ML20003F546

Nonproprietary Version of Equipment Environ Qualification, Oyster Creek Plant, Final Technical Evaluation Rept

ML20003F546
Person / Time
Site:	Oyster Creek
Issue date:	04/20/1981
From:	Crane C FRANKLIN INSTITUTE
To:
Shared Package
ML20003F545	List: ML20003F544 ML20003F546
References
CON-NRC-03-79-118, CON-NRC-3-79-118 IEB-79-01B, TER-C5257-195, NUDOCS 8104220510
Download: ML20003F546 (155)

Text

{{#Wiki_filter:_ _. _ - _ . .. - O TECHNICAL EVALUATION REPORT EQUIPMENT ENVIRONMENTAL QU ALIFICATION JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION NRC DOCKET NO. 50-219 NRC TAC NO. 42524 FRC PROJECTC5257 NRC CONTRACT NO. NRC 03-79-118 FRC TASK 195 Prepared by Franklin Research Center The Parkway at Twentieth Street Philadelphia, PA 19103 FRC Group Leader: C. J. Crane Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 NRC Lead Engineer: J. Lombardo l April 20, 1981 This report was prspared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. ah

                                                                                                                          .  . Franklin Research Center A Division of The Franklin Institute The Bengrmn Franklin Parkway. Phila.. Pa 19103 (215) 448. t 000 8104e20510

TECHNICAL EVALUATION REPORT EQUIPMENT ENVIRON MENTAL QU ALIFICATION JERSEY CENTRAL POWER & LIGHT COMPANY , 0YSTER CREEK NUCLEAR GENERATING STATION i NRC DOCKET NO. 50-219 NRC TAC NO. 42524 FRC PROJECT C5257 NRC CONTRACT NO. NRC-03-79-118 FRC TASK 195 Prepared by Franklin Research Center The Parkway at Twentieth Street Philadelphia, PA 19103 FRC Group Leader: C. J. Crane Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 NRC Lead Engineer: J. Lombardo April 20, 1981 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. 4

                                                .... Franklin Research Center A Division of The Franklin Institute The Benjarrun Frankhn Parkway. PNia.. Ps 19103(2151448 1000

DELETED MATERIAL is PROPmiETARY INFORMATION TER-C5257-195 COtrfENTS Section Title Page 1 INTRODUCTION. . . . . . . . . . . . . . 1-1 1.1 Purpose of the Evaluation . . . . . . . . . 1-1 1.2 Generic Issue Background . . . . . . . . . 1-1 1.3 Specific Issue Background . . . . . . . . . 1-5 1.4 Scope of the Evaluation . . . . . . . . . 1-6 2 NRC CRITERIA FOR ENVIRONMElffAL QUALIFICATION. . . . . . 2-1 2.1 Criteria Provided by the NRC . . . . . . . . 2-1 2.2 Staff Positions and Supplemental Criteria . . . . . 2-1 2.2.1 Service Conditions Inside Containment for a Loss-of-Coolant Acciden't. . '. . . . . . 2-1 i 2.2.2 Submergence . . . . . . . . . . . 2-2 2.2.3 Equipment I4cated in Areas Normally Maintained at Room conditions . . . . . . . . . 2-2 2.2.4 Simulated Service Conditions and Test Duration . . 2-3 2.2.5 Deferment of Qualification Review . . . . . 2-3 2.2.6 Test Sequence . . . . . . . . . . 2-4 2.2.7 Radiation . . . . . . . . . . . 2-4 3 METHODOLOGY USED BY FRC . . . . . . . . . . . 3-1 4 TECHNICAL EVALUATION . . . . . . . . . . . 4-1 4.1 Methodology Used by the Licensee . . . . . . . 4-1 4.1.1 Completeness of Equipment List . . . . . . 4-2 4.1.2 Environmental Service Conditions . . . . . 4-3 4.1.3 Aging and Qualified Life . . . . . . . 4-5 4.2 Equipment Qualified for Plant Life. . . . . . . 4-7 111 Mjdu Franklin Research Center w omam m

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l DELETED W ATERIAL O PROPRIETA*:Y INFORM ATION l TER-C5257-195 1 CONTENTS Section Title Page 4.2.1 NRC Category I.a Equipment That Fully Satisfies All Applicable Requirements of the DOR Guidelines . . . . . 4-7 4.2.2 NRC Category I.b Equipment With Acceptable Deviations From the DOk Guidelines . . . . . . . . . . . 4-7 4.3 Equipment Qualified With Restrictions . . . . . . 4-8 4.3.1 NRC Category II.a Equipment That Satisfies All Applicable Requirements of the DOR Guidelines With the Exception of Qualified Life . . . . 4-8

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4.3.2 NRC Category II.b Equipment That Satisfies All Applicable Requirements of the DDR Guidelines With the Exception of Qualified Life Provided That Specific Modifications Are Made. . . . . . 4-17 4.3.3 NRC Category II.c Equipment for Which Deviations From the DOR Guidelines Are Judged Acceptable With the Exception of Qualified Life . . . . . . . 4-17 4.4 NRC Category III Equipment That Is Exempt From Qualification . . . . 4-25 4.5 Equipment for Which Documentation Contains Deviations From the Guidelines That Are Judged Unresolved. . . . 4-27 4.5.1 NRC Category IV.a Equipment That Has Qualification Testing Scheduled but Not Completed. . . . . . . 4-27 4.5.2 NRC Category IV.b Equipment for Which Qualification Documentation in Accordance With the Guidelines Has Not Been Established. . . . . . . . . . . 4-27 iv D' A

