wyle Report 17655-PSW-6 addresses qualification; however qualification has not been fully established for the entire control circuitry at this time. A detailed

'lenglneering analysis will be conducted to establish qualification for all components within the enclosure based on the accident environmental conditions and system functional requirements. Upon completion of the engineering analysis, this equipment will be qualified.

0

Section 3.39 CABLE TER I@. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY 97 *Kerite IV Documentation not made available FR2/FR Thermocouple Cable RESOLUTION:

Suranary: Details:

Documentation was unavailable for the NRC review. At the time of the TER review, we presented informat1on which showed there The cable manufacturer has prepared a plant-specific report were no qualification deficiencies for this cable; however, test reports were (Reference 79) which documents qualification. not available for NRC's review. Kerite Company has prepared a qualification This cable is qualified for a 40 year life. report (Reference 79) for the subject cable entitled "Nine Hile Point Unit 1-NHPC - Qualification Documentation for FRII/FR Signal and Instrumentation Cables," dated October 17, 1980. This report documents loss of coolant accident testing which exceeds our accident conditions and includes analyses showing cable qualified life exceeds 40 years. We have evaluated this qualification report and concludes that the subject cable is qualified. This ~

cable is qualified for a 40 year life.

• The NRC TER/SER identified this as 16 AWG. Review of plant specific information indicates FR2/FR was used.

Section 3.40 RADIATION DETECTION TER ti0. HANUFACTURER/HODEL NRG CATEGORY DEFICIENCY General Atomics I I.a Adequate similarity between equipment and test specimen established RD-23 Program established to identify aging degradation Criteria regarding radiation satisifed RESOLUTION:

Surreary: Details:

NRC deficiencies included radiation and electrical inter- The NRC qualification deficiencies included radiation qualification and face qualification. identification and qualification of the radiation detector electrical inter-General Atomic Test Report E-254-960 (Reference 54) and Wy1e faces. Hyle Report 17655-RAD-2 has analyzed radiation effects on the Report 17655-RAD-2 (Reference 93) support qualification. detector materials and concluded that irradiation of the detector prior to This equipment is qualified. loss of coolant accident/high energy line break testing described in General Atomic Test Report E-254-960 would not have affected detector performance.

The detectors and their associated equipment (i.e., cable and electrical interface connectors) are subject to a radiation environment only. Electrical interface anomalies which occurred during General Atomic's test program were reviewed, and it was determined that these anomalies were applicable to the loss of coolant accident/high energy line break test only, not radiation. This equipment is qualified.

Section 3.41 RADIA1'ION DETECTOR TER NO. HANUFACTURER/NOBEL NRC CATEGORY DEFICIENCY N/A GE N/A N/A 194 X927 RESOLUTION:

SurMary: Details:

This equipment was added into the master list based on a detailed This equipment has been incorporated into the master list of equipment based on system review conducted subsequent to the NRC review. a detailed systems review conducted subsequent to the NRC review. These Radiation is the only harsh environm'nt. detectors (RE-RN04A-5, 048-5, 05C-10) are located in the Reactor Building.

Qualified by Hyle Report, 17655-RAO-1 (Reference 85) The only harsh environmental parameter for which these actuators should be and GE BHR EQ Suranary Report, QSR-015-A-01 (Reference 77). qualified is radiation. The equipment is qualified by Hyle Report 17655-RAD-I Equipment is qualif ied. and GE BHR EQ Sugary Report QSR-015-A-Ol. The equipment is qualified.

Section 3.42 POSITION SHITCH TER NO. MANUFACTURER/MODEL NRC CATEGORY OEF I CIENCY 45, 46 Namco I.b Adequate similarity between equipment and test specimen established 02400X 0240OXR RESOLUTION:

Surnnar y: Oetai ls:

NUS qualification assessment report 1961-N007-001 was deficient The NRC review concluded that NUS qualification assessment report 1961-with respect to similarity. N007-001 was deficient with respect to establishment of adequate simi)arity This equipment will be replaced with environmentally qualified between the loss of coolant accident test specimen, Namco model SL3 and the Namco model EA)80 position switches. installed equipment, Namco model 02400X. The Namco model 02400X position Hyle Report 17655-LSW-).l Revision A (Reference 56), Pate) Report switch will be replaced with environmentally qualified Namco model EA)80 PEI-TR-82-12-9 and Namco Test Report l}TR-)05 (Reference 5)) position switches. Hyle report )7655-LSH-).1 revision A, Pate) establish qua)ification. report PEI-TR-82-12-9 and Namco Test Report OTR-)05 establish Namco qualified EC210 series receptacle/plug connectors or Conax environmental qualification. In order to assure proper sealing of the switch, electrical seal assemblies will be used. environmentally qualified Namco series EC210 receptacle/plug connectors or A plant specific qualified life evaluation will be conducted. Conax electrical seal assemblies will be used. The vendor has established a The vendor test report for the EC210 series connectors will be qualified life of 20 years for the Namco switches and 40 years for the obtained, if required. Conax Report IPS-1079 establishes receptacle/plug connectors. A qualified life for the Conax seal assemblies qualification for the Conax seal assemblies. must be established. A plant specific engineering evaluation will be con-ducted to establish a qualified life for this equipment. The vendors test report substantiating qualification for the EC210 receptacle/plug con-nectors will be obtained if this seal is used. If the Conax seal assembly is used, Conax report IPS-1079 establishes qualification for the seal assemblies. Upon completion of the plant specific qualified life evaluation and receipt of the series EC210 connector test report (if required), this replacement equipment will be fully qualified.

Section 3.43 POSITION SWITCH TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF I CIENC Y N/A Namco N/A N/A 02400X RESOLUTION:

Sunmary: Details:

This equlixnent was added to the master list subsequent to the IlRC This equipment was added to the master list of equipment subsequent to the NRC review. Ieview based on our detailed systems review. The Namco Model 02400X This equipment will be replaced with environmentally qualified position switch will be replaced with environmentally qualified Namco Model Namco Hodel EA180 position switches. EA180 position switches. Wyle Report 17655-LSW-l.l, revision A, Patel Wyle Report 17655-LSW-1.1 Revision A (Reference 56), Patel Report qualification assessment report PEI-TR-82-12-9, and Namco Test Report PE I-TR-82-12-9 and Namco Test Report INTR-105 (Reference 5l) 11TR-I05 establish environmental qualification. In order to assure if establ ish qual ication.

Namco qualified EC210 series receptacle/plug connectors or Conax proper sealing of the switch, environmentally qualified series EC210 receptacle/plug connectors or Conax electrical seal assemblies will be used.

electrical seal assemblies will be used. The vendor has established a qualified life of 20 years for the Namco switches A plant specific qualified life evaluation wi 11 be conducted. and 40 years for the receptacle/plug connectors. A qualified life for the The vendor test report for the EC210 series connectors will be Conax seal assemblies must be established. A plant specific engineering obtained, if required. Conax Report IPS-1079 establishes qualifi- evaluation will be conducted to establish a qualified life for this equipment.

cation for the Conax seal assemblies. The vendors test report substantiating qualification for the EC210

[eceptacle/plug connectors will be obtained if the seal is used. 'f the Conax seal assemblies are used, Conax Report IPS-1079 establishes qualification for the seal assemblies. Upon completion of the plant specific qualified life evaluation and receipt of the series EC210 connector test report, this replacement equipment will be fully qualified.

Section 3.44 POS ITION. SMITCN TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY N/A Fisher Control N/A N/A 304 RESOLUTION:

Summary: /tails:

This equiixnent was added to the master list subsequent to the NRC This equipment was added to the master list of equipment subsequent to the NRC review. review based on a detailed systems review. The Fisher control model 304 This equiixa'.nt will be replaced with environmentally qualified position switch will be replaced with environmentally qualified Namco model Namco model EA180 position switches. EA180 position switches. Myle Report 17655-LSW-1.1, Revision A, Patel" Wyle Report 17655-LSW-l.l Revision A (Reference 56), Patel Report Report PEI-TR-82-12-9 and Namco Test Report QTR-105 establish environmental PE I-TR-82-12-9 and Namco Test Report OTR-105 (Reference 51) qualification. In order to assure proper sealing of the switch, estabIish qualification. environmentally qualified Namco series EC210 receptacle/plug connectors Namco qualified EC210 series receptacle/plug connectors or Conax or Conax electrical seal assemblies will be used. The vendor has established electrical seal assemblies will be used. a qualified life of 20 years for the Namco switches and ten years for the A plant specific qualified life evaluation will be conducted. receptacle/plug connectors. A qualified life for the Conax seal assemblies The vendor test report for the EC210 series connectors will be aust be established. A plant specific engineering evaluation will be con-obtained, if required. Conax Report IPS-1079 establishes ducted to ehtabiish a qualified life for this equipment. The vendors qualification for the Conax seal assemblies. test report substantiating qualification for the EC210 receptacle/plug connectors will be obtained if this seal is used. If the Conax seal assembly is used, Conax Report IPS-1079 establishes qualification for the seal assemblies. Upon completion of the plant specific qualified life evaluation and receipt of the series EC210 connector test report (if required), this lacement equipment will be fully qualified.

i

Section 3.45 POSITION SMITCH TER NO. HANUFACTURER/HOOEL NRC CATEGORY DEFICIENCY N/A Namco N/A N/A EA-180 RESOLUTION:

Suneary: Oetails:

This equipment was added to the master list subsequent to the NRC This equipment was added to the master list of equipment subsequent to the NRC review. review based on our detailed systems review. Myle Report 17655-LSW-l. 1 Myle Report 17655-LSM-l.l Revision A (Reference 56), Patel report revision A, Patel Report PEI-TR-82-12-9, and Namco Test Report QTR-105 PEI-TR-82-12-9 and Namco Test Report I}TR-105 (Reference 51) establish environmental qualification. In order to assure proper sealing establish qualification. of the switch, environmentally qualified series Namco EC210 receptacle/

Namco qualified EC210 series receptacle/plug connectors or Conax plug connectors or Conax electrical seal assemblies will be used. The vendor electrical seal assemblies will be used. has established a qualified life of 20 years for the Namco switches and 40 A plant specific qualified life evaluation will be conducted. years for the receptacle/plug connectors. A qualified life for the Conax The vendor test report for the EC210 series connectors will be seal assemblies must be established. A plant specific engineering obtained, if required. Conax Report IPS-1079 establishes qualifi- evaluation will be conducted to establish a qualified life for this cation for the Conax seal assemblies. equipment. The vendors test report substantiating qualification for the EC210 receptacle/plug connectors will be obtained, if this seal is used. If the Conax seal assemblies are used, Conax Report IPS-1079 establishes qualification for the seal assemblies. Upon completion of the plant specific qualified life evaluation and receipt of the series EC210 connector test report, this replacement equipment will be fully qualified.

I Section 3.46 POS IT ION SWITCH TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF I CIENCY 50 Namco Ir.a Criteria regarding radiation satisfied SL3L and SL3C58TW RESOLUTION:

Summary: Details:

The NRC incorrectly classified TER item 50 as The NRC considered TER item 50 to be qualified and were placed qua)i f i ed. in NRC Category I.a.; TER item 49 was found to be deficient with respect to We believe that TER item 50 is deficient with radiation and was placed in NRC Category I.b because we stated that the respect to similarity and radiation.

This equipment will be replaceii with environmentally qualified equipment would be replaced if qualification could not be established. Sub-sequent evaluation of GE report QSR-014-A-OI for this equipment has revealed Nmnco Hodel EA180 position switches. that adequate radiation qualification and similarity between the loss of Wyle Report 17655-LSW-1.1 Revision A (Reference 56), Patel Report coolant accident Namco switch model SL3HDP-OT and the installed Namco switch PE I-TR-82-12-9 and Namco Test Report QTR-105 (Reference 51) models SL3L and SL3C58TW was not established. Therefore, the TER evaluation establish qualification. was in error. Namco switch models SL3L and SL3C58TW will be replaced with Namco qualified EC210 series receptacle/plug connectors or Conax environmentally qualified Namco model EA180 position switches. Wyle electrical seal assemblies will be used. Report 17655-LSW-l.l revision A, Patel Report PEI-TR-82-12-9 and Namco Test A plant specific qualified life evaluation will be conducted. Report QTR-105 establish environmental qualification. In order to assure The vendor test report for the EC210 series conductors will be proper sealing of the switch, environmentally qualified, Namco series EC210 obtained, if required. Conax Report IPS-1079 establishes qualifi- receptacle/plug connectors or Conax electrical seal assemblies will be used.

cation for the Conax seal assemblies. The vendor has established a qualified life of 20 year s for the Namco switches and 40 years for the receptacle/plug connectors. A qualified life for the Conax seal assemblies must be established. A plant specific engineering evaluation will be conducted to establish a qualified life for this equipment. The vendors test report substantiating qualification for the EC210 receptacle/plug connectors will be obtained, if this seal is used. If the Conax seal assembly is used, Conax Report IPS-1079 establishes qualification for the seal assemblies. Upon completion of the plant specific qualified life evaluation and receipt of the series EC210 connector test r eport, this replacement equipment will be fully qualified.

Section 3.47 POSITION SMITCH TER f10. HANUFACTURER/HODEL NRC CATEGORY OEF ICIENCY N/A Namco N/A N/A EA170 RESOLUTION:

'Su+nary: Details:

This equipment was added to the master list subsequent to the NRC This equipment was added to the master list of equipment subsequent to the NRC review. review based on a detailed systems review. The Namco Hodel EA170 This equipcent will be replaced with environnentally qualified position switches will be replaced with environmentally qualified Namco Hodel Namco Hodel EA180 position switches. EA180 position switches. Myle Report 17655-LS'M-l.l, Revision A, Patel Myle Report 17655-LSM-l. 1 Revision A (Reference 56), Patel Report Report PEI-TR-82-12-9, and Namco Test Report l}TR-105 establish environmental PE I-TR-82-12-9 and Namco Test Report OTR-105 (Reference 51) qualification. In order to assure proper sealing of the switch, establish qualification. environmentally qualified Namco series EC210 receptacle/plug connectors Namco qualified EC210 series receptacle/plug connectors or Conax or Conax electrical seal assemblies will be used. The vendor has established electrical seal assemblies will be used. a qualified life of 20 years for the Namco switches and 40 years for the A plant specific qualified life evaluation will be conducted. receptacle/plug connectors. A qualified life for the Conax seal assemblies The vendor test report for the EC210 series connectors will be must be established. A plant specific engineering evaluation will be obtained, if required. Conax Report IPS-1079 establishes qualifi- conducted to establish a qualified life for this equipment. The vendors cation for the Conax seal assemblies. test report substantiating qualification for the EC210 receptacle/plug connectors will be obtained, if this seal is used. If the Conax seal assembly is used, Conax Report IPS-1079 establishes qualification for the seal assemblies. Upon completion of the plant specific qualified life evaluation and receipt of the series EC210 connector test report, this replacement equipment will be fully qualified.

Section 3.48 POSII'ION LIHIT SMITCH 87'ANUFACTURER/HOOEL TER ND. NRC. CATEGORY OEF ICIENCY

'14icroswitch Il.a Documented evidence of qualification adequate DTF2-2RN RESOLUTION:

Suraaary: Details:

Documented evidence of qualification not provided. Documented evidence of qualification was not provided at the time of the NRC Iiualification is established by Wyle Report )7655-LSW-13 review; therefore, a qualification assessment was to be conducted.

(Reference 135). Subsequently, Wyle Report 17655-LSW-13 was obtained, which The only harsh .-environment is radiation. establishes qualification for these position switches. These switches are This equipment is qualified. located in the reactor building and are required to be qualified for a radiation environment only. These position switches are qualified.

• The NRC TER/SER indicated Hicroswitch Hodel F. Plant specific information indicates Hicroswich OTF2-2RN.

Section 3.49 POSITION SMITCH TER NO. MANUFACTURER/HODEL NRC CATEGORY OEF ICIENCY N/A Laurence N/A N/A 506MA-26DC-SW-PS R ESOLUT ION:

Sumnary: Details:

This equipment was added to the master list after the NRC review. This equipment was added to the master equipment list subsequent to the NRC Qualification is established by Wyle Report 17655-LSM-8 review based on a detailed systems review. Subsequently, Wyle (Reference 116). Report 17655-LSW-8 was obtained which establishes qualification.

The only harsh environment is radiation. These switches are located in the reactor building and are required to be This equipaent is qualified. qualified for a radiation environment only. These switches are qualified.

Section 3.50 POSITION SWITCtl TER NO. HAHUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY 48 Hicroswitch I.b Adequate similarity between equipment and test specimen established 11LS1 and ILSI RESOLUTION:

Sumnary: Oe tails:

NUS Assessment Report 1961-H302-001 was deficient with respect to The NRC review concluded that NUS Assessment Report 1961-M302-001 did similarity. not establish similarity between the test specimen model LSA2B-I and the The only harsh environm'nt is radiation when the equipment is installed switches model 11LSl and 1LSl; we stated these switches will be required to function. replaced if qualification could not be substantiated. Subsequently, we have Iluaiification is established for the llLSl switch by Wyle determined that a model 1LSl switch is also mounted adjacent to the model llLSl Report'7655-LSW-12 (Reference 125). switch identified in the TER. Wyle Report 17655-LSW-12 establishes Ilualificatlon is established for the 1LS1 switch by Wyle qualification for the model llLSl switch; Hyle Report 17655-LSW-11 Report 17655-I.SW-ll (Reference 124). establishes qualification for the model 1LSl switch. The switches are located This equlpnent is qualified. in the reactor building and require qualification for a radiation environment only. This equipment is qualified.

Section 3.51 POSITION SWITCH TER tlO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY N/A Hicroswitch K/A N/A OTE6-2RN and OTE-2RN RESOLUTION:

Sumnary: Oetai is:

This equipm.'nt was added to the master list after the NRC review. This equipment was added to the master equipment list subsequent to the NRC Qualification is established by Wyle Report 17655-LSW-10 ~ review based on a detailed systems review. Subsequently, Wyle (Reference 123). Report 17655-LSW-10 was obtained which establishes qualification.

The only harsh environment Ia radiation. These switches are required to be functional after a loss of coolant accident; This equipment is qualified. the switches are located in the reactor building and are required to be qualified for a radiation environment only. These switches are qualified.

Section 3.52 POSITION SWITCH TER NO. NANUFACTURER/HODEL NRC CATEGORY DEFICIENCY N/A Hi croswi tch N/A N/A LSD4L RESOLUTION:

Sursaary: Details:

This equipment was added to the master list after the NRC review. This equipment was added to the master equipment list subsequent to the NRC Ilualification is established by Wyle Report 17655-LSW-09 review based on a detailed systems review. Subsequently, Wyle (Reference 117). Report 17655-LSW-09 was obtained which establishes qualification.

The only harsh environment is radiation. The switches are located in the reactor building and are required to be This equipment is qualified. qualified for a radiation environment only. These switches are qualified.

Section 3.53 POSITION SMITCH TER HO. HANUFACTURER/NODEL NRC CATEGORY DEFICIENCY N/A Hicroswitch N/A H/A BIE6-2RN RESOLUTION:

Sumnary: Details:

This equipment was added to the master list after the NRC review. This equipment was added to the master equipment list subsequent to the NRC Iluallfication is established by Myle Report 17655-LSW-14 review based on a detailed systems review. Subsequently, Hyle (Reference 127). Report 17655-LSW-14 was obtained which establishes qualification.

, The only harsh environment is radiation. The switches are located in the reactor building and are required to be This equipnent is qualified. qualified for a radiation environment only. These switches are qualified.

Section 3.54 POSITION SWITCH TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIEHCY Hamco I.b Adequate similarity between equipment and test specimen established 02400X RESOLUTION:

Sumnary: Details:

HUS qualification assessment report 1961-N007-001 was deficient The NRC review concluded that HUS qualification assessment report with respect to similarity. 196)-N007-001 was deficient with respect to establishment of adequate simi-This equipaent will be replaced with environmentally qualified larity between the loss of coolant accident test specimen, Namco model SL3 and Namco model EA180 position sw'itches. the installed equipment, Hamco model 02400X. The Hamco model DZ400X position Wy)e Report )7655-LSW-).1 Revision A (Reference 56), Pate) Report switch wil) be replaced with environmentally qualified Namco model EA)80 PEI-TR-82-12-9 and Hamco Test Report QTR-105 (Reference 5)) position switch and Wyle Report 17655-LSW-).1 revision A, Pate) Report establish qualification. PEI-TR-82-12-9 and Namco Test Report QTR-105 establish environmental A plant specific qualified life evaluation wi)1 be conducted. qua)ification. The vendor has estab)ished a qualified life of 20 The replacement equipment wi)l be qualified upon completion years for the Namco switches. A plant specific engineering evaluation will be of qua)if)ed life evaluation. conducted to establish a qualified life for this equipment. These switches are located in the reactor building and do not need to be environmentally sealed for high energy line breaks because they are used for loss of coolant accident mitigation only. Upon completion of the plant specific qualified life, this rep)acement equipment will be fully qualified.