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DELETED MATEAIAL IS PROPf4ETARY lNPORMATION TER-C5257-195 CONTENTS Section Title Page 4.6 NRC Category V Equipment That Is Unqualified . . . . . . . . 4-52 4.7 NRC Category VI Equipment for Which Qualification is Deferred . . . . 4-79 4.8 Summary of the Evaluation . . . . . . . . . 4-93 5 CONCLUSIONS . . . . . . . . . . . . . . 5-1 6 REFERENCES . . . . . . . . . . . . . . 6-1 APPENDIX A - ENVIRONMENTAL SERVICE CONDITIONS APPENDIX B - LISTING OF SAFETY-RELATED ELECTRICAL EQUIPMENT APPENDIX C - SAFETY SYSTDIS AND DISPLAY INSTRUMEEATION FOR WHICH ENVIRONMENTAL QUALIFICATION IS M BE ADDRESSED APPENDIX D - EVALUATION OF LICENSEE JUSTIFICATIONS FOR CONTINUED OPERATION APPENDIX E - CORRELATION OF EQUIPMEW ITEM NUMBERS WITH REPORT SECTIONS OF DRAFT INTERIM AND FINAL TECHNICAL EVALUATION REPORTS APPENDIX F - PROPERTIES OF CAST PHENOLIC RESINS APPENDIX G - EFFECTS OF NUCLEAR RADIATION DOSE RATE ON CABLE PERFORMANCE DURING A LOCA V

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DELETED MATERtAL t$ PROMWETARY INFORMAflON TER-C5257-195 ACKNOWLEDGMENTS The engineering, administrative, and editorial staff who produced this Technical Evaluation Report labored long hours giving the required intense attention to complex details. Without their extraordinary effort and devotion to this work, these reports could not have been produced in the short time that was available. The following are all to be thanked. The principal contributors to the technical effort are: J.C. Archer (ECI) G.J. Overbeck (W) C.J. Crane I.H. Sargent (W) T.J. DelGaizo (W) S.R. Schmitt J.A. Murphy W.H. Steigelmann (SRC) The associate technical contributors who supplied essential inputs are: A. Cassell K. Kauffman C.B. Chan P.N. Noell M. Hargitay J.S. Scherrer (W) J.E. Kaucher (W) K.E. Weise The editorial staff who proofread and edited all the report drafts ares l R.J. Carelli S. Reynolds M. Dank M. Rothman E.K. Friedman M. Sherritze P. Grant-Kingsberry R. Wilson M.A. Musil l The word processing group who typed and made the numerous changes needed to arrive at the final report are: 1 F. Davis A. Oponski l A. Rogers A. Mcdonald A vii Jb!5nklin Research Center ao vrs. row

DELETED MATER AL O PRoPAlETARY INFOAMATION TER-CS257-195 Overall management of the EEQ project was in the capable hands of Dr. S.P. Carfagno, FRC Project Manager, and Dr. Z. Zudans, the FRC Project Director. The following oversaw and reviewed this work for the Nuclear Regulatory Commission: J.J. Lombardo (Lead Engineer Equipment Environmental Qualification, Equipment Qualification Branch) P. DiBenedetto (Section Leader, Equipment Qualification Branch) Z.R. Rosztoczy (Chief, Equipment Qualification Branch) E.J. Butcher (Project Officer) Many others, too numerous to list, also aided in this effort and are to be thanked. l Subcontractors ECI = Energy Consultants, Inc. SRC = Synergic Resources Corporation W .= WESTEC Services, Inc. viii O-j

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I i OELETED MATEfWAL 18 PROMWETARY INPCRedATION TER-C5257-195

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i l l l IDENTIFICATION OF PROPRIETARY INFORMATION Some of the information in this technical evaluation report was obtained from manufacturer's proprietary test reports. All proprietary infor: nation used in this report has been deleted. i l 1* l 43 e..j Frankhn Research Center a w w % r,-- i

l otLETac MAfamat a peommeTany wwommanow TER-C5257-195

1. INTRODUCTION 1.1 PURPOSE OF THE EVAI,UATION The purpose of this report is to evaluate qualification documentation of nuclear power plant safety-related electrical equipment in accordance with criteria established by the NRC and to identify (1) equipment for which qualification documentation is adequate, i.e., substantiates that the equipment is capable of performing its specified design basis safety function when it is exposed to a harsh environment and (2) equipment for which qualification documentation is deficient, i.e., does not give reasonable I

assurance that the equipment is capable of performing its specified safety function. Where practical, this report presents recommendations for actions to remedy deficiencies. 1 - 1.2 GENERIC ISSUE BACKGRCUND The NRC criteria for reviewing the safety of nuclear power generating stations include the requirement that the qualification of safety-related electrical equipment be substantiated by auditable documentation of the program that establishes the ability of the equipment to function as specified in the station design. This report ir restricted to a technical evaluation of the equipment's ability to function in harsh environments resulting from design basis events (DBEs). Cualification criteria applied during the licensing of older nuclear power plants have been :nodified over the years, and specific industry standards concerning qualification have been revised as the design of reactor systems has changed and as regulatory and operating experience has accumulated. Examples of such standards are IEEE Stan.dards 279-71, 323-74, 383-74, 317-76, 334-74, 381-77, 382-80, and 627-80. NRC NUREG documents 0413 and 0588 have been developed to address this topic. In particular, NUREG-0588 (puolished for comment in December 1979) formilly presented the NRC staff positions regarding selected areas of environmental qualification of O- 1-1

      .dJ Franidin Research Center a c==a w w     a -                                                                  1

DELETED MATERIAL IS PROPRIETARY INFORMATION TER-CS257-195 safety-related electrical equipment in the resolution of General Technical Activity A-24, " Qualification of Class IE Safety Related Equipment." The positions documented therein are applicable te plants that are or will be in the construction permit or operating license review process. Although qualification standards and regulatory requirements have undergone considerable development, all of the currently operating nuclear power plants are required to comply with 10CFR50, Appendix A, General Design Criteria for Nuclear Power Plants, Section I, Criterion 4. This criterion states in part that " structures, systems and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing and postulated accidents, including loss-of-coolant accidents." In 1977, the PRC staff instituted the Systematic Evaluation Program (SEP) to determine the degree to which the older operating nuclear power plants deviated from current licensing criteria. The subject of electrical equipment environmental qualification (SEP Topic III-12) was selected for accelerated evaluation as part of this program. Seismic qualification of equipment was to be addressed as a separate SEP topic. In December 1977, the NRC issued a generic letter to all SEP plant licensees requesting that they initiate reviews to determine the adequacy of existing equipment qualification documentation. Preliminary NRC review of licensee responses led to the preparation of NUREG-0458, an interim NRC assessment of the environmental qualification of electrical equipment. This document concluded that "no significant safety deficiencies requiring immediate remedial actions were identified." However, it was recommended that additional effort should be devoted to examining the installation and environmental qualification documentation of specific electrical equipment in all operating reactors. On May 31, 1978, the NRC Office of Insrt",icn and Enforcement issued IE Circular 78-08, " Environmental Qualifica!.crs 11 F1:fety-Related Electrical Equipment at Nuclear Power Plants," sn et r?e . ired all licensees of operating D- 1-2 dM aFranidin