Section 3.55 POSITION SMITCH TER NO. MANUFACTURER/MODEL NRC CATEGORY DEP ICIENCY N/A Namco N/A N/A D2400X RESOLUTION:

Sumnary: Details:

This equipnent was added to the master list subsequent to the NRC This equipment was added to the master list of equipment subsequent to the NRC ~

review. review based on a detailed systems review. The Namco model D2400X This equiprent will be replaced with environmentally qualified position switches will be rep)aced with environmentally qualified Namco model Namco model EA)80 position switches. EA180 position switches. Myle Report 17655-LSW-1. 1, Revision A, Pate)

Myle Report 17655-LSW-).l Revision A (Reference 56), Pate) Report Report PEI-TR-82-12-9 and Namco Test Report QTR-)05 establish environmental PE I-TR-82-12-9 and Namco Test Report (1TR-)05 (Reference 5)) qualification. The vendor has established a qualified life of 20 years for establish qualification. the Namco switches. A plant specific engineering evaluation wi)1 be conducted A plant specific qualified life evaluation will be conducted. to establish a qualified life for this equipment. These switches are located The replacement equipment will be qualified upon completion of in the reactor building and do not need to be environmentally sealed for high qualified life evaluation. energy line breaks because they are used for loss of coolant accident mitigation only. Upon completion of the plant specific qualified life, this replacement equipment will be fully qualified.

Section 3.56 POS IT ION. SWITCH TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF ICIENCY N/A Namco N/A N/A EA)70 RESOLUTION:

Suenar y: Details:

This equipment was added to the master list subsequent to the NRC This equipment was added to the master list of equipment subsequent to the NRC review. review based on a detailed systems review. The Namco model EA)70 This equ)paent wi)I be replaced with environsenta))y qualified position switches will be replaced with environmentally qualified Namco model Namco model EA)80 position switches. EA)80 position switches. Wyle Report 17655-LSW-).l. Revision A, Wyle Report 17655-LSW-l. 1 Revision A (Reference 56), Pate) Report Pate) Report PEI-TR-82-12-9 and Namco Test Report QTR-105 estab)ish PE)-TR-82-)2-9 and Namco Test Report QTR-105 (Reference 5)) environmental qualification. The vendor has established a qualified estab)ish qua)if)cation. life of 20 years for the Namco switches. A plant specific engineering A plant specific qualified life evaluation wi)1 be conducted. evaluation will be conducted to establish a qua)ified life for this equip-The replacement equipment will be qualified upon completion of ment. These switches are located in the reactor building and do not qualified )ife evalution. need to be environmentally sealed for high energy line breaks because they are used for loss of coolant accident mitigation only. Upon completion of the plant specific qualified life, this replacement equipment will be fully qualiFied.

Section 3.57 POSITION SWITCH TER NO. HANUFACTURER/HOOEL NRC CATEGORY DEFICIENCY 49, 100 Namco I I.a Adequate similarity between equipment and test specimen established SL3C581W Criteria regarding radiation satisfied RESOLUTION:

Sumnary: Details:

The NRC incorrectly classified TER item 100 as qualifed. The NRC considered TER item 100 to be qualified and was placed in NRC We believe that TER items 49 and 100 are deficient with respect Category I.a., TER item 49 was found to be deficient with respect to radiation to similarity and radiation. and was placed in NRC Category I.b because we stated that the equipment would This equipment will be replaced with environmentally qualified be replaced if qualification could not be established. Subsequent evaluation Namco model EA740 position switches. of GE Report QSR-014-A-01 for this equipment has revealed that adequate Wyle Report 17655-LSW-4.1 (Reference 90), Pate) report radiation qualification and similarity between the loss of coolant accident PEI-TR-82-12-6 and Namco Test Report QTR-111 (ReFerence 89) test Namco switch model SL3HOP-OT and the installed Namco switch model establish qualification. SL3C58LW was not established. Therefore, the TER evaluation was in error.

Namco qualified EC21Q series receptacle/plug connectors or Conax Namco switch model SL3C58TW will be replaced with environmentally qualified electric'a) seal assemblies wi)1 be used. Namco model EA740 position switches. Hyle Report 17655-LSW-4.1, Pate)

A plant specific qualified life evaluation will be conducted. Report PEI-TR-82-12-6 and Namco Test Report QTR-111 establish The vendor test report for the EC210 series connectors will be environmental qualification. In order to assure proper sealing of obtained if required. Conax Report IPS-1079 establishes the switch, environmentally qualified Namco series EC210 receptacle/plug qualification for the Conax seal assemblies. connectors or Conax electrical seal assemblies will be used. The vendor has established a qualified life of 20 years for the Namco switches and 40 years for the receptacle/plug connectors. A qualified life for the Conax seal assemblies must be established. A plant specific engineering evaluation will be conducted to establish a qualified life for this equipment. The vendor's test report substantiating qualification for the EC210 receptacle/p)ug connectors will be obtained iF this seal is used. If the Conax seal assembly is used, Conax Report IPS-1079 establishes qualification for the seal assemblies. Upon completion of the plant specific qua)ified life evaluation and receipt of the ser Ies EC210 connector test report (if required), this rep)acement equipment will be fully qualified.

Section 3.58 HOTOR COHTROL CENTER TER NO. HANUFACTURER/MODEL NRC CATEGORY DEFICIENCY 72 GE Ii.a Documented evidence of qualification adequate IC'l700 RE SOLUT ION:

Surya ry: Detai ls:

Documented evidence of qualification not provided at the time of Documented evidence of qualification was not provided at the time of the NRC the HRC review. review, and an assessment was to be conducted to establish qualifi-GE is presently conducting an analysis to establish cation. Consequently, the TER indicated that the documented evidence of quali-qualification. fication was inadequate. Presently, GE is conducting an engi-Hhen the analysis is completed; the equipment will be qualified. neering analysis to establish qualification. Hhen the engineering analysis is completed, the equipment wi 11 be qualified.

Section 3.59 MOTOR GENERATOR SET TER HO. MANUFACTURER/MODEL HRC CATEGORY OEF ICIENCY N/A GE N/A N/A SLS4404 A22Y25 RESOLUTIOH:

Surnaary: Details:

This equipment was added into the master list based on a detailed This equipment (MG-162 and MG-172) was added to the master list based on a system review conducted subsequent to the NRC review. detailed systems review subsequent to the NRC review. This equipment GE is presently conducting an analysis to establish qualification. is located in the general area of the turbine building on elevation 277 feet.

When the analysis is completed, the equipment will be qualified. The postulated high energy line break, which exposes this equipment to a harsh environment is a steam or feedwater line break in the steam tunnel.

The general areas of the turbine building are less than 133F under the postulated line break event; therefore, this equipment is not exposed to an excessively harsh environment. GE is presently conducting an engineering analysis to establish qualification. When the engineering analysis is completed, the equipment will be qualified.

Section 3.60 1-KW ELECTRIC STRIP HEATERS TER NO ~ HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY N/A Thermon N/A N/A TFK RESOLUTION:

Sumnary: ~ Details:

This equipuent was added to the master list subsequent to the This equipment was added to the master list of equipment subsequent to the NRC NRC review. eview based upon a detailed system review. One 1-KW electric These heaters were originally installed to prevent condensation strip heater is located at the inlet of each of the two charcoal filters in on charcoal adsorbers of the reactor building emergency venti- the reactor building emergency ventilation system. Heater 202-72 is lation system. associated with charcoal filter $ 11; 202-73 with charcoal filter f12. In The heaters are no longer required to function because the 10-KW addition to these strip heaters, a single 10-KW electric heater (202-76) is heater upstream of the charcoal filers will keep relative humidity of located in the comnon ducting upstream of both charcoal filters. The design the air stream below 70 percent. purpose of these heaters was as follows: 10-KW - Reduce relative humidity of The 1-KW strip heaters are not within the scope of 10CFR50.49. the inlet air from 100 percent to below 70 percent to maintain charcoal filter efficiency high. I-KW - Prevent condensation on the charcoal which would impair efficiency, particularly during system startup. Subsequent contact with the equipment vendor (Hine Safety Applicances Co. letter from Hr. R. D. Parco, Assistant Hanager Nuclear Filter Systems, dated Feb. 10, 1984) has revealed the following information: 1) Current practice to meet charcoal adsorber requirements (Regulatory Guide 1.52 paragraph 2a, 3b and ANSI N-509, section 5.5) is to use a heater upstream of the adsorber to control relative humidity below 70 percent. Strip heaters to eliminate condensation on charcoal are no longer comnon. 2) Nuclear grade charcoal currently used has been qualified based on 99.99 percent efficiency on elemental iodine at 95 percent relative humidity, 30C and 99 percent efficiency on methyl iodine at 95 percent relative humidity, 80C. 3) Where an upstream heater capable of maintaining relative humidity below 70 percent exists, strip heaters are not required. In view of the above information, the 1-KW strip heaters to eliminate condensation on the charcoal are no longer required. In addition, since the 1-KW strip heaters are located in the turbine building where the only harsh environmental parameter for an accident in the containment or reactor building is radiation, the strip heaters should function normally, particularly during the system startup. Any subsequent malfunction of the 1-KW heaters due to environmentally induced failure will have no effect on the performance of the charcoal filters. Consequently, the I-KW strip heaters (202-72 and 202-73) are not within the scope of IOCFR50.49.

Section 3.61 ELECTRICAL~ NEATER TER NO. MANUFACTURER/MODEL NRC CATEGORY DEFICIENCY N/A iioneywel 1 R72838 1081 N/A N/A

~

RESOLUTION:

Details:

'urnnary:

This equipment was added into the master list based on a detailed This equipment has been incorporated into the master list of equipment based system review conducted subsequent to the NRC review. on a detailed systems review conducted subsequent to the NRC review An engineering evaluation is being conducted to qualify the An analysis of this equipment is currently being performed by Hyle equipment. Laboratories. Radiation is the only harsh environmental parameter to which The only harsh environnent is radiation. this equipment is exposed, since the heater is in the turbine building and When the engineering evaluation is completed, the equipment will be it only has to operate For an accident in the containment or reactor building.

qualified. Consequently, it is anticipated that the Wyle analysis will show this equipment to be qualified. When the engineering evaluation is completed, the equipment will be qualified.

Section 3.62 ~II 0 MONITOR TER NO. MANUFACTURER/MODEL NRC CATEGORY OEF ICIENCY N/A

, Beckman N/A N/A H202 RESOLUTION:

Sumnary: Details:

This equipment was added to the master list based on a detailed This equipment has been incorporated into the master list of equipment based on system review conducted subsequent to the NRC review. a detailed systems review conducted subsequent to the NRC review. The The only harsh environmnt is radiation. equipment is a part of hydrogen and oxygen monitoring system and is required An engineering evaluation is being conducted to qualify the to remain functional after a loss of coolant accident. The equipment is equi pmen t. located in the turbine building and is required to be qualified for a When the engineering evaluation is completed, the equipment will radiation environment. An engineering evaluation is being conducted by Wyle be qualified. to qualify the equipment. When the engineering evaluation is completed, the equipment will be qualified.

Section 3.63 RADIATION OE'IECTOR TER NO. MANUFACTURER/HOOEL NRC CATEGORY OEFICIEHCY

5) GE Il.a Oocumented evidence of equal if ication adequate 194X927 RESOLUTION:

Sumnary: Octa'i ls:

The TER concluded that documented evidence of qualification was The TER concluded that documented evidence of qualification was not adequate.

not adequate. This equipment was deleted from the master equipment list based upon a The equipaent was deleted from the master list based on a detailed system review subsequent to the NRC review. Radiation detailed system review subsequent to NRC review. detectors RE RN05A, 0, C and 0 are installed on the main steam lines. These The detectors do not fall within the scope of 10CFR50.49. are not within the scope of 10CFR50.49.

Section 3.64 RADIATION DETECTORS TER NO. HANUFACTURER/MODEL NRC CATEGORY DEFICIENCY N/A GE N/A N/A 194 X927 RESOLUTION:

Surnnary: Details:

This equipment was added to the master list subsequent to the NRC This equipment was added to the master list of equipment subsequent to the NRC review. review, based upon a detailed system review. These detectors are The detectors subsequently were determined not to be within the the radiation detectors monitoring the shell side vents of the emergency scope of 10CFR50.49. condensers (RE-RN04A-3, 04A-4, 04B-3, 04B-4). These detectors are located in the reactor building. Subsequently, these detectors were determined not to be within the scope of 10CFR50.49.

Section 3.65 CONTROL PANEL TER NO. MANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY N/A GE N/A N/A Protective Panel for HG Set HG-162 and 172 RESOLUTION:

Smanary: Details:

This equiixnent was added into the master list based on a detailed This equipment (MG-162CP and HG-172CP) was'added to the master list based system review conducted subsequent to the NRC review. on a detailed systems review subsequent to the NRC review. This GE is presently conducting an analysis to establish equipment is located in the general area of the turbine building on elevation qualification. 277 feet. The postulated high energy line break which exposes this equipment When the analysis is completed, the equipment will be qualified. to. a harsh environment is a steam or feedwater line break in the steam tunnel. The general areas of the turbine building are less than 133F under the postulated line break event; therefore, this equipment fs not exposed to an excessively harsh environment. GE is presently conducting an engineering analysis to establish qualification. When the engineering analysis is completed, the equipment will be qualified.

Section 3.66 SOLENOID VALVE TER NO. IIANUFACTURER/MODEL NRC CATEGORY DEF ICIENCY IB Asco I.b Adequate similarity between equipment and test specimen established MPLB8300872F and WPLB8300868F Aging degradation evaluated adequately I)ualified life or replacement schedule established (if required)

Criteria regarding aging simulation satisfied (if required)

RESOLUTION:

Sumnary: Details:

The NRC concluded that solenoid valves68-08C, 09C The NRC for equipment item 18, Asco Hodei WPLB8300868F, concluded that and IOC were deficient with respect to similarity, aging solenoid valves68-09C, OBC, IOC were deficient with respect to similarity, and qualified life. aging evaluation and qualified life. Therefore, we intend to replace Subsequent to the NRC review, solenoid valve 201-10, 08, 16, 32 were added to the master list.

these solenoids if qualification could not be established. Subsequent to the NRC review, Asco solenoid valves model MPLB8300872F, 201-10, I)ualification is established by Hyle Report 201-08, 201-16 and 201-32 have been added to the master list based on a 17655-SOV-3 (Reference 126). detailed system review. Solenoid valves in this equipment group are used in Radiation is the only harsh environment. the containment atmospheric dilution system and containment and reactor vessel We'e replacing gaskets and the coil. isolation system. This equipment is located in the reactor building and does When the gaskets and coil are replaced, this equipment wi 11 be not require high energy line breaks qualification because it is used for loss qualified . of coolant accident mitigation only. Therefore, the only harsh environmental requirement for qualification is radiation. Operating requirements for this equipment are 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. Myle Report 17655-SOV-3 establishes qualification for

~ this equipment based on a materials radiation analysis and thermal aging evaluation. Replacement intervals (presently 3 to 5 years) for valve components including gaskets and coils were established in Hyle Report 17655-SOV-3 based on engineering analysis and. Asco recomnendations. However, we are reviewing temperature conditions at the actual location to extend replacement intervals. He're replacing age degraded components in installed solenoid valves in order to achieve full qualification; existing class 8 coils are being replaced with class H insulated coils. When the replacement component installation is completed, these solenoid valves will be qualified.

Section 3.67 600V CIRCUIT BREAKER TER tS. HANUFACTURER/HOOEL NRC CATEGORY OEF I CIENC Y 62 GE I I.a Documented evidence of qualification adequate AKO-5 RESOLUTION:

Suamary: Details:

Documented evidence of qual1fication not provided at the time Documented evidence of qualification was not provided at the time of the NRC of the NRC review. review, and an assessment was to be conducted to establish quali-GE is presently conducting an analysis to establish fication. Consequently, the TER indicated that the documented evidence of qualification. qualification was inadequate. Presently, GE is conducting an engineering If this analysis cannot substantiate qualification, this equipment analysis to establish qualification. If the engineering analysis will be enclosed in a cabinet assembly. substantiates full qualification, the equipment will be qualified. If the Upon completion of the analysis and fabrication of a cabinet analysis cannot substantiate full qualification, the GE AKD-5 switchgear will (if required), this equipment will be qualified. be enclosed in a cabinet assembly to preclude intrusion of humidity under design basis accident conditions. Upon completion of the engineering analysis and the fabrication of a cabinet assembly, if required, this equipment will be qualified.

0 Section 3.68 VALVE ACTUATORS.

TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY N/A Limitorque N/A N/A SHB000 RESOLUTION:

Sueaary: Details:

This equipnent was added to the master list subsequent to the NRC Valves (70-92 and 70-94) were added to the master equipment list sub-review. sequent to the NRC review, based upon a detailed systems review.

This equipment subsequently has been determined not to be within They are the isolation valves on the return lines from the reactor building the scope of 10CFR50.49. closed loop cooling heat exchangers for the drywell air coolers and the recirculation pump coolers. This equipment has subsequently been determined not to fall within the scope of 10CFR50.49.

Section 3.69 HOTOR CO44TROL CENTER TER NO. HA44UFACl'URER/HOOEL NRC CATEGORY OEF ICIENCY N/A GE N/A N/A IC7700 RESOLUTION Suamary: Details:

This equiprrent was added into the master list based on a detailed These motor control centers (155, 161A/B, 171A/B," BB12) were added into the system review conducted subsequent to the NRC review. master list, based on a detailed systems review subsequent to the NRC review. .

GE is presently conducting an analysis to establish GE is presently conducting an engineering analysis to establish qualification. qualification. Nhen the engineering analysis is completed, the equipment will When the analysis is completed, the equipment will be qualified. be qualified,

Section 3.70 600V CIRCUIT BREAKER TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY N/A GE N/A N/A AKO-5 RESOLUTION:

Sunmary: Oetai ls:

This equipment was added into the master list based on a detailed This equipment (PB-16A and PB-17A) was added into the master list, based on a system review conducted subsequent to the NRC review. detailed systems review subsequent to the NRC review. GE is presently GE is presently conducting an analysis to establish conducting an engineering analysis to establish qualification.

qualification. If the engineering analysis substantiates full qualification, the equipment If the analysis cannot substantiate qualification, the equipment will be qualified. If the analysis cannot substantiate full qualification, will be enclosed in a cabinet assembly. the GE AKO-5 switchgear will be enclosed in a cabinet assembly to preclude Upon completion of the analysis and fabrication of a cabinet (if intrusion of humidity under design basis accident conditions. Upon completion required), this equipment will be qualified. of the engineering analysis and the fabrication of a cabinet assembly, if required, this equipment will be qualified.

Section 3.71 'SOLENOID VALVE TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY N/A Asco N/A N/A HTX8320A22U RESOLUTION:

Suamary: Details:

This equipment was added to the master list based on a detailed The following Asco Hodel HTII8320A22U solenoid valves were added to the master systems review.

Iiualification is established by Wyle Report 17655-SOV-4 equipment list, based on a detailed systems review subsequent to the NRC review: 201.2-431, 432, 429, 430, 421, 422, 420, 419, 201.7-24, 25, (Reference 134). 22, 27, Solenoid valves in this equipment group are used in Radiation is the only harsh environment. the H202 monitoring system and are required to remain functional after a The report will be revised with respect to qualified life and loss of coolant accident. This equipment is located in the reactor building replacement schedule. and must be qualified for radiation only. Wyle Report 17655-SOV-4 has This equipment <<ill be qualified. established qualification for radiation service conditions. This report will be revised to substantiate qualification with respect to qualified life and component replacemerit schedule consistent with plant specific operating requirements. When the report revision is completed, this equipment will be qualified.

Section 3.72 SOLENOID VALVE TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY 10 Laurence I I.a Documented evidence of qualification adequate 506WA26DC-SW-PS RESOLUTION:

Sumnary: Octa i ls:

Doc+vented evidence of qualification not provided at the time Documented evidence of qualification was not provided at the time of the NRC.

of the NRC review. review, and an assessment was to be conducted to establish qualifi-ualification is established by Myle Report, 17655-SOV-I cation. Therefore, documented evidence of qualification was indicated by the Reference 71) and NUS Analysis Report, CO-ENG-853, TER as inadequate. The solenoid valves are located in the reactor building (Reference 68). and are used as drywell hydrogen/oxygen sampling isolation valves which are Radiation is the only harsh environment. normally closed and de-energized. These valves are not required to function Equipaent is qualified. during a high energy line breaks. They are required to function du> ing and .

after a loss of coolant accident. Therefore, the equipment is in a mild environment except for radiation. Myle Analysis Report, 17655-SOV-I and NUS Report CO-ENG-853 established qualification. Therefore, the equipment is qualified.