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DELETED MATERIALIS PROPRIETARY INFORMATION TER-CS257-195 plants (except those included in the SEP program) to examine their installed safety-related electrical equipment and ensure appropriatr qualification documentation for equipment function under postulated atx (dent conditions. Subsequently, on February 8, 1979, the NRC Office of Inspection and Enforce-ment issued IE Bulletin 79-01, which was intended to raise the threshold of IE Circular 78-08 to the level of Bulletin, i.e., action requiring a licensee response. This Bulletin required a complete re-review of the environmental qualification of safety-related electrical equipment as described in IE Circular 78-08. The review of the licensee responses indicated certain deficiencies in the scope of equipment addressed, definition of harsh environments, and adequacy of qualification documentation. It became apparent that generic criteria were needed to evaluate the electrical equipment environmental qualification for both SEP and non-SEP operating plants. Therefore, during the second half of 1979, the Division of Operating Reactors (DOR) of the NRC issued internally a document entitled " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" [6] .* (The document is hereaf ter referred to as the " DOR Guidelines.") The document was prepared as a screening standard for reviewing all operating plants, including SEP plants. It was originally intended that the licensees evaluate I their qualification documentation in accordance with the DOR Guidelines.

     & wever, initial NRC review of this documentation, which was compiled to support licensee submittals, revealed the nee.d for obtaining independent evaluations and for accelerating the qualification review program.

In October 1979, the NRC awarded Franklin Research Center (FRC) a centract to provide assistance in the " Review and Evaluation of Licensing Actions for Operating Reactors," which included an assignment for review of l

      *For References, see Section 6.           Note that reference numbers are not presented in sequential order.

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DELETED MATERIAL IS PROPRtETARY lNFORM ATION TER-CS257-195 equipment environmental qualification documentation under SEP Topic III-12. FRC was to review equipment environmental qualification documentation and to present the results in the form of a Technical Evaluation Report for the 11 oldest plants (included in the SEP review) . On January 14, 1980, the NRC Office of Inspection and Enforcement issued the DOR Guidelines and IE Bulletin 79-OlB, which expanded the scope of IE Bulletin 79-01 and requested additional information on environmental qualification of safety-related electrical equipment at operating facilities, excluding the 11 facilities undergoing the SEP review. This Bulletin cited the DOR Guidelines as the criteria to be used in evaluating the adequacy of the safety-related electrical equipment qualification. The scope of the review was expanded to include high energy line breaks (inside and outside containment) in addition to equipment aging and submergence. The NRC advised the licensees that the criteria contained in the DOR Guidelines would be used in its review of licensee submittals; problems arising from this review would be resolved using NUREG-0538 as a guide. In early February 1980, the NRC decided that Indian Point Units 2 and 3 and Zion Units 1 and 2 should be included within SEP Topic III-12 for the purpose of equipment environmental qualification review. On February 21, 1980, the NRC and representatives of the SEP Plant Owners Group held an open meeting at NRC headquarters to discuss an accelerated review program in accordance with the DOR screening guidelines. Re presen-tatives of the Indian Point Units and Zion Station also attended this meeting. The NRC formally issued to all licensees represented at the meeting the DOR Guidelines document which included a second document, " Guidelines for Identification of That Safety Equipment of SEP Operating Reactors for Which Environmental Qualification Is To Be Addressed" [6], together with the request that the licensees review their plant systems and provide additional equipment O 1-4 UdC!.Enklin Research Center A Dwoon of The Frannan areasute

DEuTEc uATtmat m pmopmf7AfW INFORMADON TER-CS257-195 , environmental qualification information to the NRC on an accelerated l schedule. l In April 1980, the NRC organizational structure was modified and the Equipee- Qualification Branch was formed within the new Division of Engi-l neering. Responsibility for reviewing the status of equipment qualification l 1 for all plants was assigned to this branch. On May 27, 1980, the NRC issued Memorandum and Order CLI-80-21 [10],

specifying that Licensees and applicants must meet the requirements set forth  ;

l in the DCR Guidelines and NUREG-0588 regarding environmental qualification of safety-related electrical equipment in order to satisfy 10CFR50, Appendix A, General Design Criteria, Section I, Criterion 4. This Order also established i that the Safety Evaluation Reports on this subject, to be prepared by the NRC staff, must be issued on February 1,1981 and that all subsequent actions to l l be taken by licensees to achieve full compliance with the DOR Guidelines or l NUREG-0588 must be completed no later than June 30, 1982. l.3 SPECIFIC ISSUE BACKGROCND In a letter dated December 23, 1977, the NRC requested that Jersey Central Power & Light Company (JCPEL) review the status of environmental qualification for the safety-related electrical equipment at the Cyster Creek Nuclear Generating Station. Information requested from JCP&L included identification of electrical equipment required to perform s&fety functions while subjected l l to design basis accident environments, definitions of environmental service conditions at equipment locations, and the status of environmental qualification. In addition, documentation pertaining to qualification was to be compiled and organized for review by NRC. In response to this request, JCP&L provided information via submittals transmitted by letters dated February 24 and December 10, 1978. On March 10-13, 1980, NRC and FRC representatives visited the Cyster Creek plant, inspected safety-related systems and components, and discussed the program's requirements with JCP&L j representatives. JCPEL provided additional information in letters dated April 11 and May 7, 1980. NRC and FRC representatives held a subsequent meeting i l-5

     .d. Franklin Research Center 4 pmen e nen==memme                                                          ;

DELETED MATERIAL IS PROPRIETARY INFORMATION TER-CS257-195 with JCP&L representatives on October 9,1980. The electrical equipment requiring qualification (limited to that located within the primary containment), the plant's environmental service conditions, and the qualification documentation for the plant were identified at this meeting and in subsequent communications. FRC issued a Draf t Interim Technical Evaluation Report to NRC on October 24, 1980. Copies of the report were transmitted to JCP&L by the NRC. On August 29 and September 19, 1980, NRC notified JCP&L that all supplemental information on equipment environmental qualification must be submitted by November 1, 1980. On October 28, 1980, the Licensee sent the NRC a completely revised and expanded submittal of qualification information. 1.4 SCOPE OF THE EVALUATION Environmental qualification of safety-related electrical equipment was selected by the NRC for accelerated review. Therefore, the scope of this re por t is limited to equipment that must function to mitigate the consequences of a loss-of-coolant accident (LOCA) or high energy line break (RELB) and equipment whose environment is adversely affected by those events. Qualification aspects not included within the scope of this evaluation are: o seismic qualification o equipment protection against natural phenomena o equipment operational service conditions (e.g. , vibration, voltage, and frequency deviations) , o equipment located where it is subject to outdoor environments o equipment protection against fire hazards o equipment protection against missiles. g 1-6 uua04> FranMin s nw rm. .Resear.ch

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DELETED MATERIAL IS PROPRtETARY INFORMATION TER-C5257-195

2. NRC CRITERIA FOR ENVIRONMENTAL QUALIFICATION 2.1 CRITERIA PROVIDED BY THE NRC The DOR screening guidelines used by FRC to evaluate the electrical equipment environmental qualification programs were o " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (6) o " Guidelines for Identification of That Safety Equipment of SEP Operating Reactors for Which Environmental Qualification Is To Be Addressed" (6].

These guidelines were issued for implementation to all licensees by the NRC in February 1980. 2.2 STAFF POSITIONS AND SUPPLEMENTAL CRITERIA The NRC identified the following staff positions and supplemental criteria to be used in conjunction with the referenced DCR screening guidelines. 2.2.1 SERVICE CONDITIONS INSIDE CONTAINMENT FOR A LOSS-OF-COOLANT ACCIDENT (DOR Guidelines Section 4.1) For pressurized wdter reactors (PWRs) , the DOR Guidelines state that the containment temperature and pressure conditions as a function of time should be based on the most recent NRC-approved service conditions specified in the Final Safety Analysis Report (FSAR) or other licensee documentation. In the specific case of pressure-suppression type containments, the follcwing minimum high temperature conditions may be used: (1) boiling water reactor (BWR) drywells -- 340*F for 6 hours and (2) PWR ice condenser lower compartments -- 340*F for 3 hours. As stated in Supplement 2 to IE Bulletin 79-01B (8], "these values are a screening device, per the Guidelines, and can be used in lieu of a plant-specific profile, provided that expected pressure and humidity I conditions as a function of time are accounted for." l 2-1 4i!."J Franklin Research Center a come at w-en,en mense I

l DELETED M ATERIAL O PRoPRIETA%NFoRM ATQH TER-CS257-195 l Service conditions should bound those expected for coolant and steam line l breaks inside containment with due consideration given to analytical uncertainties. The steam line break condition should include superheated conditions, the peak temperature, and subsequent temperature / pressure profiles as functions of time. If containment spray is to be used, the impact of the spray on required equipment should be assessed. The adequacy of a plant-specific profile depends on the assumptions and design considerations at the time the profiles were developed. The DOR Guidelines and NUREG-0588 provide guidance and considerations required to determine if the calculated plant-specific temperature / pressure profiles encompass the LOCA and HELE accidents inside containment. 2.2.2 SUBMERGENCE (DOR Guidelines Section 4.1, Subitem 3; and Section 4.3.2, Subitem 3) Equipment submercence (inside or outside containment) should be addressed where the possibility exists that submergence of equipment may result from HELBs or other postulated occurrences. Supplement 2 to IE Bulletin 79-OlB [8] provides the following additional criterion: If the equipment satisfies the guidance and other requirements of the DOR Guidelines or NUREG-0588 for the LOCA and HELB accidents, and the licensee demonstrates that its failure will not adversely affect any safety-related function or mislead the operator after submergence, the equipment can be considered exempt from the submergence portion of the qualification requirements. 2.2.3 EQUIPMENT LOCATED IN AREAS NORMALLY NAINTAINEic AT ROOM CONDITIONS (DOR Guidelines Section 4.3.3) Supplement 2 of IE Bulletin 79-OlB [8] permits deferment of the review of environmental qualification for all safety-related equipment items located in plant areas where the equipment is not exposed to the direct effects of a HELB or to nuclear radiation emanating from circulation of fluids containing radioactive substances. At the licensee's option, the review may be deferred until after February 1, 1981. 4UjUU Franklin Research Center 2-2 m .ane w m .au.

i DELETED MATERIAL !s PROPRIETARY INFORMATION TER-C5257-195 By June 30, 1982, all safety-related electrical equipment potentially exposed to a harsh environment in nuclear generating stations licensed to

operate on or before June 30, 1982 shall be qualified to either the DOR Guidelines or NUREG-0588 (as applicable) . Safet.y-related electrical equipment is that required to bring the plant to a cold snutdown condition and to mitigate the consequences of the accident. It is the responsibility of the licensee to evaluate the qualification of safety-related electrical equipment to function in environmental extremes not associated with accident conditions and to document it in a f6rm that will be available for the NRC to audit.

Qualification to assure functioning in mild environments must be completed by June 30, 1982. 2.2.4 SIMULATED SERVICE CONDITIONS AND TEST DURATION (DOR Guidelines Section 5.2.1) The Guidelines require that the test chamber environment envelop the required service conditions for a time equal to the period from the initiation of the accident until the service conditions return to normal. Supplement 2 to IE Bulletin 79-013 [8] provides the following additional criterion:

  " Equipment designed to perform its safety-related function within a short time into an event must be qualified for a period of at least 1 hour in excess of the time assumed in the accident analysis. The staff has indicated that time is the most significant factor in terms of the margins required to provide an acceptaM.e confidence level that a safety-related function will be completed.