Section 3.73 SOLENOID VALVE HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY TER NO.

15 Asco I I. a Documented evidence of qualification adequate WP8300B61RU 2Uld 8300B61RU RESOLUTION:

Suamary: Details:

Documented evidence of qualification was not provided at the The NRC concluded that the Asco solenoid model 8300B61RU for solenoid time of the NRC review for solenoid valves 201.2-06, 33. valves 201.2-06 and 201.2-33 was deficient with respect to documented evidenpe Subsequently, solenoid valves 83.1-10, 12 and 58.1-01 were of qualification. Therefore, we stated that a qualification assessment added to the master list. would be conducted. Subsequently, Asco Model MP8300B61RU solenoid I)ualification is addressed by Wyle Report 17655-SOV-5 valves (solenoids 83.1-10, 83.1-12) and Asco Hodei 8300B61RU (solenoids (Reference 137). 58.1-01) we added to the master list based on a detailed systems review. Wyle Radiation is the only harsh environnmnt. Report 17655-SOV-5 addresses qualification for these Asco solenoid valves.

We'e replacing gaskets, seals and coils on these solenoids . This equipment is located in the reactor building and is required to function An analysis will be conducted to substantiate qualification for during a loss of coolant accident; high energy line break qualification is not the replacement ccmponents and to establish a plant specific required. Radiation is the only harsh environmental requirement for qualified life. qualification. Based on a materials evaluation, Wyle Report 17655-SOV-5 Mhen the subcanponents and coil are replaced and the additional addresses qualification for the radiation environment. This same report analysis is completed, this equipment will be qualified. I recoamended that valve components be upgraded with metallic seats, class H insulated coils and new gaskets and seals. We will perform these replacements, in order to achieve fully qualified equipment. Additional analysis will be performed to substantiate qualification including development of plant specific replacement interval for age-degradable parts (I.e., solenoids and gaskets). Mhen the replacement components are installed and the additional analysis is completed, these solenoid valves will be qualified.

Section 3.74 SOLENOID VALVE TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY N/A Asco 8300C68/8300C68F N/A N/A ~

R)SOLUTION:

'urnnary:

Details:

This equipaent was added to the master list of equipment sub- These solenoid valves (SV/BV 202-15, 16, 31, 32, 34, 35) were added to the sequent to the IIRC review. master list of equipment subsequent to the IIRC review, based on a detailed Wyle Report )7655-SOV-6 (Reference 138) establishes systems review. These solenoid valves are used in the reactor building emer-qualification. gency ventilation system. This equipment is located in the turbine building The only harsh environaent is radiation. and is required to function during a loss of coolant accident. .Ituaiification We'e replacing the existing solenoid coils. for high energy line break conditions is not required. Radiation is the only An analysis will be conducted to extend the current replacement harsh environmental condition for which qualification is required. Hyle interval of 4 to 5 years for the solenoid coils. Report 17655-SOV-6 has established qualification for service conditions, based When the coil replacement is complete, this equipment'will be on radiation and thermal analysisof solenoid valve materials. We are qual i f i ed. replacing existing solenoids in accordance with the recomnendations in Wyle Report 17655-SOV-6 to establish fully qualified equipment. An additional analysis will be performed to determine if the recomnended 4-5 year replacement interval for components can be extended. This equipment will be qualif led.

Section 3.75 SOLENOID VALVES TER NO. MANUFACTURER/MODEL NRC CATEGORY OEF ICIENCY 13 Asco I.b Aging degradation evaluated adequately HVA 90 40BA

~

Ilualified life or replacement schedule established (if required)

Criteria regarding aging simulation satisfied (if required)

RESOLUTION:

Sumnary: Details:

The NRC concluded that solenoid valves SV-NC-15 A, B The NRC concluded that the documentation for solenoid valves and SV-NC-16 A, B were deficient with respect to aging and SV-NC-15 A and B and SV-NC-16 A and B was deficient with respect to aging and qualified life assessment. qualified life assessment. These valves, through a detailed systems review, These valves have been determined not to fall within the scope have been determined not to fall within the scope of 10CFR50.49.

of 10CFR50.49.

Section 3.76 SOLENOID VALVES 1'ER NO. MANUFACTURER/MODEL NRC CATEGORY OEF I CIENCY N/A Asco N/A N/A 8300D6IU RESOLUTION:

Sumnar y: Details:

This equipment was added to the master list after the NRC Two solenoid valves on equipment piece number 202-'36 were added to the review. after the TER review, based on a detailed systems review. These master'ist Radiation is the only harsh environmental condition. valves are used in the reactor building ventilation system for control of air This equipment will be replaced with qualified Asco NP-1 solenoid operated dampers. The solenoid valves are located in the turbin'e building and valves. do not require qalification for high energy line breaks, since they function i}ualification of the NP-1 solenoid valve is established by Asco only during a loss of coolant accident. Radiation is the only harsh Report AljS21678/TR Revision A. environmental requirement for qualification. This equipment will be replaced A qualified life/replacement schedule assessment wi II be per formed . by qualified Asco NP-1 solenoid valves, model 206-832-3U and a plant specific IIhen the valve replacement and qualified life evaluation are completed, qualified life evaluation performed. i}ualification of the NP-1 solenoid valve the equipment will be qualified. is established by Asco Report AI}S21678/TR Revision A. This equipment will be qualified after installation and qualified life evaluations are completed.

Section 3.77 SOLENOIO VALVES TER IS. MANUFACTURER/MODEL NRC CATEGORY OEF ICIENCY N/A Asco N/A N/A WP831630 RESOLUTION:

Suamary: Details:

This equipnent was added to the master list after the NRC review. Solenoid valves (SV/BV 202-37,38,74,75) were added to the master equipment Radiation is the only harsh environmental condition. list subsequent to the TER, based on a detailed systems review. These A qualification analysis will be performed for this equipment. valves are used in the reactor building emergency ventilation system for con-This equipment will be qualified after completion of the trol of air supply to the air-operated dampers. The solenoids are located in analysis and replacement of valve components (if required). the turbine building and do not require qualification for high energy line breaks, since they function only during a loss of coolant accident. Radiation is the only harsh environmental requirement for qualification. A qualification analysis, including assessment of qualified life, wi 11 be perforned for these valves, and component parts replaced (if required). It is anticipated that this valve can be qualified by analysis because of its similarity to other Asco models. When this work is completed, these solenoid valves will be qualified.

Section 3.78 SOLENOIO VALVES TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY 17 Asco N/A Adequate similarity between equipment and test specimen established 8300 Series Aging degradation evaluated adequately Oualified life or replacement schedule established (if required)

Program established to identify aging degradation Criteria regarding aging simulation satisfied (if required)

Criteria regarding radiation satisfied Criteria regarding functional testing satisfied RESOLUTION:

Sundry: <Details:

Asco solenoid valves (SV39-05E, 5F, 06E, 6F) have been replaced The TER concluded that Asco Hodel 8300 solenoid valves by Valcor solenoid valves. (SV39-05E, 05F, 06E, 06F) were deficient with respect to similarity, aging, Valcor valves also were added into the master list, based on a de- ~

qualified life, radiation and functional testing. Subsequent to the NRC tailed systems review conducted subsequent to the NRC review. review, these solenoid valves were replaced with Valcor Hodel V70900-21. In Qualification addressed by Wyle Report 17655-SOV-2.1 addition, the following additional solenoids have been added to the master list Revision A (Reference 113), Valcor Test Report HR70900-39-1-1 based on a detailed systems review: SOV39-11C, 110, 12C, 120, 13C, (Reference 82) and Valcor Test Report QR70900-21-1 (Reference 81). 130, 14C, 140; SYNC-15C, 150;60-170, 17E;60-180, 18E. Valcor Test An additional engineering analysis will be conducted to establish Report HR70900-39-1-1 and I)R70900-21-1 were obtained, which a qualified life based on vendor test data. establish qualification. Wyle Report 17655-SOV-2. 1 Revision A, will be revised When the additional qualified life engineering evaluation is com- to establish a qualified life based on plant specific conditions and pleted, the equipment will be qualified. >actual aging test data presented in Valcor Report HR-70900-39-1.1. When the additional qualified life analysis is completed, the equipment will be qualified.

Section 3.79 SOLENOID VALVE TER N. HANUFACTURER/NOBEL NRC CATEGORY OEF I C IENCY 16, 17 and 79 Asco I.b for TER 16'7 Documented evidence of qualification adequate (TER 79)

NP8344A71E Il.a for TER 79 Adequate similarity between equipment and test -specimen 8300861RU established (TER 16, 17)

HT8320A90 Aging degradation evaluated adequately (TER 16, 17)

HT8300858BU Iiualified life or replacement schedule establish (TER 16, 17) 83008 &lF Criteria regarding aging simulation satisfied (TER 16, .17)

HT830086RU Program established to identify aging degradation (TER 17)

WPHTX8300861U Criteria regarding radiation satisfied (TER 17)

HT8317A30 Criteria regarding functional testing satisfied (TER 17)

LB8320A-25 RESOLUTION:

Sumnary: Details:

t For TER No. 16, the NRC review concluded that the documentation ,

For TER Item 16 (Asco Hodel 8300861RU), the,NRC review concluded was deficient with respect to similarity, aging and qualified that the qualification documentation for solenoid valves 201.2-32 and li fe. 201.2-03 was deficient with respect to similarity to the test specimen, aging For TER No. 79, the NRC review concluded that qualification and qualified life. For TER Item 79, the NRC review concluded that documentation was not provided. the qualification documentation for solenoid valves 122-04 through 122-11 was Subsequent to the NRC review, additional solenoid valves were added not provided to substantiate qualification. We stated that these solenoid to the master list based on a detailed systems review. valves would be replaced if qualification could not be established. Subsequent This equipment will be replaced by qualified Asco NP series to the NRC review, the following solenoid valves were added to the solenoid valves. master list based on a detailed systems reviews:

Asco Test Report Ai)S21678 Revision A (Reference 178) establishes Hanufacturer Hodel Plant I.O.

qualification. Asco NP83AVIE 05-01R, 02K, 53R, RR, 05-11, 05-12 A plant specific qualified life evaluation wi II be conducted. Asco HT8320A90 40-328, 32C,40-338, 33C When the equipment is replaced and the qualified life evaluation is Asco iiT83008588U 39-05, 06 ccmpleted, this equipment will be qualified. Asco 8300B61F 201.2-02, 04 Asco HT830086RU 201.7-20, 21, 23, 26 Asco WPHTX8300861U 01-03, 04 Asco NP8344A71E 201. 1-09, 11 Asco HT8317A30 201.9-91, 92 Asco LB8320A-25 201.7-01, 02 We have determined that all the solenoid valves mentioned above will be replaced with qualified Asco NP series solenoid valves. Ilualification of the Asco NP solenoid valves is established by Asco Report AI}521678/TR Revision A.

A aualified life evaluation will be performed for these replacement solenoid vaTves based on plant specific conditions. When this equipment IS replacqg and the additional qualified life analysis is completed, this equ>pment will be qualified.

A Section 3.80 FLOW TRANSHITTER TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY N/A Endevco N/A N/A 2273AH20 RESOLUTION:

Sumaary: Details Equipment was added to the master list subsequent to the NRC review. This equipment (FT-19, 20, 21, 22, 23, 24) was added to the master equipment ~

This equipment was installed as a THI Action Plan item to provide list subsequent to the NRC review, based upon a detailed systems the operator with indication of relief valve opening. review. These items are accelerometers and have been added to the automatic Preliminary testing for the BWR Owner's Group indicates the equipment depressurization system relief valves as THI action plan items. Their purpose to be qualified. is to provide the operator with positive indication of the opening of a relief The test report is being evaluated for qualification assessment. valve. The BWR Owners Group has undertaken to qualify this equipment. Pre-When this is completed, the equipment will be qualified. liminary testing has shown the equipment to be qualified. The testing performed envelopes the environmental conditions at our plant. The test report is expected to be available for review shortly. We will add heat shrink tubing over the connectors. The test report is being evaluated for qualification assessment. When this is completed, the equipment will be qualified.

Section 3.8I SOLENOIO VALVE TER NO. NANUFACTURER/NOBEL t)RC CATEBORY OEF ICIENCY N/A Asco N/A N/A ilT8320A90NB RESOLUTION:

Sugary: Details:

This equiixient was added to the master list after the NRC These solenoid valves (122-03B, 03C) were added to the master equipment list review. subsequent to the NRC review, based on a detailed systems review. This Radiation is the only harsh environmental condition. equipment is used in the containment and reactor vessel isolation system for Wyle Report 17655-SOV-8.1 (Reference 162) addresses qualification. post-accident sample line isolation. The solenoid valves are located in the Additional analysis will be performed with respect to similarity, reactor building and are required to function after a loss of coolant qualified life and long-term post-accident functional capability. accident. Radiation is the only harsh environmental requirement for This equiixnent will be qualified . qualification. Wyle Report 17655-SOV-8.1 sumnarizes an analysis which addressed qualification for short-term post-accident monitoring. Additional analysis will be performed to establish similarity between HT8320 and NP-1 Asco solenoid models, extend qualification to long-term post-accident operational requirements and revise qualified life/replacement intervals, based on plant specific conditions. This equipment will be qualified.

C' Section 3.82 SOLENOID VALVE TER NO. MANUFACTURER/MODEL NRC CATEGORY DEFICIENCY 11 and 12 Oecco Documented evidence of qualification adequate 24166 RESOLUTION:

Sumnary: Details:

We have determined that these solenoid valves are manufactured The TER listed these solenoid valves as Oecco Model 24166; subsequent to the .

by Numatics. NRC review, we have determined that these solenoids are Numatics Model Documented evidence of qualification was not provided at the time 1JSP3 and 46JLSAD3. The TER concluded that documented evidence of qualifi-of the NRC review. cation had not been provided for these solenoid valves (SV/IV 80-15, 16, This equipnent will be replaced by qualified ASCO HP service 35, 36, 01-05, 06) and we stated that these so)enoid vaves would be re-solenoid valves. placed if qualification could not be established. The Humatics solenoid Asco Test Report AI)S2)678/TR Revision A (Reference )78) establishes valves will be replaced by qualified Asco NP series solenoid valves. Itua)-

qualification. ification of the ASCO NP solenoid valves is established by Asco Report A plant specific qualified life will be conducted. A(S2)678/TR Revision A. A qualified life evaluation will be performed for When the equipment is replaced and the qualified life evaluation is these replacement solenoid valves, based on plant specific conditions.

completed, this equipment wi)1 be qualified. When this equipment is replaced and the additional qualified life analysis is completed, this equipment will be qualified.

Section 3.83 DIFFERENTIAL PRESSURE TRANSMITTERS TER NO. MANUFACTURER/MODEL NRC CATEGORY DEF I CIENCY 25, 38 GE/MAC I.b Documented evidence of qualification adequate 551, 553, 554 RESOLUTION:

Suamary: Details:

The TER evaluated GE/HAC aadels 551 and 553 as lacking documented Three of these transmitters were evaluated by the TER as deficient in documen-evidence of qualification. ted evidence of qualification as follows:

Subsequently, additional GE/MAC eadel 553 and 554 transmitters TER Item No. Hodel No. Plant ID No.

were added to the master list. K 3%i p~=QA, les Hodel 553 has been tested by Hyle Laboratories, with similarity 38 553 LT IA-12 established to models 551 and 554. Preliminary results indicate The remaining transmitters were added to the master equipment list subsequent all models are qualified. to the NRC review, based on a detailed systems review, as follows:

As soon as the test report is obtained and reviewed, the transmitters Plant IO No. Model No: Function will be qualified. FT 80-4%A, 964, 7TK, 76A AS Containment spray flow FT RV-268 553 Core s'pray flow FT 202-49A, 92A 554 Emergency ventilation flow FT 93-30A, 32A, 33A, 34A 553 Containment spray raw water flow LT IG-06A, 068 553 Emergency condenser level GE/MAC transmitter, model 553, has undergone testing by Wy1e Laboratories.

The test parameters envelope the accident environmental conditions at Nine Mile Point 1. The test report wi 11 also show similarity to GE/MAC models 551 and 554. Preliminary results of this testing indicate that the transmitters are qualified. As soon as the test report is obtained and reviewed, these transmitters will be qualified.

Section 3.84 TRANSHITTERS TER NO. MANUFACTURER/MODEL NRC CATEGORY OEF I CIENCY 80, 81, 82, 83 Rosemoun t II.a Documented evidence of qualification adequate (TER 80, 81, 83) 1153 DA5, 1153 GA9, 1153 GA7 Aging degradation evaluated adequately, 0ualified life or replacement schedule established (if required) (TER 82)

RESOLUTION:

Suamary: Oetai ls:

For TER items 80, 81 and 83, the NRC review concluded that For TER item no. 80, Rosemount Model 1152T0280 (LT-58-05 and -06), the NRC documented evidence of qualification had not been provided. review concluded that documented evidence of qualification had not been pro-For TER item 83, the NRC revigw concluded that the qualifica- vided. Subsequent to the NRC review, we have determined that these .trans-tion documentation was deficient with respect to aging and qualified mitters are Rosemount Model 1153OAS. For TER item 81, Rosemount Model li fe. 1153GA9 (PT-36-23A and 23B), the NRC review concluded that documented evidence Subsequent to the NRC review, level transmitters LT-60-22 and 23 were of qualification had not been provided. For TER item 83, Rosemount Model added to the master list. 1153OAS (LT-36-24A and 24B), the NRC concluded that documented evidence of Iiualification is established by Patel Report PE I-TR-82-12-14 qualification was not provided. For TER item 82, Rosemount Model 1153 series (Reference 60) including Rosemount Test Report 3788 Revision A and A (PT-201.2-483 and 484), the NRC review concluded that the qualification Test Report 108026 and Wyle Report 17655-XHR-I.I Revision A documentation was deficient with respect to aging and qualified life eval-(Reference 66). uation. Subsequent to the NRC review, we have determined that these trans-An evaluation wi 11 be conducted to resolve the TER concern regarding mitters are Rosemount Hodel 1153GA7. Subsequent to the NRC review, Rosemount o-ring failures on recent Rosemount testing programs. Model 1153OAS transmitters for LT-60-22 and LT-60-23 were added to the master When the additional evaluation is completed, this equipment will be list based on a detailed systems review. Subsequent to the NRC review, fully qualified. the following qualification documentation was obtained which establishes qualification of these transmitters: Pate) Report PEI-TR-82-12-14 including Rosemount Test Report 3788 Revision A and Test Report 108026 and Wyle Report 17655-XHR-l. I Revision A. An evaluation wi 11 be conducted to resolve the TER concern regarding o-ring fai lures on recent Rosemount testing programs. Upon completion of these additional analyses and evaluations, these transmitters will be qualified.

Section 3.85 TRANSHITTER TER NO. MANUFACTURER/HOOEL NRC CATEGORY DEFICIENCY Foxboro Il.a Documented evidence of qualification adequate EI3OL RESOLUTION:

Suneary: Details:

For TER item 84, the NRC review concluded that documented for TER item 84,.Foxboro Hodel E13DL transmitter (FET 664), the NRC evidence of qualification was not provided. review concluded that documented evidence of qualification had not been Subsequent to the NRC review, transmitters (FT201.8-41, FT201.9-31; provided. Subsequent to the NRC review, the following have been added to PT201.8-35, 45; PT201.9-26, 80; FT201.2-533, 534) were added to the the master list:

mm-.Ar mr master list. ~ Plant- I.D. No.- Model No.'-

~

All of these transmitters are being replaced by qualified Rose-mount transmitters 1153 series D. F T201. 9-31 2340 IIhen these transmitters are replaced with qualified transmitters, PT201.8-35 2340 this equipm'nt is qualified. PT201.8-45 2340 PT201.9-26 2340 PT20I.9-80 2340 FT201.2-533 To be determined FT201.2-534 To be determined All of these transmitters are being replaced by environmentally qualified Rosemount Model 1153 series 0 transmitters. Qualification for these replacement units is established by Rosemount Test Report D8300040. Nhen these transmitters are replaced, this equipment will be qualified.

t Section 3.B6 ELECTRONIC TRIP UNIT TER NO. HANUFACTURER/HOOEL NRC CATEGORY DEF ICIENCY 19 and 20 Ros emoun t Il.a Aging degradation evaluated adequately 510 OU i}ualified life or replacement schedule established (if required)

Peak pressure adequate Steam exposure (if required) adequate Criteria regarding radiation satisfied RESOLUTION:

Sumnary: Details:

The NRC concluded that the equipment was deficient The NRC concluded that the qualification documentation for the Hodel with respect to aging, qualified life, pressure, steam and 510DU Rosemount electronic trip units was deficient with respect to aging and radiation exposure. qualified life assessment, exposure to peak pressure, exposure to steam (high Wyle Report 17655-TU-1.1 (Reference 141),BWR Equipment humidity) and radiation. Subsequent to the NRC review, Wyle Report I}ualification Suarnary Report i}SR-Oll-A-01 (Reference 142) 17655-TU-I. I, BWR Equipment I}ualification Summary Report i}SR-Oll-A-Ol and Rosemount Report 12777D (Reference 143) address qualification. and Rosemount Report 127770 were obtained which address Additional engineering analysis and evaluations will be performed qualification of the Hodel 5.10 OU electronic trip units.

to resolve all TER concerns. Additional engineering evaluations and analysis will be performed to completely When the additional analysis is completed, this equipment will be resolve all TER concerns. When the additional engineering analysis and eval-qualified. uations are completed, this equipment will be qualified.

Section 3.87 VALVE ACTUATORS TER t@. HANUFACTURER1HODEL NRC CATEGORY DEFICIENCY 2, 3, 7, 75 Limitorque Il.a Documented evidence of qualification adequate SHB RESOLUTION:

Suenary: Details:

TER Items 2, 3 and 7 were deficient with respect to qualification documentation; TER Item 75 had similarity qualified life, evidence of qualification and aging deficiencies.

Me added valve actuators (38-01, 02) to the master list after the TER review based on a systems review.

This equipment is located in the drywell and will be replaced with Plant Tiqirm I VOI-02 IV33-01

'ir At the time of the NRC review, the TER concluded that deficiencies existed in .

the qualification documentation of these valves as follows:

ID No. Hodel SH84 SHBO TER No.

2 2

TER Deficiencies Documeot'eeev>dence Documented Documented evidence evidence of quatiftcatioo of of qualification qualification qualified Limitorque valve actuators. IV33-02 SHBO 2 Documented evidence of qualification I)ua)ification testing was performed and reported in Limitorque IV40-01 SHB3 7 Documented evidence of qualification Reports 80058 and 600376A. Wy)e Report 17655-HOV-).1 and I V40-09 SH83 7 Documented evidence of qualification 1.2 (References 179 and )80) and Patel Report PEI-TR-82-12-8) IV40-IO SHB3 3 Documented evidence of qualification (Reference 92) address qualification. I V40-11 SHB3 3 Documented evidence of qual ification Additional analysis will be performed for qualified life and I V83. ) -09 SHBOOO 7 Documented evidence of qualification dern)nera)ized water spray effects. IV83.)-11 SHBOOO 7 Documented evidence of qualification Valve actuators wi)1 be inspected to verify components. I V110-127 SHBOOO 75 Adequate similarity This equipment will be qualified after completion of qual- Documented evidence of qualification ification analysis and replacement installation. Aging degradation, qualified life Subsequent to the NRC review, valves IV38-01 and IV38-13 were added to the master equipment list, based on a detailed systems review. Since the TER review, the following valves have been replaced or are scheduled for rep)acement:

Replaced Scheduled for Replacement AUSlA fVZl=dT I V40-01 I V-01-02 I V40-09 IV-33-02 IV40-10 IV-40-11 I V-83. 1-09 I V-83. 1-11 IV-)10-127 IV-38-01 IV-38-13

Section 3.87 VALVE ACTUATORS

~ont>nue8$

Details: (Continued)

Wyle Report 17655-HOV-I.I and l7655-HOV-I.2 assessed the environmental qualification for these valves. These assessments were based on engineering evaluations performed by Patel. Report PEI-TR-82-12-8, and testing described in Limitorque Test Reports 80058 and 600376A. Additional analyses will be performed to determine qualified life and evaluate the effects of demineralized water spray. We will perform inspections of these Limitorque actuators to ensure installed components (i.e., torque switches, limit switches, terminal blocks, wires, etc.) conform to vendor component lists and materials subjected to qualification testing. This equipment wi 11 be qualified when installation of qualified replacement actuator s, engineering analyses and inspections is completed.

Section 3.88 TRANSHI TIER TER NO. HANUFACTURER/HOOEL NRC CATEGORY DEFICIENCY 27, 29, 30, 35, 39 Rosemount I I. a Documented evidence of qualification adequate 1)51 OP, )151 GP RESOLUTION:

Suenary: Details:

Documented evidence not provided at the time of the t)RC review. The NRC concluded that documented evidence of qualification for the Oua) if)cation is established by Pate) Report PEI-TR-82-)2-10 following Rosemount transmitters was not provided: FT36-06A, 8, C, 0; (Reference 59) and Wyle Report )7655-XHR-2.1, Revision 8 (Reference 67). FTRV-26A; LT36-03A, 8, C, 0; LT36-04A, 8, C, D: LT36-05A, 8, C, D; An additional evaluation will be conducted to resolve the TER concern PT36-07A, 8, C, 0; PT36-08A, 8, C, 0; PT201.2-476A, 8, C, D. Therefore, regarding o-ring failure on recent Rosemount testing programs. we stated that qualification assessment would be conducted to establish Upon completion of the additional evaluation, these transmitters qualification. Subsequent to NRC review, Pate) Report PEI-TR-will be qualified. 82-12-10 and Wyle Report 17655-XHR-2. I Revision 8 were obtained which establish the qualification for the 1151 Rosemount transmitters. An evaluation will be conducted to resolve the TER concern regarding o-ring failure on recent Rosemount testing programs. Upon completion of the additional evaluations, the transmitters will be qualifed.

t Section 3.89 VALVE. ACTUATORS TER NO. HANUFACTURER/MODEL NRC CATEGORY DEFICIENCY I, 6, 8 Limitorque I).a Documented evidence of qua)if ication adequate SMB Adequate similarity between equipment and test specimen established Aging degradation evaluated adequately Oua)ified life or replacement schedule established (if required)

RFSOLUT ION:

Surnaary: Details TER Items 6 and 8 were deficient with respect to similarity, aging and qualified life; TER Item I had documentation deficiencies.

We added valve actuator 34-0) to the master list after the TER review.

At the time of the t)RC review, Limitorque Report 8003 was pro-vided as evidence of qualification. Subsequent investigations de-termined similarity was not adequately established.

At the time of the Plant ID No.

$$ -Q 34-01 39-07 39-08 NRC Hodel 3RSl SHBOOO SM82 SHB2

~No..

review, the qualification deficiencies existed:

TER N/A 6

6 TER concluded that the following (Added TER-Def)cien~c Docaae'naae to Similarity, Simi)arity, evidence list) aging, aging, of qoaltficatton qualified qualified life life Wyle Report 17655-HOV-2 (Reference 105) addresses qualification. 39-09 SH82 6 Similarity, aging, qualified life Field inspections and additional evaluations will be performed to 39-10 SHB2 8 Similarity, aging, qualified life determine qualification status based on similarity to components 201-31 SM8OOO 8 Similarity, aging, qualified life tested in qualification reports. Valve actuator 34-01 was added to the master list based on a detailed This equipment will be qualified after completion of inspections, systems review. Subsequent to the TER, Wyle Labs determined that no specific evaluations and replacement of components, as necessary. qualification reports exist for these valve actuators. Hyle Report 17655-HOV-2 identified typical components and materials used in the construction of qualified Limitorque actuators. The report recoranended that field inspections be performed to determine the components used in equipment installed in Nine Hile Point l. An assessment will then be made of the qualification status based on similarity of installed components to Limitorque components tested in existing Limitorque qualification reports. These actuators are scheduled for inspection to determine specific actuator components (torque switches, limit switches, wiring and motors). A qualification assessment/analysis will be performed and components replaced with previously qualified items, as necessary. This equipment will be qualified when the analysis and component replacements are completed.

Section 3.90 VALVE ACTUATOR TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY 77 L imi torque II.a Documented evidence of qualification adequate SHB-0 Adequate similarity between equipment and test specimen established Aging degradation evaluated adequately 0uallfied life or replacement schedule established (if required)

RESOLUTION:

Suarnary: Details:

This equilxn.nt item was deficient with respect to qualification The TER concluded that valve actuator 80-118 was deficient in documented documentation, similarity, aging and qualified life. evidence of qualification. Additional deficiencies included similarity between A field inspection was performed to identify components. installed and tested equipment, aging and qualified life. We stated an Wyle Report 17655-HOV-3.1 (Reference 164) reviewed the quali- ongoing qualification assessment would be performed. Subsequent to the TER, a fication status of this equipment and determined that the actuators field inspection of the valve actuator was performed. Based were not tested to envelope plant specific conditions. on motor information obtained from this inspection, Wyle Report 17655-HOV-Additional analysis or replacement will be performed. 3.1 concluded that Limitorque actuators of the type used on valve 80-118 were This valve actuator will be qualiFied after qualification not tested to the maximum steam temperature/pressure conditions.

analysis or replacement is completed. Additional analyses will be performed to determine if existing qualification test results can be extrapolated to our service conditions. If the qualification cannot be achieved through analysis, the valve actuator will be replaced with qualified equipment. Upon completion of the analysis or replacement of this equipment (if necessary), this equipment will be qualified.

Section 3.9l VALVE ACTUATOR TER NO. HANUFACTURER/HOOEL NRC CATEGORY DEF ICIENCY 78 Limitorque II.a Documented evidence of qualification adequate SHB-000 (TER Item 78);

SHB-00 RESOLUTION:

Sumaary: Details:

Inadequate qualification documentation was provided for TER The NRC review concluded that documented evidence of qualification was inade-Item 78. quate for TER Item 78 valve actuators (IV05-05 and 05-07). At that time, Additional equipment was added to the master list after the TER Me stated that an ongoing qualification assessment would be performed. Sub-review. sequent to the TER, we added the following Limitorque valve actuators to the This equipm nt is located in the reactor building and is essentially equipment master list based on a detailed systems review:

exposed only to radiation during an accident. Plant 10 No. Hodel TER Deficiencies Wyle Report 17655-HOV-4.1 (Reference 167) and Limitorque Report No. 80058 (Reference 165) estab)ish qualification.

K)II=61 IVBO-02 SHIII)6 TER No.

HIA 'TTA SHBOO N/A N/A An analysis will be perforaed to determine qualified life. IVBO-21 SHBOO N/A N/A Upon completion of the analysis, this equipment wi 11 be qualified. IVBO-22 SHBOO N/A N/A These valve actuators are located in open areas of the Reactor Building where limiting temperature/pressure conditions are less than 110/1 psig due to line breaks; radiation is essentially the only harsh environment. Subsequent to the TER, field inspections of actuator nameplates were performed to identify motor manufacturer and insulation class. Hyle Report 17655-HOV-4. 1 esta-blished qualification for this equipment based on similarity of Nine Hi le Point 1 actuator models and motor s to equipment tested by the manufacturer and reported in Limitorque Report 80058, Appendix O. An aging analysis will be performed to determine a qualified life for this equipment. Upon completion of the qualified life analysis, this equipment will be qualified.

Section 3.92 VALVE ACTUATOR TER NO. HANUFACTURER/HODEL NRC CATEGORY DEFICIENCY 73, 74, 76 Limitorque Il.a Documented evidence of qualification adequate SHB-000 (TER Items 73, SHB-0 (TER Item 74) 76); Adequate similarity between equipment test specimen .established Aging degradation evaluated adequately Itualified life or replacement schedule established (if required)

RESOLUTION:

Sumnary: Details:

This equiprent was deficient with respect to qualification The NRC concluded that valve actuators were deficient with respect to documentation, similarity, aging and qualified life. documented evidence of qualification, similarity, aging degradation A field inspection was performed to identify components. Hyle evaluation and qualified life. Me stated at the time of the TER Report )7655-HOV-5.1 and 17655-HOV-6.1 (Reference 167 and review that an ongoing qualification assessment would be performed.

172) and Limitorque Report 80058 (Reference 165) establish This equipment is located in the reactor building where (peak) temperature/

qualification. pressure is 212F/1 psig and radiation TID is 1 Hrad under accident conditions.

An analysis will be performed to determine qualified life. Subsequently, field inspections were performed to identify actuator motor Upon completion of the analysis, this equipment will be qualified. insulation system class and manufacturer (i.e., Peerless or Reliance). Wyle Report 17655-HOV-5.1 established qualification for TER Items 73, 76; Wyie Report 17655-HOV-6.1 established qualification for TER 74. 0ualification was based on similarity of installed actuator models and motors to equipment tested by Limitorque. These test results are documented in Limitorque Report 80058. An aging analysis will be performed to determine a qualified life for this equipment. On completion of the qualified life analysis, this equipment will be qualified.

Section 3.93 VALVE ACTUATOR TER NO. MANUFACTURER/HODEL NRC CATEGORY DEFICIENCY 4, 5, 9 Limitorque Ii.a for TER Item 5 Documented evidence of qualification adequate SHB 11.c for TER Items 4 and 9 Adequate similarity between equipment test specimen established Aging degradation evaluated adequately Iiualified life or replacement schedule established (If required) I RESOLUTION:

Sugary: Octa i ls:

TER Items.4 and 9 were deficient with respect to aging and Limitorque valve actuators addressed in this section are listed be-qualif ied life. low, including TER deficiencies at the time of the NRC review for TER Items 4, TER Item 5 was deficient with respect to documented evidence of 5 and 9. The additional Limitorque valve actuators listed in the table were qualification, similarity, aging and qualified life. added to the master list based on a detailed systems review.

Additional valve actuators were added to the master list based on a detailed systems review.

Specific qualification documentation has not been identified for 5lllm6 SHBOO llWlr55 I V40-06

~

Hodel No; Plant ID No. TER No; 4 Aging, TER-Defici~enc Relete~no TR scope qualified life of -1005050.49 this equipment at this time. SHBOO I VBI -Ol N/A N/A Field inspections and evaluations wi)l be performed to determine SHBOO I VBI-02 N/A N/A qualification status based on similarity to components tested in SHBOO I V81-21 N/A N/A Limitorque qualification reports. SHBOO IV81-22 N/A N/A After completion of field inspections, evaluations and replacement SMB3 IV40-30 N/A N/A of components, as necessary, this equipment will be qualified. SHB3 IV40-31 N/A N/A SHBO BV93-27 5 Evidence, similarity, aging, qualified life SHBO . BV93-28 5 Evidence, similarity, aging, qualified life SHBO BV93-26 5 Evidence, similarity, aging, qualified life SHBO BV93-25 5 Evidence, similarity, aging, qualified life SHBO IV93-49 5 Deleted, not in scope of 10CFR50.49 SHBO IV93-50 5 Deleted, not in scope of IOCFR50.49 SHBOOO IV201-07 9 Aging, qualified life SHBOOO IV201-09 9 Aging, qualified life SHBOOO IV201-17 9 Aging, qualified life SHB2 IV31-08 N/A N/A SHB2 IV31-07 N/A N/A SHB2 IV38-02 N/A N/A

~Contfnue8)'ALVE Section 3.93 ACtUATOR Oetai1s: (Continued)

Specific qualification documentation for this equipment has not been identified at this time. Field inspections will be performed to determine the components used in equipment installation in Nine Nile Point l. An assessment will then be made of the qualification status based on similarity of installed components to Limitorque components tested fn existing Limitorque qualification reports. These actuators are scheduled for inspection to determine specific actuator components (torque switches, limit switches, wiring and motors). A qualification assessment/analysis will be performed and components replaced with previously qualified items, as necessary. This equipment will be qualified when the field inspections, analysis and component replacements are completed.

Section 3.94 TRANSM ITTER TER NO. MANUFACTURER/MODEL NRC CATEGORY DEFICIENCY 23 GE/MAC I.b Documented evidence of qualification adequate 551 RESOLUTION:

Surnnary: Details:

This TER item was deleted frcm the master equipment list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of 10CFR50.49. detailed systems review. This information was previously transmitted in our May 20, 1983 response to 10CFR50.49.

0 Section 3.95 TRANSMITTER TER NO. MANUFACTURER/MODEL NRC CATEGORY DEFICIENCY 24 GE/MAC I.b Documented evidence of qualification adequate 551 RESOLUTION:

Sumnary: Details:

This TER item was deleted from the master equipment list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of 10CFR50.49. detailed systems review. This information was previously transmitted 'in our May 20, 1983 response to 10CFR50.49.

Section 3.96 TRANSHITTER TER NO. HANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY 26 GE/HAC I.b Documented evidence of qualification adequate 551 RESOLUTION:

Sumnar y: Details:

This TER item was deleted. from the master equipm'nt list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of 10CFR50.49. detailed systems review. This information was previously transmitted in our-Hay 20, 1983 response to 10CFR50.49.

Section 3.97 TRANSMITTER TER NO. MANUFACTURER/MODEL NRC CATEGORY DEF lCIENCY 31 Rosemoun t il.a Documented evidence of qualification adequate 1151DP RESOLUTION:

Sumnary: Details:

This TER item was deleted from the master equipment list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of 10CFR50.49. detailed systems review. This information was previously transmitted in our May 20, 1983 response to 10CFR50.49.

Section 3.98 PRESSURE SMITCH TER NO. HANUFACTURER/HODEL NRC CATEGORY OEFICIENCY 33 Hercoid Ii.a Oocumented evidence of qualification adequate 508136 RESOLUTION:

Suaeary: Details:

This TER item was deleted from the master equipment list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of 10CFR50.49. 'etailed systems review. This information was previously transmitted in our Hay 20, 1983 response to 10CFR50.49.

Section 3.99 PRESSURE SWITCH TER 110. HANUFACTURER/HODEL NRC CATEGORY DEF ICIENCY 34 Hercoid Ii.a Documented evidence of qualification adequate DA5432 RESOLUTION:

Sunmary: Details:

This TER item was deleted from the master equipment list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of IOCFR50.49. detailed systems review. This information was previously transmitted in our Hay 20, 1983 response to 10CFR50.49.

Section 3.100 TRAHSMITTER TER HO. MN(UFACTURERJMOOEL HRC CATEGORY OEF ICIEHCY 40 Ros emoun t II.a Documented evidence of qualification adequate 115IOP RESOLUTIOH:

Suamary: Details:

This TER item was deleted from the master equipment list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of 10CFR50.49. detailed systems review. This information was previously transmitted in our May 20, 1983 response to 10CFR50.49.

Section 3.101 THERHOCOUPLE TER NO. HANUFACTURER/HODEL NRC CATEGORY OEF I CIENCY Omega . II.a Documented evidence of qualification adequate HWANSA223)2DN))4T834 RESOLUTION:

Surnnary: . Details:

Th)s TER item was deleted from the master equipment 1)st, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of 10CFR50.49. detailed systems review. This information was previously transmitted in our Hay 20, 1983 response to 10CFR50.49.

Section 3.102 HOTOR SQARgg TER NO. HANUFACTURER/HODEL NRC CATEGORY OEFICIENCY 98 GE Il.a Documented evidence of qualification adequate CR 2078 223AAA RESOLUTION:

Suaeary: Details:

This TER item was deleted from the master equipment list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of IOCFR50.49. detailed systems review.

Section 3.103 CIRCUIT BREAKER TER NO. MANUFACTURER/HOOEL NRC CATEGORY OEF ICIENCY 98 GE Oocumented evidence of qualification adequate NTE20 RESOLUTION:

5uranary: Oetai ls:

This TER item was deleted from the master equipment list, since This TER item was deleted from the master equipment list based on a it does not fall within the scope of IOCFR50.49. detailed systems review.

Section IV Justification for Continued Operation

APPENDIX A 5 B JUSTIFICATION FOR CONTINUEO OPERATION Each equipment item lacking documented qualification was evaluated and a justification for continued operation (JCO) was prepared. Even where an item was fully qualified except that an elastomeric seal or other degradable component had exceeded its rated lifetime (and therefore was scheduled for maintenance), a justification for continued operation was prepared. These justifications for continued operation verify that no significant degradation of any safety function results from failure of the item or items. In the case of display instruments, it was determined on an individual instrument basis, that there were sufficient alternate indicators available to the operator to prevent misleading in the event that a display instrument should fail.

Appendix A:

Justification for continued operation narratives (JCO's) prepared since our Nay 20, 1983 submittal for equipment environmental qualification are provided in Appendix A.

Appendix B:

Justification for continued operation narratives (JCO's) submitted as a part of the iMay 20, 1983 submittal are summarized in Appendix B. These have been resubmitted at this time in order to provide complete documentation covering all equipment items..