The 1-hour qualification requirement is based on the acceptance of a type test for a single unit and the spectrum of accidents (small and large breaks) bounded by the single test." 2.2.5 DEFERMENT OF QUALIFICATION REVIEW Supplement 3 to IE Bulletin 79-013 [91 permits the submittal of qualificaticn documentation regarding the TMI Action Plan equipment and the equipment required to achieve and maintain a cold shutdown condition to be delayed as follows:

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DELETED W ATERIAL O PROMUETARY INFORM ATION TER-C5257-195 o " Qualification information for installed TMI Action Plan equipment must be submitted by February 1, 1981. o Qualification information for future TMI Action Plan equipment (ref. NUREG-0737, when issued), which requires NRC pre-implerentation review, must be submitted with the pre-implementation review ccta. o Qualification information for TMI Action Plan equipment currently under NRC review should be submitted as soon as possible. o Qualification information for mI Action Plan equipment not yet installed which does not require pre-implementation review should be submitted to NRC for review by'the implementation date. o "he qualification information for equipment required to achieve and maintain a Cold Shutdown condition ... will be submitted not later than February 1, 1981." 2.2.6 TEST SEQUENCE (DOR Guidelines Section 5.2.3) Supplement 2 to IE Bulletin 79-01B [8] provides the following additional criteria:

     " Sequential testing requirements are specified in NUREG-0588 and the DOR Guidelines. Licensees must follow the test requirements of the applicable document.

1. If the test has been completed without aging in sequence, justification for such a deviation must be submitted.

2. If testing of a given component has been scheduled but not initiated, the test sequence / program should be modified to include aging.

3. Test programs in progress should be evaluated regarding the ability to comply by incorporating aging in the proper sequence. These would then fall in the first or second category."

2.2.7 RADIATION (DCR Guidelines Sections 4.1.2, 4.2.2, and 4.3.2, Subite= 2) Supplement 2 to IE Bulletin 73-OlB [8] provides the following additional criteria:

     "Both the DCR Guidelines and NUREG-0588 are similar in that they provide the methods for determining the radiation source term when considering 4.2 Frankhn Research Center                  2-4 s w an w m moue

DELETED MATEMlAL IS PHOPallETARY INFORMATION TER-C5257-195 LOCA events inside containment (1004 noble gases /50% iodine /lt partic-ulates). These methods consider the radiation source term resulting from an event which completely depressurizes the primary system and releases the source term inventory to the containment. NUREG-0578 provides the radiation source term to be used for determining the qualification doses for equipment in close pecximity to recirculating fluid systems inside and outside of containment as a result of LOCA. This method considers a LOCA event in which the primary system may not depressurize and the source term inventory remains in the coolant. NUREG-0588 also provides the radiation source term to be used for qualifying equipment following non-LOCA events both inside and outside containment (10% noble gases / lot iodine /04 particulates) . When developing radiation source terms for equipment qualification, the licensee must ensure consideration is given to those events which provide the most bounding conditions. The following table summarizes these considerations: LCCA Non-LOCA HELB Cutside Containment NUREG-0578 NUREG-0588 (100/50/1 (10/10/0 in RCS) (*] in RCS) Inside Containment Larger of NUREG-0588 NUREG-0588 (100/50/1 (10/10/0 in containment) in RCS) or NUREG-0578 (100/50/1 in RCS) , Gamma equivalents may be used when consideration of the contributions of beta exposure has been included in accordance with the guidance given in the DCR Guidelines and NUREG-0588. Ccbalt 60 is one acceptable gamma radiation source for environmental qualification of safety-related equipment. Cesium 137 may also be used." l

 *The numbers in parentheses represent t noble gases /t iodine /t particulates.

RCS means reactor coolant system. 46 2-5

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3. METHODOLOGY USED BY FRC The Licensee, Jersey Central Power and Light Company, listed an extensive l

( number of safety-related electrical equipment items in various locations of the Oyster Creek Nuclear Generating Station in its submittals to the NRC. FRC analyzed the Licensee's list and grouped together all identical equipment items located within plant areas that are exposed to the same environmental service conditions. This analysis reduced the list to 73 different equipment items to be reviewed. In this report, the term " equipment item" refers to a specific type of electrical equipment, designated by manufacturer and model, , which is representative of all identical equipment in a plant area exposed to the same environmental service conditions (e.g., Flow Transmitter, Fischer & Porter, Model 10B2496, located within containment) . Appendix A contains the environmental service conditions for each location, Appendix B contains a tabulation. of the equipment items and locations (the tabulation does not include equipment covered by the evaluation deferment described in Section 2.2.3 of this report), and Appendix C lists the plant systems identified by l the Licensee and the NRC as being essential to safety. 1 Using the list of safety-related electrical equipment items,* FRC reviewed each item in relation to: o NRC DOR Guidelines, as modified by NRC staff interpretations o Licensee definition of harsh service environments (Appendix A) o results of plant visit and equipment inspection o qualification documentation o analysis and/or justification of qualification o Licensee-proposed remedies for qualification deficiencies o Licensee-stated position concerning system or component function.

                   *In this report, the term " safety-related electrical equipment" refers to the equipment defined by the two NRC Guidelines referenced in Section 2.1.

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DELETED M ATERIAL t$ PROPRIETARY INFORMATION TER-C5257-195 Topics not within the ope of FRC evaluation are: o completeness of the Licensee's listing of safety-related equipment o acceptability of Licensee-provided environmental service conditions. The initial results of FRC's review of the equipment environmental documentation were issued to NRC as a Draft Interim Technical Evaluation Report (DITER) on October 24, 1980 [7). Qualification data summary forms used to summarize salient data compiled from the various information sources were included in the DITER. In developing the present final Technical Evaluation Report (TER) , FRC used the DITER and the Licensee submittals [1,2,3,4,5] . This information was analyzed by FRC to determine: o what specific response was made to the FRC DlTER o whether the Licensee made any changes to the initial submittal o what additional information was supplied (e.g., analysis, test report, or justification for qualification) o whether any changes were made in the environmental conditions o whether any equipment was added or deleted. All information wac eviewed by FRC for conformance to the NRC criteria referenced in Section 7 of this report. As requested by the NRC, all qualification information developed in the Equipment Environmental Qualification (EEQ) program was.used by the FRC reviewers, whether referenced by the Licensee or not. The qualification data summary forms were updated as appropriate and were then used to identify deviations from NRC criteria and the Licensee's qualification program. The final TER text was written primarily to address these deviations from the criteria. Items or test results not specifically cited by FRC implicitly satisfy the qualification criteria. Upon completion of the fir.al review for each equipment item, FRC developed an overall evaluation of the component and a specific conclusion 1 4 3-2 El Franklin Research Center aon, .erh. n.a inm.m

DELETED MATERfAL Q PRoPRIE7RY INFORMAflON TER-CS257-195 with respect to its qualification. At the NRC's request, recommendations were made to resolve questions of deficient qualification. Based on the FRC conclusion, each equipment item was assigned to one of the generic qualification categories provided by the NRC. The NRC category descriptions follow. NRC CATEGORIES AND DEFINITIONS o NRC Category I.a EQUIPMENT THAT SATISFIES ALL APPLICABLE REQUIREMENTS OF THE DOR GUIDELINES This category includes equipment items which are fully acceptable on the basis that all applicable criteria defined in the DOR Guidelines are satisfied and the equipment has been found to be qualified for the life of the plant. o NRC Category I.b EQUIPMENT WITH ACCEPTABLE DEVIATIONS FROM THE DOR GUIDELINES This category includes equipment items which do not satisfy one or more of the applicable criteria defiried in the DOR Guidelines; however, sufficient information has been presented to determine that the specific deviations are acceptable and the equipment has been found to be qualified for the life of the plant. o NRC Category II.a EQUIPMEtC THAT SATISFIES ALL APPLICABLE REQUIREMENTS OF THE COR GUIDELINES WITH THE EXCEPTION OF QUALIFIED LIFE This category includes equipment items that are acceptable on the basis that all applicable criteria defined in the DOR Guidelines are satisfied with the exception of the qualified life criterion. With respect to qualified life, the equipment items have been found to have a qualified life which (1) is limited to a time interval less than plant life, (2) has not been adequately established in terms of calendar time, or (3) has not been evaluated by the licensee. o NRC Category II.b EQUIPMENT THAT SATISFIES ALL APPLICABLE REQUIREMENTS OF THE DOR GUIDELINES WITH THE EXCEPTION OF QUALIFIED LIFE This category includes equipment items which will be acceptable and will satisfy all applicable criteria defined in the DOR Guidelines with the exception of qualified life, provided that specific modifications are made on or before the designated date. When the modifications are complete, the equipment can be considered qualified with the exception of the qualified life criterion. With respect to qualified life, ene equipment items have been 3-3

    .50 Franklin Research Center
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DELETED MATERIAL 23 PROPRIETAZf INFORM ATION TER-CS257-195 found to have a qualified life which (1) is limited to a time interval less than plant life, (2) has not been adequately established in terms of calendar time, or (3) has not been evaluated by the licensee, o NRC Category II.c EQUIPMENT FOR WHICH DEVIATIONS FROM THE DOR GUIDELINES ARE JUDGED ACCEPTABLE WITH THE EXCEPTION OF QUALIFIED LIFE . This category includes equipment items which do not satisfy one or more of the applicable criteria defined in the DOR Guidelines; however, either (1) sufficient bases have been presented to allow a determination that the specific deviations are judged to be acceptable with the exception of qualified life criterion, or (2) the specific deviations are judged to be acceptable with the exception of qualified life criterion, based on review of the applicable qualification documentation associated with the overall equipment environmental qualification program. With respect to qualified life, the equipment items have been found to have a qualified life which (1) is limited to a time interval less than plant life, (2) has not been adequately established in terms of calendar time, or (3) has not been evaluated by the licensee. o NRC Category III EQUIPMENT THAT IS EXEMPT FROM QUALIFICATION This categor" includes equipment items which are exempt from qualifi-cation on ti. : t ais that (1) the equipment does not provide a safety function (i.e., shorid not have been included in the equipment list submitted by the licensee), or (2) the specific safety-related function of the equipment can be accomplished by some other designated component which is fully qualified. In addition, any failure of the exempt equipment must not degrade the ability of qualified equipment to perform its required safety-related function. o NRC Category IV.a EQUIPMENT THAT HAS QUALIFICATION TESTING SCHEDULED BUT NOT COMPLETED The qualification of equipment items in this category has been judged deficient or inadequate based upon review of the documentation provided by the licensee. However, the licensee has stated that the equipment item is scheduled to be tested by a designated date. The results of the testing will dictate the specific qualification category of the equipment item. o NRC Category IV.b EQUIPMENT FOR WHICH QUALIFICATION DOCUMEN"'ATION IN ACCORDANCE WITH THE GUIDELINES HAS NOT BEEN ESTABLISHED The qualification of equipment items in this category is deficient or inconclusive based upon review of the documentation provided by the licensee. This equipment is judged to have a high likelihood of operability for the specified environmental service conditions; however, complete and auditable 4_ 3-4 W.0 Franklin Research Center A Dwoon of N Frannan meetute

DELETED MATERIAL 13 PROPRIETARY INFORMATION TER-CS257-195 records reflecting comprehensive qualification documentation have not been made available for review. o NRC Category V EQUIPMENT THAT IS UNQUAI.IFIED The DOR Guidelines require that complete and auditable records reflecting a comprehensive qualification methodology and program be referenced and made available for review of all Class IE equipment. The qualification of equipment items in this category has been judged to , be deficient or inadequate, based upon review of the documentation provided iy the licensee. The extent to which the equipment items fail to satisfy the criteria of the DOR Guidelines can be categorized as follows: (1) documen-tation reflecting qualification as specified in the DOR Guidelines has not been made available for review, (2) the documentation is inadequate, or (3) the documentation indicates that the equipment item has not passed the required tests. o NRC Category VI EQUIPMENT FOR WHICH QUAIIFICATION IS DEFEP. RED This category includes equipment items which have been addressed by the licensee in the equipment environmental qualification submittals; however, the l qualification review of this equipment has been deferred by the NRC in [ accordance with criteria presented in Sect. ions 2.2.3 and 2.2.5 of this report. l [ l l l ( i l l 4JJ Franklin Research Center 3-5 a Onomen af ?>e Fw ineeswe

DELETED MATERIAL IS PROPRIETARY INFORMATION TER-C5257-195

4. TECHNICAL EVALUATION General observations concerning the Licensee's approach to qualification are included in Section 4.1. Sections 4.2 through 4.7 identify the equipment items placed in each of the major NRC qualification categories in accordance with FRC's technical evaluation of the Licensee's documentation. The results of the evaluation are summarized in Section 4.8.