APPENDIX A

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Temperature elements located in the emergency condenser isolation valve cubicle, elevation 281 feet (TE IB08-26-11, 12, 13, 14, 15, 21, 22, 23 5 24)

FRC Equipment Item No.: 32 Manuf actur er: Minco Nickel Model: R-T-0 Saf et Functi on: Alerts operator to leak in emergency cooling system loop Oualification Oeficienc: Oocumented evidence of qualification Justification for Continued Operation:

These temperature elements are installed to aid the plant operator in identifying steam leakage from one or the other of the emergency cooling system loops to permit manual isolation of the faulted loop (See Operating Procedure Nl-OP-13). Failure of these temperature elements could either ( 1) prevent timely isolation- of a leaking emergency cooling loop (7ailure to operate) or (2) cause unnecessary isolation of one or both emergency cooling loops by the operator (actuation not caused by emergency cooling loop leaking). In either event, the emergency cooling system is backed up by the high pressure coolant injection system, which is located in a mild environment under conditions when the harsh environment is in the reactor building.

Furthermore, the core spray/automatic depressurization systems are also available to provide core cooling functions should the emergency cooling system be either unavailable or a source of leakage.

Therefore, justification for the continued safe operation of the plant is demonstrated. I

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Temperature elements located adjacent to the emergency condensers (TEIG01 A, 8, C 8 0),

FRC Equipment Item No.: N/A Manufacturer: Pall Trinity Model: CU/6 Safety Function: Identifies possible emergency condenser tube leak or condenser low level condition

/uglification Deficiency: Documented evidence of qualification Justification for Continued Operation:

These temperature elements monitor emergency condenser shell-side temperature. High shell side temperature indicates either insufficient shell side water inventory or possible condenser tuoe leak. Consequently, malfunction of these detectors can lead to (1) reduced performance of the emergency cooling system (due to low condenser water level) or (2) premature condenser isolation by the operator (due to faulty indicators). In either event, the emergency cooling system is backed up by the high pressure coolant injection system (wnicn would be in a mild environment because the harsh environment is in the reactor building) and also by the core spray/automatic depressurization system.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW SUMNRY SHEET

~Equi ment: Thermocouple located in the reactor building (TE 7O-23)

FRC Equipment Item No.: 22 Manufacturer: Pall Trinity Model: CU/6 Safety Function: Provides temperature signal for automatic control of reactor building closed loop cooling water and service water flow.

/uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

This temperature element monitors reactor building closed loop cooling heat-exchanger outlet temperature and provides an alarm in the control room.

The temperature signal is also sent to,an E/P converter (E/P 70-137) which controls heat-exchanger flow both on the reactor building closed loop cooling side and the service water side. Failure of either the temperature element or E/P converter can result in increased heat-exchanger flow (both reactor building closed, loop cooling and service water), no change in flow, or decreased flow.

If flow is increased, excess cooling will be applied to the system loads.

This is not desirable from a moisture and equipment life standpoint but is not an imminent threat to continued system operation. If flow remains as-is, the system will continue to operate essentially normally, particularly since reactor building closed loop cooligg heat loads will remain fairly constant even with a high energy line break in the reactor building.

If cooling flow is greatly decreased or lost, however, the reactor building closed loop cooling loads may eventually be lost. During a high energy line break, the systems of concern which rely on reactor building closed loop cooling are high pr essure coolant injection, instrument air, and H2-02 monitoring. Even if all three of these systems are lost, the plant can still be placed in a safe condition, as follows.

In the case of loss of nigh pressure coolant injection, both the emergency cooling system and the core spray/automatic depressurization system are available for core cooling. The emergency cooling system will start automatically, either on high reactor pressure or low-low reactor water level, or due to valves isolation valves 39-05/06 failing-open on loss of instrument air . With instrument air unavailable, the cool down rate can be controlled by the emergency cooling system by cycling motor-operated valves in the steam supply lines to the condenser.

~ .

With core. cooling assured. by the. emergency cooling system and the high energy line break located in the reactor building (causizg the harsh environment to reactor building closed loop cooling equipment), the remaining instrument air loads and the Hp-Op monitoring system are not required for mitigation of this accident.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid valves located in the emergency condenser return valve area (SV/IV39-05, 06; SV39-05E, 05F, 06E, 06F)

FRC Equipment Item No.: 17 Manufacturer: Asco/Valcor Model: 8300 Series/V70900-21-1 Safety Function: De-energize to open emergency condenser return valves Oualification Oeficiency: Similarity, aging, qualified life, radiation, functional testing Justification for Continued Operation:

These solenoid valves are normally energized to maintain air pressure to the normally closed emergency condenser air-operated isolation valves (39-05,

-06). On reactor protection system signal (high reactor pressure or low-low reactor pressure level, either for 10 seconds) or on loss of electrical power, these solenoids de-energize and vent the air to isolation valves 39-05 and 06, allowing the isolation valves to open and natural circulation to begin through the condensers.

e The emergency condensers are used to assist in core cooling and depressurization only if the nigh pressure coolant injection system is unavailable. In this case, emergency cooling will oe initiated at the start of an accident, after whicn, failure of the solenoids is of no concern.

Furthermore, it is hard to conceive of a credible fai lure mecnanism wnich will Also, if keep the solenoids energized when they are signaled to de-energize.

necessary, power to the solenoids can be manually secured.

Finally, if the emergency cooling system cannot be initiated (either loop),

core cooling is assured by the hign pressure coolant injection system and/or core spray/containment spray/automatic depressurization system.

Therefore, justification for the continued safe operation of the plant is

'demonstrated.

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Pressure transmitter located in the reactor building (PT36-07A, B, C, 0;36-08A, B, C, 0)

FRC Equipment Item No.: 27 Manufacturer: Rosemount blode 1: 1151 Safety Function: Initiates protective action on high reactor vessel pressure Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued 0 eration:

These transmitters monitor reactor pressure and initiate protective action on high reactor vessel pressure. Ouring a loss-of-coolant accident or high energy line break, protective action occurs nearly immediately. Consequently, instrumentation devices whicn initiate protective actions will have performed their intended safety function before being exposed to the adverse environment for more than just a few .seconds.

The transmitters are located in the reactor building where any credible high energy line break can be isolated from the reactor vessel in a short period of time. Consequently, the extent and duration of the adverse environment wi 11 be limited and it is likely that these transmitters wi 1,.1 be available for long-term post-accident monitoring, as well as performing their short-term safety function. Furthermore, should all reactor pressure indication fail, the reactor vessel remains protected from an over-pressure condition by relief-valves. Consequently, eventual loss of these transmitters does not seriously affect accident mitigation.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW SUMNRY SHEET

~Equi ment: O/P transmitter located in the reactor building (FT 36-06A, 8, C, 0)

FRC Equipment Item No.: 29 Manufacturer: Rosemount Model: 1151 OP Safety Function: Monitors steam line break in emergency condenser cooling loops

/uglification Oeficienc : Oocumented evidence of qualification Justification for Continued Operation:

These transmitters monitor the emergency condenser cooling loops. They will detect a break in the cooling loop as indicated by high differential pressure

{i.e. high flow) and signal the reactor protection system to isolate the appropriate loop.

In the case of a high energy line break in an emergency condenser cooling loop, the high differential pressure condition will occur nearly immediately.

Consequently, these transmitters will perform their safety function prior to the on-set or sustained adverse environmental conditions. Furthermore, should the. transmitters fail to perform their function, other indicators, such as hign area temperature alarms and high radiation detectors, are availaole to identify the break. In any event, core cooling is assured by the operable emergency cooling loop, the high pressure coolant injection system, and/or core spray/containment spray/automatic depressurization system.

Therefore, justification for the continued safe oper ation of the plant is .

demonstrated.

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Flow transmitter located in reactor building (FT RV-26A, 26B)

FRC Equipment Item No.: 39 Manufacturer: Rosemount - GE/MAC Model: 11510P - 553 Safety Function: Core spray flow measurement

/uglification Deficiency: Documented evidence of qualification Justification for Continued Operation:

This flow detector monitors core spray flow. Core spray flow detectors provide no control or automatic safety functions. They provide low flow rate indication to the operator. There are numerous other indicators of this system, such as filter differential pr essure indicators, to aid the operator should the flow transmitters fail. Failure of these transmitters does not impair system performance.

Therefore, justification for the continued safe operation of the plant is demonstrated.

~l NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Level transmitters located in the reactor building (LT 58-05, 58-06)

FRC Equipment Item No.: 80 Manufacturer: Rosemount iMode1: 1)53OA Safety Function: Monitors torus water level

(}uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Torus water level transmitters provide no automatic control or safety function. They provide level alarms indications to the operator. Water level in the torus at the start of an accident would be sufficient for accident .

mitigation. Should the level transmitters fail post-accident, there is no reason to suspect subsequent loss of water invento'ry in the torus. Loss of these transmitters will not adversely affect accident mitigation.

Therefor e, justification for the. continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Flow transmitters located in the reactor building (FT 80-49A, 56A, 71A, 76A and 93-30A, 32A, 33A, 34A)

FRC Equi ment Item No.: N/A Manufacturer: GE/MAC Model: 553 Safety Function: Monitor containment spray flow Qualification Deficiency: Documented evidence of qualification Justification for Continued Operation:

These flow transmitters monitor containment spray flow to each of four containment spray heat exchangers (torus water and.raw water sides). They perform no automatic control or safety functions but provide indication of flow to the operator. In view of the numerous other indications of flow availaole to the operator (e.g. pumps running, filter differential pressure, temperature at heat exchanges, etc.); loss of these instruments will not adversely affect accident mitigation.

Therefore, justification for the continued safe operation of the plant is demonstrated.

10

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Level transmitters located in the reactor building (LT IG06A, B and LT60-22, 23)

FRC Equipment Item No.: N/A Manufacturer: GE/MAC - Rosemount Model: 553 - 1153DA5 Safety Function: Provides for automatic control of emergency condenser level.

(}uglification Deficiency: 'ocumented evidence of qualification Justification for Continued Operation:

These transmitters monitor emergency condenser water level and provide for automatic control of the make-up supply valves. Should these transmitters fail, emergency condensers can be filled by opening the make-up supply valves (EC 111-112 LCV or EC 121-122 LCV) and .filling the emergency condensers manually.

(Note: The make-up supply valves fail open on loss of air, therefore the condensers can always be filled, either to overflowing or based on pre-determined amounts of make-up).

justification for the continued safe operation of the plant is

'herefore, demonstrated.

C NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Flow transmitters located in the turbine building (FT 202-92A and 49A)

FRC Equi ment Item No.: N/A Manufacturer: GE/MAC Model: 554 Safety Function: Control of exhaust flow from reactor building emergency ventilation system

/uglification Oeficiency: Oocumented evidence of qualification Justification for Continued 0 eration:

These transmitters monitor exhaust air flow from the reactor building emergency ventilation system and control the inlet damper to the exhaust fans to automatically regulate flow. Failure of the transmitters could result in loss of reactor building emergency exhaust.

Should reactor building emergency exhaust flow be lost due to transmitter failure, the inlet dampers can either be manually controlled or failed open (by isolating instrument air) in order to regain emergency ventilation flow.

Consequently, transmitter failure does not adversely impact accident mitigation.

In addition, indication of exhaust air flow is available from other sources (e.g. filter differential pressure, fan operation/valve line-up, etc.)

Therefore, justification for the continued safe operation of the plant is demonstrated.

12

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Flow transmitters located in the reactor building (FTRO-15)

FRC Equipment Item No.: N/A Manufacturer: GE/MAC Model: 553 Safety Function: Provides signal to regulate control rod drive hydraulic system flow to a constant 65 gpm.

Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued 0 eration:

The control rod drive hydraulic system provides water for charging scram accumulators, for normal rod movement, for control rod drive cooling, and also provides a small amount of high pressure coolant injection, which provides core cooling for certain primary system leakage or line breaks up to 0.003 square'feet.

This flow transmitter monitors control rod drive hydraulic system flow and provides the input signal to two E/P converters which in turn operate control rod drive flow control valves. These valves maintain control rod drive hydraulic system flow at 65 gallons per minute.

Should the transmitter fail, flow control could be lost. However, with the scram accumulators charged at the start of an accident, the requirement to, constantly charge them terminates. Once the scram occurs, normal rod motion and control rod drive cooling are also unneeded. Finally, any hign pressure injection function can be provided by the high pressure coolant injection system, which is also backed-up by core spray/automatic depressurization system.

Therefore, justification for the continued safe operation of the plant is demonstrated.

13

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Pressure transmitter located in reactor building (PTID-46A and 468)

FRC Equipment Item No.: 25 Manufacturer: GE/MAC Model: 551 Safety Function: Provides indication of reactor vessel pressure.

/uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

These pressure transmitters provide control signals to the feedwater control system and also provide indication of reactor vessel pressur e for the operator. It is expected that these transmitters wi 11 operate in a post-accident condition because they are located in the reactor building, where radiation is the only adverse parameter for an accident within containment or because a high energy line break in the reactor building can be quickly terminated. Nevertheless, post-accident failure of these transmitters is acceptable for the following reasons:

1. Short-term protective functions will oe performed by separate and independent pressure transmitters.

2. Automatic feedwater control is not required to mitigate the accident.

3. Once core spray has been initiated, vessel pressure will essentially be core spray system head and vessel pressure is not a critical parameter at this time.

Therefore, justification for the continued safe operation of the plant is demonstrated.

14

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Level transmitter located in reactor building (LTIA-12)

FRC Equi'pment Item No.: 38 Manufacturer: GE/MAC lloae 1: 553 Safety Function: Provides indication of reactor vessel level.

Qualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

This level transmitter provides control signals to the feedwater control system and also provide indication of reactor vessel level for the operator.

It is expected that transmitter will operate inwhere a post-accident condition because it is located in the reactor building, radiation is the only adverse parameter for an accident within containment or because a nigh energy line break in the reactor building can be quickly terminated. Nevertheless, post-accident failure of the transmitter is acceptable for the following reasons:

1. Short-term protective functions will be performed by separate and independent level transmitters.

2. Automatic feedwater control is not required to mitigate tne accident.

3. Once core spray has been initiated, the core will be cooled as long as core spray flow is maintained.. Reactor vessel level is not a significant parameter in this operating configuration.

Therefore, justification for the continued safe operation of the plant is demonstrated.

15

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Electric motors located in the reactor building elevation 198 feet (M-81-23, 24, 03, 04)

FRC Equipment Item No.: 44 Manufacturer: GE Model: 5K6336XC-166A Safety Function: Orives core spray pumps gualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Core spray pump motor drives are located on the 198 foot level of the reactor building where the limiting post-accident environment, except for radiation, is not highly stressful on equipment (100F, 1 psig, 1 x 106 Rads). In addition the two core spray loops are physically separated, therefore at least one of the 100 percent capacity loops will be in an even less stressful environment during a specific accident. Consequently, it is expected that one, if not both, core spray loops will remain operational throughout any post-accident environment.

In the unlikely event of loss of all core spray pumping capacity, core spray flow can be maintained by cross-connecting the core spr ay discharge headers with the containment spray raw water system. This cross-connect permits pumping of raw lake water directly from the intake tunnel. to the core using raw water pumps which are located in a mild environment.

Even in the worst possiole scenario, wnere the pipe break exists in the reactor recirculation piping, core flooding/cooling using raw water would continue until containment water volume reaches a level wnere the core remains covered without continuous raw water addition. Long-term cooling can then be performed by the emergency condensers, or even by the shutdown cooling system.

Therefore, justification for the continued safe operation of the plant is demonstrated.

16

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Electric motors located in the reactor building elevation 237 feet (M-81-49, 50, 51, 52)

FRC Equipment Item No.: 42 Manufacturer:

Model: 5K828837C7 Safety Function: Orives core spray topping pumps gualification Oeficiency:

S Similarity, qualified life, pressure Justification for Continued Operation:

Core spray topping pump motors are located on the 237 foot level of the reactor building wher e the 1imiting post-accident environment, except for radiation, is not highly stressful on equipment (126F, 1 psig, 1 x 106 Rads). In addition the two core spray loops are physically separated, therefore at least one of the 100 percent capacity loops will be in an even less stressful environment during a specific accident. Consequently, it is expected that one, if not both, core spray loops will remain operational throughout any post-accident environment.

In the unlikely event of loss of all core spray pumping capacity, core spray flow can be maintained by cross-connecting the core spray discharge neaders with the containment spray raw water system. This cross-connect permits pumping of raw lake water directly from the intake tunnel to the core using raw water pumps which are locted in a mild environment.

Even in the worst possible scenario, where the pipe break exists in the reactor recirculation piping, core flooding/cooling using raw water would continue until containment water volume reaches a level wnere the core remains covered without continuous raw water addition. Long-term cooling can then be performed by the emergency condensers, or even by the shutdown cooling system.

Therefore, justification for the continued safe operation of the plant is demonstrated.

I NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Electric Motors Located in the reactor building Elevation 198 feet (M-80-03, 04, 23, 24)

FRC Equipment Item No.: 43 Manufacturer:

Model: 5K6328XC-136A Safety Function: Drives containment spray pumps Oualification Deficiency: Documented evidence of qualification Justification for Continued Operation:

Containment spray pump motor drives are located on tne 198 foot level of the reactor building where the limiting post-accident environment, except for radiation, is not highly stressful on equipment (100F, 1 psig, 1 x 106 Rads). In addition the two containment spray loops are physically separated, therefore at least one of the 100 percent capacity loops will be in an even less stressful environment during a specific accident. Consequently, it is expected that one, if not both, containment spray loops will remain operational throughout any post-accident environment.

In the unlikely event of loss of all containment spray pumping capacity, containment spray flow can be maintained by cross-connecting the containment spray discharge headers with the containment spray raw water system. This cross-connect permits pumping of raw lake water directly from the intake tunnel to the spray header using raw water pumps which are located in a mild environment.

Therefore, justification for the continued safe operation of the plant is demonstrated.

18

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Electric Motors Located in the reactor building Elevation 237 feet (NC 08A -and NC 08B)

FRC Equi ment Item No.: N/A Manufacturer:

Model: 5K814316A73 Safety Function: Orives control rod drive hydraulic pumps Qualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

The control rod drive hydraulic system motor drives are located on the 237 foot level of the reactor building where the limiting post-accident environment, with exception of radiation, is not highly stressful on equipment (126F, 1 psig, 1 x 106 Rads). Consequently, it is expected that tnese motors will function in a post-accident environment.

The control rod drive hydraulic system provides water for charging scram accumulators, for normal rod movement, for control rod drive cooling, and also provides a small amount of nigh pressure coolant injection which provides core cooling for certain primary system leakage or line breaks up to 0.003 square feet.

Control rod drive hydraulic functions are not required once an accident has occurred. The scram accumulators are charged at the start of the accident and therefore they will function. Once the scram occurs, normal rod motion is irrelevant. Finally, any high pressure injection needs wi.ll be accommodated by the high pressure coolant injection system.

Therefore, justification for the continued safe operation of the plant is demonstrated.

19

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Electric motors located in the reactor building (M202-33, 202-53)

FRC Equipment Item No.: N/A Manufacturer:

Model: 5K184AL 218 Safety Function: Orives for reactor building emergency ventilation exhaust fans Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Reactor building emergency exhaust fans function following an accident inside Containment or the reactor building in order to filter exhaust air wnile maintaining a negative pressur e in the reactor building relative to the outside atmosphere. Since they are physically located in the turbine building, the only harsh environmental parameter to which these. motors are exposed in an accident condition for which they must operate is radiation.

Furthermore in an accident, the radiation level is mostly caused by the air which the fans are exhausting. Therefore under non-accident conditions, the fans are not subjected to any significant radiation levels.

Consequently, it is expected that these fans will function normally under accident conditions. This is particularly true since the radioactive air being exhausted wil.l be greatly reduced once post-accident containment pressure is reduced or the leakage into the reactor building has been terminated.

Therefore, justification for the continued safe operation of the plant is demonstrated.

20

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Electric motors located in the reactor building elevation 298 feet (M 70-01, 02, 03)

FRC Equipment Item No.: 41

.Manufacturer:

Node1: 5K445AK249A Safety Function: Drives reactor building closed loop cooling water pumps

/uglification Deficiency: Documented evidence of qualification Justification for Continued Operation:

These motors drive the three 50 percent capacity reactor building closed loop cooling pumps. Normally, one or two of the thr ee pumps is in operation and under accident conditions, one pump can supply the cooling loads with accident mitigation functions. The motors are located in the reactor building where the only harsh environmental parameter during an accident inside containment is radiation. During a high energy line break in the reactor building, the motors may be exposed to a.temperature of 300F but this temperature can be quickly lowered by isolating tne high energy line break at the containment boundary. Consequently, it is considered highly unlikely that all thr ee motors would fail under these conditions.

Should all reactor building closed loop cooling flow be lost following an accident, the following cooling loads with accident mitigation functions may be jeopardized: high pressure coolant injection, instrument air, H2-02 monitoring. In case of loss of these systems, the following backups are avai able:

1 Load Backup High Pressure Core spray/automatic depressurization system Coolant Injection or emergency cooling System Instrument Air Essential components operated oy instrument air fail to their safety position. Motor-operated valves can oe used to control the cooldown in the absence of air-operated valves.

H2-02 Monitoring Oxygen content can be maintained below 4 percent by excess inerting of the containment with ni tr ogen.

Therefore, justification for continued safe operation of the plant is demonstrated.

21

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Valve Actuator located inside the drywell (IV38-Ol and 38-13)

FRC Equipment Item No.: N/A Manufacturer: Limitorque Model: SMB2 Isolation of shutdown cooling system

(}uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Under normal plant operating conditions, the shutdown cooling system is not in operation and its drywell isolation valves are closed. Should these valves be open when an accident occurs, isolation valves 38-01 and IV38-13 close to isolate the drywell. Since isolation occurs at the start of an accident, it is expected that the valves will function. If they should be open and if they should fail to close, isolation is still performed by the valves located outside the drywell (38-02, a motor operated valve and 38-12, a check valve).

Therefore, justification for the continued safe operation of the plant is demonstrated.

22

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid valves located in the reactor building (SV68-08C,68-09C, 68-10C)

FRC Equipment Item No.: 18 Manufacturer: Asco Model: WPLB 8300 868F Safety Function: Operates torus vacuum relief valve Oualification Oeficiency: Similarity, aging degradation, qualified life, aging simulation Justification for Continued 0 eration:

The torus vacuum relief valves ar e normally closed valves which open to prevent a vacuum in the torus with respect to outside atmosphere. The solenoid valves are energized to keep the relief valves normally shut. On loss of electric power or air, the relief valves fai 1 open. Containment integrity is not br eached with a relief valve open because of check valves in each relief line.

The solenoid valves are located in the reactor building wnere the only harsh environmental parameter during an in containment accident is radiation. Since they will operate early in an accident, they are expected to function. Should one or more of the solenoids fail, it is most likely that the solenoid will deenergize, causing the relief valve to go to its safety position, open.

Therefore, justification for the continued safe operation of the plant is justified.

23

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid valves located in the reactor building (SOV 40-328,40-32C, 40-33B,40-33C)

FRC Equipment Item No.: N/A Manufacturer: Asco Model: HT 8320A90 Safety Function: Closes the core spray high point vents, on high drywell pressure or reactor low-low level gualification Deficienc  : Documented evidence of qualification Justification for Continued Operation:

Isolation valves 40-32 and 40-33 are air-operated isolation valves in the 1 inch vent lines from the high points of the core spray headers. These valves are normally shut and are opened only to vent the core spray line to the equipment drain system. Each air -operator is controlled by two solenoid valves in series (40-32B and C or 40-338 and C). Both solenoids must be energized simultaneously for the air to open the isolation valve.

Should an isolation valve be open when an accident occurs, the valve will automatically close on either a high drywell pressure or reactor low-low level signal. Since these signals wi 11 be received at the start of an accident, the solenoids are expected to operate. Furthermore, the only way that 40-32 or 40-33 will remain pressurized (i.e. open) is for a solenoid to fail in such a way that it either remainS energized or sticks. Finally, should 40-32 or 40-33 remain open, motor-operated valves inside the drywell (40-30 and 40-31) can still isolate the vent line.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid valves located in the turbine building (SV/BV-202-15, 16, 31, 32, 34, 35, 36, 37, 38, 74, 75)

FRC Equi ment Item No.: N/A Manufacturer: Asco Model: 8300C68/8300C68F Safety Function: Operation of dampers in reactor building ventilation supply, normal exhaust, and emergency exhaust Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

These solenoid valves control the air supply to air-operated dampers in the reactor building ventilation system as follows:

SV/BV202-15, 16 Ventilation supply discharge duct SV/BV202-31, 32 Normal exhaust inlet duct SV/BV202-34, 35 Emergency exhaust discharge ducts SV/BV202-36 Emergency exhaust common inlet duct SV/BV202-37, 38 Emergency exhaust inlet ducts SV/BV202-74, 75 Turbine building supply to emergency exhaust inlet duct Under accident conditions, these dampers are shifted to isolate normal exhaust and establish emergency exhaust, and also to isolate the normal supply and exhaust systems under high radiation conditions.

The solenoid valves are located in the turbine building, where the only harsh environment parameter with an accident in containment or the reactor building is radiation. Consequently, these solenoids are expected to operate, particularly at the start of an accident. In addition, the normal supply and exhaust isolation dampers are each controlled by two solenoid valves which have to fail to de-energize in order to prevent the dampers from shifting.

Therefore, justification for the continued safe operation of the plant is demonstrated.

.NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Solenoid valves located in the reactor building

( SV/I V05-1A/B,-2R, -3R, -4A/B, -11B, -12A)

FRC Equipment Item No.: N/A Manufacturer: Asco Model: NP8344A71 E Safet Function: Isolates emergency condenser vent line to main steam line

/uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

The emergency condenser steam lines are vented to the main steam lines in order to remove air and non-condensaole gases which could interfere with natural circulation. Under normal conditions, these air-operated vent valves are open. The solenoid valves control the air to the vent valves. Under accident conditions, the vent valves are snut and the emergency condensers are manually vented to the torus.

The solenoid valves are exposed to a harsh environment by a break outside the drywell in the emergency condenser piping. In this case, the emergency condenser cooling loop will be isolated and therefore the isolation of the 1 inch vent line is not of concern. For other high energy line breaks, the solenoids are in a much less severe environment and are expected to operate normally.

Therefore, justification for the continued safe operation of the plant is demonstrated.

26

NINE NILE POINT UNIT 1 COHPONENT REVIEW SUHHARY SHEET

~Equi ment: Position limit switches FRC Equipment Item No.: See Below Manufacturer: Namco/Fisher Controls Model: Various {see below)

Safety Function: Indication (and alarm) of valve position for the operator (uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

The limit switches listed below provide indication (and alarm) of isolation valve position. The switches provide no automatic control or safety related function, other than indication and alarm. By emergency procedure, the operator checks these indicators after isolation occurs. Since isolation occurs at the beginning of an accident, it is expected that the indicators will work properly. Should a switch indicate that a valve did not achieve its safety position, the operator would attempt to put it in its safety position or would verify that it actually was in its safety position. These position switches are:

Plant I.O. No. ~Mf r. Model Mo. FRC Mo. Valve Indication POS39-ll to 14 Fisher 304 N/A Emergency cooling drain line POS39>>05, 06 Namco SL3L 50 Emergency cooling condensate return POS01-05, 06 Namco 02400X 46 Hain steam by-pass POS40-32,33 iVamco EA170 N/A Core spray vent line POS68-02 to 07 Namco 02400X-2 N/A Torus vacuum relief POS80-15, 16,35,36 Namco 02400X N/A Containment spray discharge POS83. 1-10 Namco 02400X 45 Equipment drain line from drywell POS83. 1-12 Namco 02400X 45 Floor drain line from drywell POS201-08 Namco 02400XR 45 Torus air vent POS201-10 Namco 02400XR 45 Orywell air vent POS201-16 Namco 02400XR 45 N2 supply to condenser POS201-32 Namco 02400XR 47 iV2 vent and fill POS201.2-32,03 Namco 02400X 47 iV2 makeup and bleed POS201.2-33,06 Namco 02400X-2 45 N2 makeup and vent 27

Page 2 Plant I.O. No. ~Mf r. Model No. FRC No. Valve Indication POS201. 1-09, 11 Namco EA180 N/A N2 to emergency vent system POS201.2-109, 110, Namco EA170 N/A Containment atmospheric dilution 111, 112 sample lines POS201.7-08,09, 10, 11 Namco 02400X-2 N/A Turbine building sample sink POS58. 1-01 Namco 02400X N/A Condenser makeup to torus POS122-03 Namco EA170 N/A Post-accident sample line In addition to the above, each line from the drywell or torus is isolated oy at least two isolation valves, either check valves or other air-operated or motor-operated valves. Consequently, should an isolation valve fail to close and should its position indicator fail to correctly indicate valve position, the operator has other valves and indicators to check to ensure isolation has occurred.

Therefore, justification for the continued safe operation of the plant is demonstrated.

28

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Pressure indicating switches located in the reactor building (NR-108A to F)

FRC Equipment Item No.: N/A Manufacturer: Barksdale Model: 1539VX Safety Function: Operates electro-matic relief valves of the automatic depressurization system

/uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

The automatic depressurization system relief valves are installed to reduce reactor vessel pressure to allow core spray system operation for piping breaks smaller than 0.3 square feet. Three of the six installed valves are sufficient to provide the design blowdown. Relief valves are operated manually or au'tomatically upon simultaneous low-low-low reactor level and high drywell pressure.

Since a reactor vessel blowdown will occur at the start of a small-break loss of coolant accident (i.e. brief exposure to harsh evironment) it is expected that at least three of the six valves will function. Once reactor pressure has been reduced to permit core spray flow, the automatic depressurization system relief valves no longer perform an accident mitigating function. In addition, these switches are located in the reactor building and are in permanent enclosures which makes it'even more likely that they wi 11 operate.

Should a condition occur where automatic depressurization system was required (i.e. loss-of-'oolant-accident with pipe break less than 0.3 square feet) aqd should the system fail to achieve the necessary blowdown (i.e. at least four of six valves fail to open), adequate core cooling will be provided and reactor vessel pressure will be lowered by the high pressure coolant injection system, the emergency condensers, or a combination of the two. In addition, with the loss-of-coolant accident inside containment, the high pressure coolant injection system is essentially located in a mild environment (turbine building) and the emergency cooling system ui ll be subjected to high radiation only (reactor building).

Therefore, justification for the continued safe operation of the plant is demonstrated.

29

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Main steam line outboard isolation valve position switches {POS01-Ol to 01-04)

FRC Equipment Item No.: 49 4 100 Manufacturer: Namco Model: SL3C-58T-M Safety Function: Initiates reactor scram on main steam line isolation valve position

/uglification Oeficiency: Radiation Justification for Continued Operation:

The switch provides a signal to the reactor protection system for reactor scram on main steam line position. Failure of the switch to generate a reactor scram on valve position would result in increased reactor pressure and power and a decrease in indicated reactor coolant level, all of which generates reactor scram signals for off normal values of these parameters.

Mitigation of a high energy line break or loss of coolant accident is not dependent on these switches. These switches provide an anticipatory-reactor trip due to loss of normal heat sink, the main condenser.

Although time/temperature aging analysis currently indicates limited qualifications based on conservative switch would fail. Normal technical ambient temperatures, it is unlikely the specification surveillance testing provides a means to monitor switch operability for age related failures. (The switches are scheduled for replacement with qualified EA740 models.)

Therefore, justification for the continued safe operation of the plant is demonstrated.

30

I NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET ment: Solenoid valves located in the reactor building

~Equi controlling inert gas purge and fill valves FRC Equipment Item No.: 15/16 Manufacturer: Asco Model: Various (see below)

Safet Function: Operate to purge the drywell, torus, emergency ventilation system, and condenser

(}uglification Oeficienc  : Documented evidence of qualification, similarity, aging degradation, qua'lified life, aging simulation Justification for Continued Operation:

The following solenoid valves located in the reactor building control air operated 'valves which perform functions associated with purge and fill of the drywell, torus, and associated ventilation lines:

Valve Model No. Function SOYA . 2-02 83~BI F pp suppppy o drywell SOV201.2-04 8300B61F supp ly to torus SOV201. 9-91 HT8317A30 N2 supply to drywell SOV201.9-92 HT8317A30 N2 supply to torus SV/I V201. 2-06 WPLB8300B72F N2 supply to torus SV/IV201.2-33 8300861RU N2 supply to torus SV/IV201-08 WPLB8300872F Torus atmospheric vent SV/I V201-10 WPLB8300872F Torus atmospheric vent SV/I V 201-16 WPLB8300872F N2 supply to condenser SV/I V201. 1-09 NP8344A71E N2 supply to emergency Ventilation SV/I V201. 1-11 NP8344A71E N2 supply to emergency ventilation SV/IV201.2-03 8300861 N2 supply to drywell SV/IV201.2-32 8300861RU supply to drywell SV/I V201-32 WPHV202-302-1F N2 N2 vent and fill (drywell)

The containment atmospheric dilution system prevents the build-up of a combustible concentration of hydrogen and oxygen within the containment following a loss of coolant accident. The system functions by a combination of purging with nitrogen and venting to either the reactor building emergency ventilation system or the main condenser. Ouring an accident within the containment, radiation is the only harsh environmental parameter to which these valves are exposed. Since the maximum hydrogen and oxygen generation 31

occurs at the beginning of an accident, it is highly improbable that any of these valves will fai 1 due to radiation before the time when all vent and purge functions have been accomplished. In addition, several solenoid valves would have to fail in order to prevent the introduction of nitrogen to both the drywell and the torus and to prevent purging of the ventilation paths.

Therefore,. justification for the continued safe operation of the plant is demonstrated.

32

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Solenoid valves located in the reactor building

( SV/ I V80-15, -16, -35, -36)

FRC Equipment Item No.: 12 Manufacturer: Numatics Model: 46JLSA03, 48JLSA03 Safety Function: Open isolation valves to initiate containment spray

/uglification Oeficienc : Oocumented evidence of qualification Justification for Continued 0 eration:

One air-operated isolation valve (80-15,-16,-35 or-36) are located in each of the four containment spray discharge headers to the drywell. These valves are normally open and fail open on loss of air. Each isolation valve is controlled by a set of two normally de-energized solenoid valves. The isolation valves will remain open (their safety position) unless both of their respective solenoid valves fail to their energized positions.

Following a loss-of-coolant accident within the containment, the solenoid valves only need remain de-energized to keep the isolation valves in their safety position (open). Radiation is the only harsh environmental parameter to which the solenoids are exposed. The signals to energize the solenoids (to shut the isolation valves) come from the control room, which is in a mild environment. Cogsequently, there is no credible scenario by which radiation could result in the energizing. of both solenoids in a set with the subsequent closure of the respective isolation valve. Furthermore, three of the four isolation valves would have to shut (i.e. six of tne eight solenoids energized) before 100 percent capacity of the containment spray system would be lost.

Therefore, justification for the continued safe operation of the plant is demonstrated.

33

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Solenoid valves located in the reactor building (SV/IV 58.1-01)

FRC Equipment Item No.: N/A Manufacturer: Asco Model: WP8300861RU Safety Function: Isolation of condensate make-up line to torus Qualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Valve 58.1-01 isolates the torus make-up line from the condensate storage tank. Valve 58.1-01 is normally closed and remains closed during an accident. Should valve 58.1-01 nappen to be open at the start of an accident, the solenoid valves must de-energize in order to cause tne isolation valve (58.1-01) to close. Since the solenoid valve is outside of the containment where the only harsh parameter is radiation, the solenoids are expected to operate normally at the start of the accident. Furthermore, should ooth solenoids fail, isolation of the make-up line will be accomplished by the self-actuating check-valve 58. 1-02.

Therefore, justification for the continued safe operation of the plant is demonstrated.

34

NINE MILE POINT UNIT 1 COMPONENT REVIEH

SUMMARY

SHEET

~Equi ment: Solenoid valves located in the reactor building (SV/IVOl-03 and 01-04)

FRC Equi ment Item No.: N/A Manufacturer: Asco Model: WPHTX8300B61U Safety Function: Close main steam isolation valves 01-03 and 01-04 gualification 'Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Main steam isolation valves 01-03 and 01-04 are normally-open/fai 1-closed valves. At the start of an accident, they close to isolate the main steam lines. The solenoid valves which operate the main steam isolation valves merely have to de-energize in order to close the main steam isolation valve. =

Since main steam isolation occurs at the start of an accident, it is expected that the solenoids will operate normally and will perform their safety function. Should both solenoids fail and should the maio steam isolation valves remain open (main steam isolation valves also close on loss of air),

the main steam line can still be isolated by, the motor -operated isolation valve in each line located within the drywell. Since the solenoids are outside the drywell and the motor-operated isolation valves are inside the drywell, they will not oe simultaneously exposed to a harsh environment (except for radiation). Even if all these solenoid valves fail to operate simultaneously, valves 01-03 and 01-04 can be manually closed by isolating and venting the instrument air supply.

Therefore, justification for the continued safe operation of the plant is demonstrated.

35

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Solenoid valves located in the reactor building (SV/IV83.1-10 and 83.1-12)

FRC Equi ment Item No.: N/A Manufacturer: Asco Model: WP8300861RU Safety Function: Actuate to isolate drywell to floor drain and equipment drain waste collector tanks (83.01-10 and 83.1-12)

Qualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Floor drain and equipment drain lines from the drywell are each isolated at the start of an accident by motor-operated isolation valves inside the drywell (83.1-09 and 83.1-11) and air-operated valves outside the drywell (83.1-10 and 83.1-12). Except for radiation, either the motor-operated or air-operated valves (including their solenoids). will.not be exposed to a harsh environment. Since the isolation occurs at the star t of an accident, the solenoids are expected to function normally. Should the solenoids fail to de-energize and the air-operated isolation valves remain open, the drain lines can be isolated by the motor-operated valves or by isolating and venting the air to the air-operated valves.

Therefore, justification for the continued safe operation of the plant is demonstrated.

36

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Solenoid valves located in the reactor building FRC Equipment Item No.: N/A Manufacturer: Asco Model: Various (see below)

Safety Function: Isolate H2-02 and turbine building sample sink lines/return lines Qualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

These solenoids function to open and close sample lines from the drywell as fol lows:

Val ve Model No. Function SOV IVER-Ol LIILc~A- 5 Hp Op sam~Se Tine SOV IV201.7-02 LB8320A-25 H2 0~ sample line SOV 201.7-ZO HT830086RU Turbine Building sample sink SOV, 201. 7-21 HT830086RU Turbine Building sample sink SOV 201.7-23 HT8300B6RU H202 sample return SOV 201.7-26 HT8300B6RU H2 02 sample return SOV 201.7-22 HTX832A22V H2 02 sample return SOV 201.7-24 HTX8320A22V Turbine building sample sink SOV 201.7-25 HTX8320A22V Turbine building sample sink SOV 201.7-27 HTX832A22V H202 sample return SOV 201.2-419 HTX8320A22V H2 02 sample return SOV 201.2-420 HTX8320A22V H2 02 sample return SOV 201.2-421 HTX8320A22V H2 02 sample return SOV 201.2-422 HTX8320A22V H2 02 sample return SOV 201.2-429 HTX8320A22V H2 02 sample line SOV 201.2-430 HTX8320A22V H2 02 sample line SOV 201.2-431 HTX8320A22V H2 02 sample line SOV 201.2-432 HTX8320A22V H2 02 sample line SOV 201.7-03 LB8320A-25 H2 02 sample line SOV 201. 7-04 LB8320A-25 H2 02 sample line 37

V These valves are in the reactor building where the only harsh environmental parameter to which they are exposed during an accident in the drywell is radiation. Consequently, they are expected to function normally during the relatively short period of time when drywell atmospheric samples would be needed.

Therefore, justification for the continued safe operation of the plant is demonstrated.

38

4 NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Valve actuator located inside the drywe 1 1 (40-11)

FRC Equipment Item No.: 3 Manufacturer: Limitorque model: SMB3 Safet Function: Opens to initiate core spray gualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

I Valve 40-11 is one of four motor operated isolation valves which open to initiate core spray. The other three actuators (40-01, 09, 10) have recently been replaced with actuators procured from Limitorque as qualified actuators.

(40-11 is scheduled to be replaced in the .1984 refueling outage). The valves are arranged in two sets of two valves, one set in each core spray loop discharge header. Opening of either valve in a set allows full core spray flow to pass to the spray nozzles. Since each loop provides 100 percent of required core spray flow, only one valve of the four needs to open to provide 100 percent spray flow. One valve open in each loop provides 200 percent spray flow.

Since core spray is initiated at the start of an accident, it is expected that 40-11 will function before the onset of any environmentally induced failure.

However, should 40-11 fail to open, the opening of any one of three isolation valves with recently replaced actuators provides 100 percent spray flow.

Therefore, justification for the continued safe operation of the plant is demonstrated.

39

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET'quipment:

Valve actuators located inside the drywell (IV40-30 and IV40-31)

FRC Equipment Item No.:

Manufacturer: Limitorque ilode1: SBMOOO Safety Function: Closes the high point vent on high drywell pressure or reactor low-low level.

gualification Oeficienc Oocumented evidence of qualification Justification, for Continued Operation:

Valves 40-30 and 40-31 are drywell isolation valves. They are opened to vent the high points of the core spray discharge headers to the equipment system (via one inch lines). They are automatically closed by either hign drywell pressure or reactor low-low level signals.

I Should these valves fail to close, the one inch vent lines will be isolated by air operated valves (40-32 and 40-33) located outsid'e the drywell. Valves 40-32 and 40-33 are energized open and will fail closed on loss of air or electric power.

Therefore, justification for the continued safe operation of the plant is demonstrated.