The technical evaluation of each equipment item is documented in the following formats o Original Text Taken From Draf t Interim Technical Evaluation Report o Licensee Response o FRC Evaluation o FRC Conclusion. _ All equipment item

• numbers are associated wi'.h Reference 1.

4.1 METHODOLOGY USED BY THE LICENSEE The final submittal of electrical equipment qualification documentation from the Licensee [1] was well organized and addressed the basic qualification requirements by means of system, equipment, and environmental analysis techniques. An FRC review of the documentation provided by the Licensee has generated the following observations.

                *In this report, the term " equipment item" refers to a specific type of electrical equipment, designated by manufacturer and model, which is representative of all identical equipment in a plant area exposed to the same environmental service conditions (e.g., Flow Transmitter, Fischer & Porter, Model 10B2496, located within containment) .

4-1 1 O l i JU) aFranklin ca n w th r Research n.en - u. C. enter l

DELETED MATE 2iAL 0 PROPRIETA0Y INFORM ATION TER-CS257-195 4.1.1 COMPLETENESS OF EQUIPMENT LIST In the final submittal, the Licensee provided information for a large number (approximately 200) of equipment items. (Tne previous submittal [3] considered only equipment in the drywell of the primary containment.) The

 . Licensee's equipment item list included only those safety-related electricol equipment items that are (i) installed in potentially " harsh" areas and (ii) needed for hot shutdown. The Licensee has elected to defer the review of equipment installed in " mild environments" and items needed for cold shutdown until after February 1, 1981, as discussed in Sections 2.2.3 and 2.2.5, and is continuing to assemble and review qualification information for these equipment items.

In Reference 1, the Licensee presented System Component Evaluation Work (SCEW) sheets for each safety-related equipment item for which the review is not deferred. These sheets summarize the pertinent environmental service conditions and identify available documentation references. FRC has analyzed the information in the SCEW sheets and has compiled a list of 73 equipment item groupings (henceforth referred to as " equipment items") for review in this Technical Eva'uation Report. These equipment items consist of identical units having similar operational requirement s and exposed to similar environmental conditions. Discussions with the Licensee have indicated that motor control centers and possibly some switchgear have been overlooked as safety equipment located I in " harsh" areas and required for hot shutdown. The Licensee stated that a revision to its most recent submittal [1] would be tran,smitted to rectify the oversight. The Licensee should also investigate the torus vacuum relief valve system to determine whether the vacuum relief valve solenoid and the differential pressure transmitter should be qualified. In addition, the Licensee should verify that no safety-related connectors or terminal blocks are located outside of the containment drywell. 4-2 p2 L. Franklin Research Center

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l DELETED MATERIAL IS PROPRIETAAY INFORMATION TER-C5257-195 4.1.2 ENVIRONMENTAL SERVICE CONDITIONS 4.1.2.1 TDtPERATURE AND PRESSUR2 PROFILES FOR THE CONTAINMENT DRYWELL The Licensee states: The Oyster Creek containment temperature and pressure profile to be used for the environmental qualification of electrical equipment inside containment is derived from the most severe MSL break response with heat sinks and containment spray considered. This is the 0.75 ft2 MSL break analysis. The results for this case are repeated in Figure 7-1 [ Figure A-3 in Appendix A] . This plant-specific analysis represents a significant reduction from the 340*F for 6 hours recommended in NUREG-0588. The major reasons for the departure from the NUREG-0588 generic profile are the consideration of containment heat sinks and the initiation of containment spray. The Licensee submittal indicates that Oyster Creek Station has automatic / manual and redundant drywell containment sprays that can provide the long-term drywell heat sink and reduce the drywell temperature and pressure. The Licensee's drywell analysis used conservative energy release data and heat removal parameters, together with the assumption that the spray would be initiated 10 minutes af ter the break occurs. The NRC has reviewed the analysis and concurs that the MSL3 accident analysis sets the limiting drywell service condition (11).

 ..l.2.2     TEMPERATURE CONDITIONS IN THE REACTOR BUILDING The Licensee has conducted extensive analysis to determine the environmental service conditions to which the safety-related electrical
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equipment needed for hot shutdown would be exposed in the event of postulate

• MSLB and HELB accidents. For areas where this equipment was located and tLe temperatures will exceed 100*F, temperatures as functions of time were presented in graphical form, and the peak temperature, pressure, and radiation levels were listed in Table 1. This information is included in Appendix A of f this report.

4.1.2.3 RADIATION DOSE The Licensee provided a description of methods for calculation and evaluation of dose values. Key statements from Reference 1 on the methodology are quoted below: i l E. nklinwRes,earc.h.C.