40

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Valve actuators located inside the drywell (IV01-01, 01-02, 33-01, 33-02)

FRC Equipment Item No.:

Manufacturer: Limitorque Model: SMB4, SMBO Safety Function: Main steam line isolation valves (01-01, 01-02);

reactor water cleanup isolation valves (33-02) gualification Oeficienc  : Oocumented evidence of qualification Justification for Continued Operation:

Valves 01-01 and 01-02 are the main steamline isolation valves inside the drywell. Since these valves close at the start of an accident, it is expected that they will operate before they can be affected by a harsh environment.

Should one or both valves fail to close the respective main steam line will be isolated by the air operated isolation valve located outside the drywell (01-03 and 01-04). Valves 01-03 and 01-04 are energized to open and will fail closed on either loss of air or electric power.

Valves 33-01 and 33-02 are reactor water cleanup isolation valves inside the drywell. They are expected to operate in an accident since the isolation will occur at the onset of the accident. Should it fail to close, the system will be isolated by other motor operated valves (33-04 and 33-05) which are located outside the drywell.

Therefore, justification for the continued safe operation of the plant is demonstrated.

41

NINE MILE POINT UNIT 1 COMPONENT REVIEN SUMNRY SHEET Equipment: Valve actuators located inside the drywel 1 (IV 83.1-09, 83.1-11)

FRC Equipment Item No.: 7 Manufacturer: Limitorque Model: SMBOOO, SMB4 Safety Function: Isolates equipment drain and floor drain discharge lines from the drywell

/uglification Oeficienc : Oocumented evidence of qualification Justification for Continued Operation:

Valves 83.1-09 and 83.1-11 are the drywell isolation valves on the equipment drain and floor drain discharge lines, respectively. They close on a drywell isolation signal. Since isolation will occur at the star t of an accident, the valves are expected to operate before they can be affected by the harsh environment. Should the valves fail to close, the lines will be isolated by air operated isolation valves outside the drywell (83.1-10 and 83.1-12).

These valves are energized to open and wi 11 fail closed on loss of air or electtic power.

Therefore, justification for the continued safe operation of the plant is demonstrated.

42

I NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Valve actuators located inside the drywel 1 (IV 110-127)

FRC Equipment Item No.: 75 Manufacturer: Limitorque Model: SMBOOO Safety Function: Actuates isolation valve 110-127 to isolate reactor coolant sample line

/uglification Oeficienc  : Oocumented evidence'f qualification, similarity, aging degradation, qualified life Justification for Continued 0 eration:

Valve 110-127 is a normally closed isolation valve in the one inch reactor coolant sample line. If open, it will close on drywell isolation. Since isolation occurs at the start of an accident, it is expected that the isolation function will be performed. If the valve fails to close, the sample line will be isolated by valve 110-128 which is outside the drywell.

Therefore, justification for the continued safe operation of the plant is demonstrated.

43

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Valve actuators located in the main steam tunnel (IV31-07 and IV31-08)

FRC Equipment Item No.: N/A Manufacturer: Limitorque Model: SM82 Safety Function: Close feedwater line isolation valves

/uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Valves 31-07 and 31-08 isolate the feedwater lines at the penetration to the drywell. They are normally open, valves which remain open on a. loss of coolant accident or high energy, line break .since the feedwater system will shift to the high pressure coolant injection mode. These valves are closed by the operator when it is necessary to isolate a feed line (i.e. feedwater line break). Since this isolation would occur early in an accident and since the valves are outside the drywell, it is expected that they will function.

Should they fail to close, however, check valves 31-01 and 31-02 immediately next to their respective isolation valves provide the drywell barrier.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid valves in the reactor building (SOV 122-03B, 03A)

FRC Equipment Item No.: N/A Manuf acturer: Asco Model: HT8320A90MB Safety Function: Actuates IV122-02 to provide containment isolation of the post-accident sample line qualification Oeficienc  : Oocumented evidence of qualification Justification for Continued Operation:

These solenoids control valve 122-03 which is the containment isolation valve for the post-accident sample line. The isolation valve is normally closed with the solenoids de-energized. These. solenoids need to be qualified to ensure that post-accident samples can be drawn through this system.

Nevertheless, they fail safe (i.e. to isolate containment) and if they are failed, 'post-accident samples will have to be drawn by other means.

Therefore, justification for the continued safe operation of the plant is demonstrated.

45

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid valves in the reactor building (SV IV01-05, 01-06)

FRC Equipment Item No.:

Manufacturer: Numatics Model: 463-567 1JSP3 Safety Function: Main steam isolation valve by-pass line isolation Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

Valves 01-05 and 01-06 isolate the outboard main steam isolation valve by-pass line. These valves would only be open when warming up the main steam line.

Ouring normal operations, the isolation valves are shut and the solenoid valves (SV/IV 01-05 and SV/IV 01-06) wnich operate tne isolation valves, would be de-energized.

These solenoid valves do not have to operate to perform an accident mitigation function unless the by-pass lines happened to be open at the start of an accident. In this case, the solenoids merely have to de-energize to return to their safety position.

Therefore, justification for the continued safe operation of the plant is demonstrated.

46

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: H>-02 monitors located in turbine building (Tl, 12)

FRC Equipment Item No.: N/A Manufacturer: Beckman Model: H2-O2 Safety Function: Monitors post-accident hydrogen and oxygen concentrations in Containment.

/uglification Oeficiency: Oocumented evidence of qualification.

Justification for Continued 0 eration:

These monitors measure post-accident hydrogen and oxygen concentration in containment so that the containment atmospheric dilution system can maintain oxygen concentration below 4 percent. Since they will only be exposed to radiation when they are required to operate, they are expected to function normally. Should both detectors fail, the oxygen concentration can be maintained below 4 percent by over-inerting the containment with nitrogen.

Therefore, justification for the continued safe operation of the plant is demonstrated.

47

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Electric heater located in turbine building (202-76)

FRC Equi ment Item No.: N/A Manuf acturer: Honeywell Model: R72838 1081 Safety Function: Reduces relative humidity of reactor building emergency ventilation exhaust air to below 70 percent to retain charcoal efficiency.

gualification Oeficiency: Oocumented evidence of qualification.

Justification for Continued Operation:

This heater functions to retain high charcoal adsorber efficiency in the reactor building emergency ventilation system by reducing relative humidity of the filtered air to below 70 percent. The heater is located in the turbine building where the only harsh environmental parameter (wnen functioning is required) is radiation and therefore the heater is not expected to fail prematurely.

Therefore, justification for the continued safe operation of the plant is demonstrated.

48

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Insulating Tape, Sealant, and Undercoat for 5KV Terminal Insulation FRC Equi ment Item No.: N/A Manufacturer: Kerite Model: Splicing Compound Tape/Cement/ 3/4 inch Friction Tape Safety Function: Insulation for SKV Terminals Qualification Oeficiency: Qualification details not fully substantiated.

Justification for Continued Operation:

The insulating tape, sealant, and undercoat provides insulation for 5KV power terminals. A Wyle Laboratories Assessment Report (17655-TPE-4.1) addresses the qualification of these items, however, the Wyle report relies upon actual test details of Isomedix Report I-R975-Ol. Once the full test details and results can be verified, the materials are qualified.

In view of the nature of the remaining qualification deficiencies of these items, there is substantial confidence that the items will perform as designed under accident conditions. There is no reason to believe these materials ui 11 fail due to an accident environment since they were tested and only verification of test details and results remain.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Flow transmitters in containment (FT-19,20, 21, 22, 23, 24)

FRC Equipment Item No.: N/A Manufacturer: Endevco Model: 2273AM20 Safety Function: Provides operator with positive position indication of automatic depressurizatio'n 'system relief valves.

gualification Oeficiency: gualification testing not fully documented.

Justification for Continued Operation:

These items are accelerometers installed on the automatic depressurization system relief lines as a TMI action plan item. The equipment is to provide the operator with direct indication of a relief valve opening. There are no control functions, automatic safety functions, or accident mitigating functions associated with this equipment, other than providing the operator with another indication of relief valve opening.

Therefore, justification for the continued safe operation of the plant is demonstrated.

50

e' NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Solenoid valves located in the reactor building (SOV39-1 1C, -1 10,-12C,-120,-13C, -130,-14C,-140)

FRC Equi ment Item No.: N/A Manufacturer: Valcor Model: V70900-21-3 Safety Function: Isolation of emergency condenser drain lines Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

~

These solenoids close their respective isolation valves which isolate the emergency cooling drain lines from the main steam drain lines. The isolation valves are normally open valves which allow draining of condensation from the emergency cooling steam lines when the emergency condenser system is in the standby mode. They are automatically closed on an emergency cooling system isolation signal which results from either high radiation in. the emergency cooling system or an emergency cooling line break (excess flow).

In a loss-of-coolant accident or high-energy-line-break condition, these valves merely back up the 10 and 12 inch emergency cooling isolation valves which also close on an emergency cooling system isolation signal. Since the emergency cooling system will be isolated from the reactor vessel by the 10 and 12 inch valves, the drain line isolation valves perform no accident mitigating functions.

Therefore, justification for the continued safe operation of the plant is demonstrated.

51

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid valves in the reactor building (SOV 60-170,-17E,-180,-18E)

FRC Equi ment Item No.: N/A Manuf acturer: Valcor Model: V70900-21-1 Safety Function: Control of emergency condenser make-up supply valves

(}uglification Deficienc  : Documented evidence of qualification Justification for Continued 0 eration:

Each set of two emergency condensers has a single make-up supply valve (LCV 60-17 or LCV 60-18) which controls flow from its respective 40,000 gallon make-up tank. The make-up valves are controlled by one of two E/P converters, each with an individual level-transmitter providing the level signal. The purpose of solenoids60-170, 17E, 180, and 18E is merely to allow the operator to manually select which of the E/P converters controls the make-up valve.

During an accident, there is no need for the solenoids to operate unless the selected E/P converter or level transmitter malfunctions. Further, should a solenoid fai 1 (i.e. shift position), it merely places the back-up E/P converter into operation. Level control should be unaffected.

Finally, should the operator need to change E/P converters and the solenoids are inoperable, the make-up valve can be failed to the full-open position by isolating the instrument air supply.

Therefore, justification for the continued safe operation of the plant is demonstrated.

52

NINE MILE POINT UNIT 1 COMPONENT REVIEW SUMNRY SHEET

~Equi ment:. Electrical Insulation and Sealant Materials Located in the Steam Tunnel (Elevation 240 feet)

.FRC Equipment Item No.: See Below Manuf acturer: See Below Model: See Below Safety Function: Insulation of high-voltage terminals necessary for powering safety-related equipment.

/uglification Oeficienc  : Oocumented evidence of qualification.

Justification for Continued Operation:

~

These materials provide insulation. and sealing of high-voltage terminations and cables which power safety- elated equipment. The following materials are used:

FRC COMPONENT NO. MFGR. MOOEL FUNCTION Electrical Sealant 55 Ouxseal Filler for 5KV Terminals Electrical Tape 63 83 Cable connection insulation tape El ectri cal Sealant 64 GE 227 Filler for 5KV Terminals Insulating Varnish 66 Westinghouse 1309 5KV Terminal Insulation Varnish Electrical Tape 67 8380 Cable connection insulation tape 53

Experience has shown that these types of materials will survive harsh environmental conditions. Tests have been performed of commercially comparable materials, used. on heat shrinkable high-voltage terminations, with successful results.

In addition, a main-steam line break in the steam tunnel will be automatically terminated by closure of main-steam isolation valves on a high steam-flow signal. The differential pressure transmitters which signal the closure are located outside the steam tunnel, as is the inboard isolation valve. The isolation valve, which is located in the tunnel, is air-operated.and

'utboard fails closed on loss of air or power. Therefore, the harsh environment in the steam tunnel will be of short duration and there is no reason to suspect failure of insulation or sealant materials.

Therefore, justification for the continued safe operation of the plant is demonstrated.

54

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Motor Generator Sets and Control Panels in the Turbine Building (MG-162, MG-172, MG-162CP, MG-172CP)

FRC Equipment Item No.: N/A Manufacturer: General Electric Model: SLS4404 A22425/6PA4404 A22424 Safety Function: Provides Uninterruptible 120V AC Power to Reactor Protection Buses Qualification Oeficienc  : Oocumented evidence of qualification.

Justification for Continued Operation:

These motor-generator sets provide uninterr uptible power to the rea'ctor protection buses. They are normally in continuous operation as an AC motor/AC generator and automatically shift to the OC motor/AC generator mode on fai lure of the normal AC source.

The motor-generators and their control panels are located in the tur oine building where the maximum post-accident temperature is 133F and there is negligible post-accident radiation. Consequently, there is no reason to suspect that any failure of these generators will occur as a result of environmental conditions.

Finally, should a motor-generator fail (i.e. loss of the AC generator end),

its reactor protection bus can be powered from Panel 4167A by either off-site power or emergency diesels.

Therefore, justification for the continued safe operation of the plant is demonstrated.

55

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Circuit breakers located in the reactor building (PB-16A, 16B, 17A and 17B)

FRC Equipment Item No.: 62 Manufacturer: General Electric Model: AKO-5 Safety Function: Provides 600 volt electric power to safety-related equipment.

Oualification Deficiency: Documented evidence of qualification.

Justification for Continued 0 eration:

These circuit breakers provide power to various electrical equipment which are needed for loss of coolant accident/high energy line break mitigation. The circuit breakers are passive in the safeguard mode; that is they are normally closed and are not required to operate during the course of the loss of coolant accident/high-energy line break. In addition, the harsh environmental conditions in the reactor bui lding (with exception of radiation) are of a short duration and it is not reasonable to expect these conditions to result in the failure of a normally closed circuit breaker.

Therefore, justification for the continued safe operation of the plant is demonstrated.

56

1 'I NINE M'ILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Motor control centers located in the reactor building (MCC-155, 167, 161A/B, 171A/B) and battery board located in the turbine building (BB-12)

FRC Equipment Item No.: 72 Manuf acturer: General Electric Model: IC7700 Safety Function: Electric power distribution to various safety-related equipment

(}uglification Oeficienc: Oocumented evidence of qualification.

Justification for Continued Operation:

These motor control centers and battery board provide electric power to various safety-related equipment such as containment isolation valves, reactor building emergency ventilation fans, diesel-generator auxiliaries, and back-up instrument bus power.

A qualification analysis of this equipment is currently in progress by General Electric. A preliminary qualification report was issued in November 1983 (NEDC-30322-P). This report indicates a qualified life for the components of this equipment as follows:

Circuit Breaker 26 years ta 80 percent load and 25 percent duty cycle Magnetic Starter 13 years 9 25 percent duty cycle Control Power Transformer 18 years 9 25 percent duty cycle In view of the conservatism of the analysis (i.e. 25 percent duty cycle) and the preliminary results, it is concluded that this equipment will continue to function satisfactorily until such time as the analysis is completed and the equipment is included in the maintenance and surveillance program to ensure continuing qualification.

Therefore, justification for the continued safe operation of the plant is demonstrated.

57

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Control cable located in the reactor building FRC Equipment Item No.: 94 Manufacturer: Rockbestos Model: RSS6104 Safet Function: Instrument cable to post-accident radiation detectors (uglification Oeficienc  : Evidence of qualification not documented.

Justification for Continued 0 eration:

This control cable is associated with post-accident radiation detector s installed as a TMI action item. The cable is scheduled to be replaced with qualified Rockbestos RSS6104-1081 cable. The radiation detectors provide no automatic control function and are for indication only.

Therefore, justification for the continued safe operation of the plant is demonstrated.

58

t 1

NINE MILE POINT UNIT 1 COMPONENT REYIEH

SUMMARY

SHEET Equipment: Cable terminations in the reactor building FRC Equipment Item No.: 60 Manuf acturer: AMP Inc.

Model: Ring tongue terminal Safet Function: Cable termination for cable to various safety-related equipment Oualification Oeficiency: Anomaly during testing not resolved.

Justification for Continued Operation:

These cable terminations are associated with various equipment required for accident mitigation. During testing of these terminals, an anomaly occurred during which insulation sleeves slipped off the wire barrels on some unenergized "plasti-grip" specimens mounted vertically with the ring tongue end up during the accident simulation. An engineering evaluation is being conducted to resolve the anomaly and to determine the physical configuration of the terminations at Nine Mile Point Unit l.

Although the test identified this anomaly, there is no presently identified problem at Nine Mile Point Unit 1 for the following reasons:

1. There are no known vertically mounted insulation sleeves with the ring tongue end up.

2. Even if the insulation sleeves were to slip under accident conditions, there is no evidence that there will be an equipment failure.

3. Should a vertical configuration be discovered, the configuration will be immediately rectified or a justification for continued operation will be prepared based upon the accident mitigating function of the affected equipment.

Therefore, justification for the continued safe operation of the plant is demonstrated.

59

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Containment penetrations/connectors FRC Equipment Item No.: 52 Manufacturer: O.G. 0'8rien Model: Various Safety Function: Provides electric power for equipment inside the containment and containment boundary integrity.

gualification Oeficiency: Additional materials data and spray effects not documented.

Justification for Continued Operation:

Three separate test reports (FIRL F-C4879-1, Wyle 17655, Patel PEI-TR-82-12-101) all support the environmental qualification of the penetrations/connectors. While additional evaluation is required to fully establish qualification, testing and analyses p'erformed to date provides sufficient confidence in the capabilities of these assemblies to perform under accident conditions. In addition, the integrity of the penetrations is demonstrated every 3 years under peak containment accident pressure during 10CFR50, Appendix J, Type A testing.

Therefore, justification for the continued safe operation of the plant is demonstrated.

60

P NIN) NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Electrical terminal blocks inside containment (EB-5 and EB-25)

FRC Equipment Item No.: 53 Manufacturer: General Electric Model: N/A Safet Function: Provide electric power to various safety-related equipment inside containment

(}uglification Oeficiency: Similarity with tested component not fully documented.

Justification for Continued Operation:

Wyle analysis report 17655-TB-1.1 addresses the qualification of the terminal blocks. However, there'is a question as to the similarity between terminal blocks EB-5, 25 and the tested CR-151 terminal block. Another Wyle report, 17436-15, appears to fully qualify terminal blocks EB-5 and EB-25. However, these reports have not yet been obtained and reviewed. Nevertheless, there has been sufficient testing and analysis to conclude that the terminal blocks will perform satisfactorily in a post-accident environment and that the remaining qualification deficiencies can be adequately resolved.

Therefore, justification for the continued safe operation of the plant is demonstrated.

61

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Pressure switch located in the reactor. building (PS-R0-68A, B)

FRC Equipment Item No.: N/A Manuf actur er: Mercoid Model: OA23-156 R.63/X-OAW-43-103 R26E Safety Function: Trips control rod drive hydraulic pump on low suction pressure (5 psig decreasing)

(}uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

These pressure switches trip their respective control rod drive hydraulic pump on low suction pressure. Their malfunction can cause an i,nadvertent loss of the pump.

The control rod drive hydraulic system provides water for cnarging scram accumulators, for normal rod movement, for control rod drive cooling, and also provides a small amount of high pressure coolant injection which provides core cooling for certain primary system line breaks up to 0.003 square feet.

Control rod drive hydraulic functions are not required once an accident has occurred. The scram accumulators are charged at the start of the accident and therefore they will funcion. Once the scram occurs, normal rod motion is irrelevant. Finally, any high pressure injection needs will be accommodated by the high pressure coolant injection system.

Therefore,. justification for the continued safe operation of the plant is demonstrated. I 62

NINE MILE POINT UNIT 1 COMPONENT REYIEW

SUMMARY

SHEET Equipment: Pressure switches located in the reactor. building (PS70-108, 109, 110)

FRC Equipment Item No.: N/A Manufacturer: Mercoid Model: OA23-156-RSS Trips reactor building closed loop cooling pumps on low suction pressure.

Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

These pr essure switches trip their respective reactor building closed loop cooling pump on low suction pressure. Malfunction of these switches can cause inadvertent loss of one or more closed loop cooling pumps.

There are three 50 percent capacity reactor building closed loop cooling pumps. Normally, one or two of the three pumps are in operation and under accident conditions, one pump can supply the cooling loads having accident mitigation functions. The motors are located in the reactor building where the only harsh environmental parameter during an accident inside containment is radiation. Ouring a high energy line break in the reactor building, the motors may be exposed to a temperature of 300F but this temperature wi 11 be quickly lowered by isolating the high energy line break at the containment boundary. Consequently, it is considered highly unlikely that all three motors would fail under these conditions.

63

Should all reactor building closed loop cooling flow be lost following an accident, the following cooling loads with accident mitigation functions may be jeopardized: high pressure coolant injection, instrument air, Hp-Op monitoring. In case of loss of these systems, the following backups are avail able:

LOAD BACKUP High Pressure Core spray/automatic depressurization system Coolant Injection System and/or emergency cooling Instrument Air Essential components operated by instrument air fail to their safety position.

Motor-operated valves can be used to control the cooldown in the absence of air-operated valves.

Hp-Op Oxygen content can be maintained below 4 percent by excess inerting of the containment with Hp.

Therefore; justification for the continued safe operation of the plant is demonstrated.

64

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Valve actuator s outside containment FRC Equipment Item No.: 1, 4, 5, 6, 8, 9, 73, 74, 76, 77, 78 Manufacturer: Limitorque Model: SMBO, 000, 3/SB1, 2 Safety Function: Actuate valves for isolation and emergency core cooling system operation Qualification Oeficiency: Oocumentation deficiencies exist.

Justification for Continued Operation:

The motor operated valves listed below are either containment isolation valves or are safety system valves with accident mitigating functions:

PLANT ID NO. MOOEL NO. SYSTEM BV05-05 SMBOOO Emergency condenser vent to torus BV05-07 SMBOOO Emergency condenser vent to torus BV93-25 SMBO Containment spray raw water line BV93-26 SMBO Containment spray raw water line BV93-27 SMBO Containment spray raw water line BV93-28 SMBO Containment spray raw water line IV201-31 SMBOOO N2 fill and vent IV33-04 SB1 Reactor clean-up IV39-07 SB2 Emergency cooling steam line IV39-08 SB2 Emergency cooling steam line IV39-09 SB2 Emergency cooling steam line IV39-10 SB2 Emergency cooling steam line I V 110-128 SMB000 Reactor coolant sample line IV201-07 SMBOOO Torus vent and purge IV201-09 SMBOOO Orywell vent and purge IV201-17 SMBOOO Torus vent and purge IV34-01 SMBOOO Head spray line 65

PLANT IO NO. MOOEL NO. SYSTEM IV38-02 SMB3 Shutdown cooling IV40-05 SMBOO Core spray test line IV40-06 SMBOO Core spray test line IV80-01 SMBOO Containment spray suction IV80-02 SMBOO Containment spr ay suction IV80-114 SMBOOO Containment spray drain Line IV80-115 SMBOOO Containment spray drain Line IV80-21 SMBOO Containment spray suction IV80-22 SMBOO Containment spray suction IV81-01 SMBOO Core spray suction IV81-02 SMBOO Core spray suction I V81-21 SMBOO Core spray suction IV81-22 SMBOOO Core spray suction In the case of the containment isolation valves, they will close at the start of an accident. In addition, they are also backed up by a redundant isolation valve, often a check-valve or air-operated valve located on the other side of the containment boundary (i.e. not exposed to the same environment).

In the case of the safety system valves, these valves are either normally-open/remain open valves (i.e. core spray suctions,. containment spray suctions, emergency cooling steam lines, control road drive hydraulic return lines) or are normally shut/remain shut valves (i.e. core spr ay test line, drywell vent and purge, head spray) or function at the start of an accident (i.e. emergency condenser vent to torus, sample line).

Limitorque operators have been extensively tested throughout the industry and there is substantial evidence that they will not only perform a short-term accident function but will remain operable long-term. There is even evidence that these operators will survive an accident environment without an absolute seal. Any qualification deficiencies which remain outstanding are believed to be documentation problems.

Therefore, justification for the continued safe operation of the plant is demonstrated.

66

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Level transmitters located in the reactor building (LT36-03 A/8/C/D, 04 A/8/C/0, 05 A/8/C/0)

FRC Equipment Item No.: 35 Manufacturer: Rosemount ModeI: 1151 Safety Function: Reactor level indication and input to the reactor protection system or feedwater control

(}uglification Oeficienc  : Oocumented evidence of qualification Justification for Continued Operation:

These transmitters monitor reactor vessel level and provide indication to the operator, provide input to the reactor protection system or input to feedwater control:

LT 36-03A, 038, 03C, 030 High/Low Level LT 36-04A, 04$ , 04C, 040 Low-Low Level LT 36-05A, 058, 05C, 050 Low-Low-Low Level The trip functions performed by these transmitters will oe performed at the start of an accident. In addition, since they are located in the reactor building where the only harsh environmental parameter is radiation during a loss of coolant accident, it is unlikely that they will fail at onset of harsh environment. Failure of several transmitters is even less likely. Finally, even if all level transmitters are lost during the long term post-accident period, core cooling is ensured by continuing core spray flow.

Therefore, justification for the continued safe operation of the plant is demonstrated.

67

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Pressure transmitters in the reactor building (PT 201.2-476 A/B/C/0)

FRC Equi ment Item No.: 30 Manufacturer: Rosemount Model: 1151 OP Safety Function: Provides containment pressure signal to the reactor protection system Oualification Oeficienc  : Oocumented evidence of qualification Justification for Continued Operation:

These transmitters provide the high containment pressure trip to the reactor protection system. Since they are located in the reactor building where the only harsh environmental parameter, during an accident inside containment, is radiation, and since they will trip at the start of an accident, there is no reason to suspect that the trip function will not be performed. This is particularly true because simultaneous failure of these transmitters at the start of an accident due to radiation is highly unlikely.

Therefore, justification for the continued safe operation of the plant is demonstrated.

68

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Flow transmitters located in the reactor building (FT 201.8-35, 41, 45 and 201.9-26, 31, 80)

FRC Equi ment Item No.: N/A Manuf acturer: Fisher Model: 2340 Safet Function: Provides for controls of containment air dilution N2 flow and pressure Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued 0 eration:

These transmitters function to regulate flow control valves in the containment air dilution system to control the pressure and flow rate of make-up N2.

These valves are needed only in the case of a design basis accident within containment where hydrogen and oxygen are being generated as a result of the accident. Failure of all of the transmitters could result in a loss of N2 make-up.

Since these transmitters are located in the reactor building where the only harsh environmental parameter (during a drywell accident) is radiation and since inerting wi 11 occur in the early stages of an accident, there is no reason to suspect that failure will occur before they have performed their safety function.

Therefore, justification for the continued safe operation of the plant is demonstrated.

69

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Valve actuators located inside the drywell (IV40-01, 09, 10)

FRC Equipment. Item No.: 3, 7 Manuf acturer: Limitorque Model: SMB 3 Safety Function: Open to initiate core spray qualification Deficiency: . Documented evidence of qualification Justification for Continued Operation:

These actuators were recently purchased as qualified actuators and were installed. There appear to be some documentation problems regarding qualified life. Nevertheless, these actuators are believed to be qualified and wi 11

'certainly perform their short-term safety function.

Therefore, justification the for continued safe operation of the plant is demonstrated.

70

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid actuator s and position switches located in containment (PSV NR-108A to F and POS NR-108A to F)

FRC Equipment Item No.: 14 Manufacturer: GE/Unimax blode1: CR9503-213C/KBL7HB-5 Safety Function: Operates and indicates position of electro-matic relief valves of the automatic depressurization system Oualification Deficiency: Documented evidence of qualification Justification for Continued Operation:

The automatic depressurization system relief valves are install'ed to reduce reactor vessel pressure to allow core spray system operation for breaks smaller than 0.3 square feet. Three of the six installed valves are sufficient to provide the design blowdown. Relief valves are operated manually or automatically upon simultaneous low-low-low reactor level and high drywell pressure.

Since a reactor vessel blowdown wi 11 occur at the start of a small-break loss coolant accident (i.e. brief exposure to narsh environment), it is expected

'f that at least three of the six valves will function. Once reactor pressure has been reduced to permit core spray flow, the relief valves no. longer perform an accident mitigating function. .

Should a condition occur where automatic depressurization system was required (i.e., loss of coolant accident with pipe break less than 0.3 square feet) and should the system fail to achieve the necessary blowdown (i.e., at least four of six valves fail to open), adequate core cooling will be provided and reactor vessel pressure will be lowered by the high pressure coolant injection system, the emergency condensers, or a combination of the two. In addition, with the loss of coolant accident inside containment, the high pressure coolant injection system is essentially located in 'a mild environment (turbine building) and the emergency coo'ling system will be subjected to hign radiation only (reactor building).

Therefore, justification for the continued safe operation of the plant is demonstrated.,

71

APPENDIX B

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Temperature switches located in steam tunnel (TS IB-10A thru 10H, J thru N, P, g, R)

FRC Equipment Item No.: 28 Manufacturer: Fenwal Model: =

1700240 Safet Function: Closes main steam isolation valves on high temperature Oualification Oeficienc  : Simi liarity, aging, pressure, radiation Justification for Continued Operation:

The temperature switches monitor steam tunnel temperature and initiate main steam isolation ~alve closure on high steam tunnel temperature.

If the break is in the main steam line, increased pressure drop across the main steam line flow limiter will initiate main steam isolation valve closure. The pressure drop is measured by instruments dPT 01-26A thru H located outside th'e steam-tunnel (reactor building elevation 237 feet north instrument room) which are not subject to the harsh environment.

A main steam line break in the tunnel will result in low reactor water level, which will cause a scram and low-low water level, which will initiate both main steam isolation valve closure and containment isolation. This instrumentation is located in the reactor building which is not suojected to a harsh environment from a high energy line break in the steam tunnel.

Therefore, justification for the continued safe operation of the plant is, demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Containment pressure indication (PT201.2-483, 484)

FRC Equi ment Item No.: 82 Manuf actur er: Rosemount Model: 1153 Saf et Functi on: Provide containment pressure indication to the control room operator following a loss of coolant accident

/uglification Oeficiency: Aging, qualified life Justification for Continued Operation:

Pressure transmitters, 201.2-483 and 484, provide the control room operator with indication of pressure inside the primary containment following a loss of coolant accident. These transmitters do not automatically initiate any safety systems, they provide indication only. In the event of failure of one of these devices due to age-related degradation, the same function will be provided by the redundant transmitter. Simultaneous failure of both components due to aging is highly improbable. In addition, both devices are located outside of the primary containment and will not be subjected to a steam environment during a loss of coolant accident.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Pressure transmitters located in the reactor building (PT36-23A, B)

FRC Equipment Item No.: 81 Manufacturer: Rosemount Model: 1153GA9 Safety Function: Monitors reactor pressure to correct for pressure change effects on LT36-24A, B

(}uglification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

These pressure transmitters were recently installed as required by NUREG 0737 Item II.F.2.3, Instrumentation for detection of inadequate core cooling. This level monitoring system does not provide for actuation of any safety systems.

Level monitoring installed to mitigate the consequences of a design basis accident is provided by separate and independent instruments.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Level transmitters located in the reactor building (LT36-24A, 8)

FRC Equipment Item No.: 83 Manuf acturer: Rosemount Model: 11530A5 Safety Function: Monitor reactor vessel level in fuel zone Qualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

These level transmitters were recently installed as required by NUREG 0737 Item II.F.2.3, instrumentation for detection of inadequate core cooling. This level monitoring system does not provide for actuation of any safety systems.

Level monitoring installed to mitigate the consequences of a design basis accident is provided by separate and independent instruments.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Solenoid valves located in the reactor building (SOV 122-04 through 122-11)

FRC Equipment Item No.: 79 Manufacturer: Asco Model: HT 8317A29 Safet Function: Actuates air operated post accident sampling blocking valves,(122-04 through 06, 122-08 through 11)

Oualification Oeficiency: Oocumented evidence of qualification Justification for Continued Operation:

These valves were recently installed as part of the post accident sampling system.

Failure of these valves would not prevent safety-related equipment from performing its intended function. Failure would not affect the plant's accident mitigation capability or its capability to prevent the release of radioactive material to the environment.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Control cable splice located in the steam tunnel at elevation 240 feet FRC Equipment Item No.: 59 Manufacturer: AMP Inc.

Model: Butt Safet Function: Butt connectors for various safety-related control cables Oualification Oeficienc  : Assessment report does not provide sufficient similarity information Justification for Continued Operation:

The test report does not provide sufficient similarity information. An engineering evaluation will be conducted to establish similarity.

The failure of equipment which is fed by cable routed through the steam tunnel at elevation 240 feet does not prevent safe shutdown or result in unacceptable off-site doses as a result of a main steam line break inside the steam tunnel.

A large break in the main steam line downstream of the main steam flow limiter causes increased pressure drop across the limiter. This initiates main steam isolation valve closure. The pressure drop is measured by instruments PT Ol-26A thru H located outside the steam tunnel (reactor building elevation 237 feet, north instrument room) which are not subject to the harsh environment.

The inboard main-steam isolation-valve and its related electrical components are not subject to the harsh environment as they are located in the drywell.

The outboard main-steam isolation-valves which are located in the steam tunnel are air operated valves which fail closed on loss of air or power.

A main steam line break in the tunnel will result in low reactor water level whicn wi 11 cause a scram, and low-low reactor water level which will initiate main steam isolation valve closure and containment isolation. The motor-operated feedwater isolation valves do not receive any isolation signal. In addition, there is a check valve in series which serves the isolation function.

Failure of the steam tunnel temperature switches, or main steam line radiation detectors does not prevent safe shutdown or result in unacceptable off-site doses.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Vacuum switches (VCS68-11A, 11B, 12A, 12B, 13A, 13B)

FRC Equipment Item No.: 37 Manufacturer: Mercoid Model: CP 4122 Safet Function: Controls air operated isolation valves 68-08, 09, 10 to admit building atmosphere to drywell/torus ualification Oeficienc  : Oocumented evidence of qualification Justification for Continued Operation:

The function of vacuum switches68-11A, 118, 12A, 12B, 13A, 138 is to open air operated isolation valves 68-08, 09 and 10 when torus pressure is negative with respect to the reactor building. This will admit reactor building atmosphere to the torus, thus equalizing the pressure. There are two vacuum switches associated with each valve. The switches are installed in parallel so that either switch can operate the valve. Simultaneous failure of both switches due to age-related degradation is highly improbable.

In addition, air operated check valves will provide the isolation function should these valves fail to reclose.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Control rod drive scram dump volume (SV NC-15A, 8, C, 0, 16A, 8)

FRC Equipment Item No.: 13 Manuf acturer: Asco (A&B Valves)/Valcor (C&0 Valves)

Model: HVA904058A/V70400-21-1 Safety Function: Isolate dump volume drain and vent during reactor scram Oualification Oeficiency: Aging degradation, qualified life, aging simulation Justification for Continued Operation:

Ouring a reactor scram, the scram discharge volume isolates by the closure of two air operated vent valves and two air operated drain valves which are in series.

Solenoid operated valves NC 15A and 158 control the operation of only one set of isolation valves (one drain and one vent). The redundant set of isolation valves is controlled by solenoid operated valves NC 15C and 150. These valves were not included in the original program but are now included. Simultaneous failure of both systems due to aging is highly improbable.

Solenoid operated valves NC 16A and 8 (backup scram valves) perform a function whicn is redundant to the two scram valve pilots on each of the 129 control rod drive hydraulic units. Instrument air will be vented from the scram valve pilot air header, during a scram initiation, by either the backup scram valves or the scram valve pilots. Failure of solenoid operated valves NC 16A and 8 to operate would not adversely affect the scram function. Loss of power causes the valves to go to the correct position. Simultaneous failure of the scram valve pilots and the backup scram valves is highly improbable.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Master trip unit located in the reactor building (dPTOl-26A to H, dPT201.2-476A to 0, dPT36-06A to 0)

FRC Equipment Item No.: 19 Manufacturer: Rosemount Model: 510 OU

~!<< Orywell pressure trip unit/emergency condenser high flow trip/main steam line flow trip

/uglification Oeficiency: Aging, qualified life, pressure, spray, radiation Justification for Continued Operation:

These instruments are located in the reactor building at elevation 281 feet, one in each of the four corners.

Or ell Pressure Trip Unit For breaks within the drywell, temperature, pressure or humidity does not increase in the reactor building where these'rip units are located.

Radiation is the only effect that these units would experience after a loss of coolant accident.

Only the drywell pressure trip units, are needed to mitigate a loss of coolant accident. These units automatically perform their safety function within seconds. Once they have initiated their safety function, no failure could negate it. They fail safe on trip unit failure or loss of power.

Emer enc Condenser Hi h Flow Trip The effects of the emergency condenser flow trip units for dPT.-36-06A, B, C and 0 are needed to mitigate an emergency condenser steam line break outside of containment. These trip units are not in close proximity to the break locations and automatically perform their safety function within the first few seconds of the event.

Main Steam Line Flow Tri Main steam line flow trip unit's dPT's01-26A to H are not exposed to the harsh environment of a steamline break, since they are not located near the breaks. These instruments automatically perform their safety function ny isolating the steam line at the start of the event and once performed, instrument failure cannot open these valves. Subsequent instrument/trip unit fai lure has no effect on the operator's response.

Therefore, justification for the continued safe operation of the plant is demonstrated.

10

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Trip units located in the reactor building (LT 36-03A, 8, C, 0; LT 36-04A, 8, C, 0; LT 36-05A, 8, C, 0; PT 36-07A, 8, C, 0; PT 36-08A, 8, C, 0)

FRC Equipment Item No.: 20 Manuf acturer: Rosemount Model: 510 OU Safety Function: Oetects high, low, low-low and low-low-low vessel level and high and low reactor pressure

(}uglification Oeficienc  : Aging, qualified life, pressure, spray, radiation Justification for Continued Operation:

These trip units are located in the reactor building, one in each of .the four analog trip system cabinets. These cabinets are located in the four corners of the building, one per corner.

They are all de-energnized to actuate and provide inputs to the reactor protection system. Their safety functions are automatically performed early in the event sequence as a result of process conditions or as the result of a trip unit failure. Failure of these level and pressure transmitter trip units can only affect long term monitoring and will not affect accident mitigation.

Therefore, justification for the continued safe operation of the plant is demonstrated.

NINE NILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Radiation detector located in reactor building Elevation 261 feet (201.7 - 36A, 37A)

FRC Equipment Item No.: 86 Manufacturer: General Atomic Model: R023 Safety Function: High range radiation monitor Oualification Deficiency: Similarity, aging, radiation Justification for Continued Operation:

These radiation detectors were recently installed as r equired by NUREG 0737 item II.F.1.3. The system does not provide either actuation or indication that a safety system is performing its intended function.

Therefore, justification for the continued safe operation of the plant is demonstrated.

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NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Electrical cable located inside the drywell FRC Equipment Item No.: 65 Manufacturer: General Electric Model: Vulkene Power and Control 600V and 1000V Ou a 1 i fi cat i on Oef i ci ency: qualification documentation lacks similarity analysis Justification for Continued Operation:

Vulkene insulated cable without a jacket and with Neoprene and Irradiated Crosslinked Polymer jackets have operated properly under conditions exceedin9 the environment at Nine Mile Point Unit 1 and have been tested within the industry. Unjacketed Vulkene cable and Vulkene insulated cable with a PVC jacket are utilized at Nine Mile Point Unit 1. The addition of. a PVC jacket would further improve the environmental resistance of the overall system:

Therefore, justification for the continued safe operation of the plant is demonstrated.

13

NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET

~Equi ment: Flow switch, FET-664 FRC Equipment Item No.: 84 Manufacturer: Foxboro Model: El 3DL Safety Function: Post accident sampling

/uglification Deficiency: Documented evidence of qualification Justification for Continued Operation:

This flow switch was recently installed as required by NUREG 0737 item II.B.3, post accident sampling system. The system does not provide information to indicate whether any plant safety functions are being accomplished.

Failure of this flow switch would not prevent any safety-related equipment from performing their intended safety function. Failure would not affect the plant's accident mitigation capability or its capability to prevent the release of radioactive material to the environment.

Therefore, justification for the continued safe operation of the plant is demonstrated.

14

l NINE MILE POINT UNIT 1 COMPONENT REVIEW

SUMMARY

SHEET Equipment: Power cable FRC Equipment Item No.: 71 Manuf acturer: Ker ite Model: SkV quadruplex Cable Safet Function: Power distribution to engineered safeguard equipment Oualification Oeficienc  : Oocumented evidence of qualification Justification for Continued 0 eration:

The Kerite power cable is used as distribution circuits to motors and transformers. The power conductor insulation is Kerite with a neoprene jacket. The ground conductor insulation is polyvinyl cnloride (PVC). The three individually insulated and jacketed power conductors and the insulated ground conductor are cabled together in quadruplex form with no further cover ing. The cable is installed primarily in conduit. Safety-related use of the cable is limited to areas outside the primary containment. Power cable with Kerite insulation have been utilized extensively over the years in various Niagara Mohawk Power Corporation steam generating facilities uith no known age related failures. Temperatures would be consistent with those existing within the reeactor building. Temperature and pressure conditions resulting from a high energy line break at cable terminations which are enclosed in junction boxes to emergency core cooling motors are only slightly above normal operating values. Wyle gualification Test Report 47176-1 demonstrates qualification pending review.

Based on the above, continued operation is justified until the cable is qualified or replaced with qualified cable.

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