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DELETED MATERIAL is PRoPR6ETARY INFORMATION TER-C5257-195 Analytical Methodology - EDS has calculated post-accident radiation exposures to vital equipment located inside the Oyster Creek containment due to airborne contamination and reactor vessel streaming. In addition, the radiation exposure contribution due to the station's normal forty-year operation has been considered. These calculations were performed by using the computer program QAD-PSA ... and simplified manual techniques. In all instances, the accident case source terms provided by JCP&L were utilized. In order to calculate radiation exposures inside containment due to reactor vessel streaming, a one-hour post-accident source term composed of one hundred percent each of the noble gases, halogens, and the remainder isotopes was calculated using the source term data supplied by JCP&L. This source term was distributed within the region defined by the active volume of the fuel resulting in the reactor vessel source model input into the computer program QAD-P5A. Appropriate shielding credit was taken for the reactor vessel wall, the coolant within the reactor vessel, the self-shielding afforded within the fuel region, and the biological shield wall. The calculated exposure rates outside the biological shield were held constant for forty years to determine the normal operation lifetime exposure. One-year post-accident integrated exposures were determined by applying an integration factor that accounted for the fission product radioactive decay during the one-year period following the accident. Analysis Results and Discussion Results of the inside containment exposure calculations due to reactor vessel streaming are shown in Table No.1 for.several locations and two electrical connector penetration lead shie'd thicknesses. The values shown indicate the normal operation forty-year lifetime exposure, the one-year post-accident integrated exposure, and the total. It should be pointed out that the reactor vessel streaming exposures presented here

   .      are for containment locations external to the biological shield.

, Exposures inboard of the biological shield would be considerably greater. i Table No. 2 indicates the results of calculations performed to determine l radiation exposures inside the containment due to post-accident airborne activity. Values are presented for both the electrical connector I pc..<cration area and for a point midway between the outer biological shield wall and inner drywell wall. No credit has been taken for the lead end shields supplied with the penetrations as discussed in Section 3.0. l l i l I l l 1 4 4-4 Udu Fmnklin Research Center .

             = >=.n s tw r   mo-a.,                                                            ;

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OsLETED MATEmAL IS PROPMETARY MORMADON TER-C5257-195 FRC agrees with the methodology applied by the Licensee. However, it appears that only the gamma contributions were listed in the tables. For beta-sensitive items such as electrical cables and pe: haps some other items, the sum of the gamms plus beta dose should have been provided and identified. 4.1.2.4 " MILD AREA" ASSUMPTIONS FOR REACE R BUILDING AND TURBINE BUILDING The plant's environmental study of MSLB/HELB occurrences outside the containment identified several areas in the reactor building and turbine building which during ar. accident would not experience any change in temperature and pressure from the normal ambient conditions. From FRC's review of Reference 1, it was not clear Gether the HVAC systems were assumed to be operational in order for the environment.1 service conditions of pressure and temperature to remain essentially normal in several areas. If the HVAC systems were assumed to be operational, then they are required by the DOR Guidelines to be redundant and powered from emergency electrical power systems. FRC has not had the opportunity to deterrine if redundant HVAC systems are available. The Licensee should either show that HVAC system operation was not assumed for the environmental calculations or provide evidence that the HVAC systems are redundant and fed by emergency power. 4.1.3 AGING AND QUALIFIED LIFE The Licensee has not adequately addressed the related topics of aging and qualified life. The DOR Guidelines require that the Licensee: o establish (numerically) the qualified life for all equipment items containing components susceptible to degradation produced by heat and nuclear radiations o implement programs to review detailed surveillance and maintenance records to assure that equipment that exhibits age-related degradacion is identified and replaced (or modified) as necessary. Qualified life is the maximum time of normal service, under specified conditions, for which it can be demonstrated that the functional capability of the equipment at the end of the period is still adequate for it to perform its specified safety function (s) for applicable design basis even:s. The 1 4-5 4xjd Franklin Research Center l 4 o- a# The w m.en

OELETED M ATERI AL 0 PROPRIETARY INFoRMATioN TER-CS257-195 qualified life may be contingent on implementation of a specified maintenance program. It is acceptable for the qualified life of some subcomponents of an equipment item to be less than the qualified life of the item itself, provided a program for replacement of such subcomponents at intervals not exceeding their qualified lifetimes is specified and fulfilled. The qualified life of an equipment item may be changed during its installed life when justified by new information that permits a reanalysis of the qualification program. Establishing the qualified life for equipment is a technically challenging task because of the paucity of information concerning the degradation of materials and components under the long-term exposure to the environmental service conditions in a nuclear power generating station. As is discussed more fully in Reference 13, with the possible exception of certain simple materials, there is no rigorous basis for establishing equipment qualified lifetimes for periods approaching an installed lifetime of 40 years. Furthermore, applicable information regarding possible long-term synergistic effects of temperature, humidity, nuclear radiations, etc. is extremely limited. On virtually every SCEW sheet in Reference 1, the Licensee has stated (next to the parameter " Aging") a value of 40 years under both the " Specification" and " Qualification" headings. Presumably, these entries are intended as the qualified life. In accordance with the Guidelines in this program, the licensees are required to establish a qualified life for equipment subject to thermal and radiation aging. In addition, surveillance, maintenance, and replacement programs should be established for equipment that may be subject to age-related degradation. The licensees should review the qualified life values and the present installed life of the equipment in accordance with the DOR Guidelines to determine a replacement schedule for each equipment item (or subcomponents thereof). As noted above, these schedules may be revised as new information becomes available. 4 4-6 f)J Franklin Research Center A Dmeien of The F eruan inename

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DELETED MATERIAt. IS PROPRIETARY !NFORMA7)ON TER-C5257-195 4.2 EQUIPMENT QUALIFIED FOR PLANT LIFE This section includes equipment items which are fully acceptable on the basis that (1) all criteria defined in Section 2 of this report are satisfied or (2) sufficient data exist to determine that specific deviations are acceptable. 4.2.1 NRC Category I.a QUIPMENT THAT FULLY SATISFIES ALL APPLICABLE REQUIREMENTS OF THE DOR GUIDELINES The equipment items in this section are fully acceptable on the basis that all applicable criteria defined in the DOR Guidelines are satisfied and the equipment has been found to be qualified for the life of the plant. For the Oyster Creek Station, no equipment falls within this category. 4.2.2 NRC Category I.b EQUIPMENT WITH ACCEPTABLE DEVIATIONS FROM THE. DOR GUIDELINES The equipment items in this section do not satisfy one or more of the applicable crit.eria defined in the DOR Guidelines; however, sufficient information has been presented e.o determine that the specific deviations are acceptable and the equipment has been found to be qualified for the life of the plant, j For the Oyster Creek Station, no equipment falls within this category. i l I g 4-7 MJ Franklin Research Center s m anan w m m mse

k I i DELETED W ATERIAL IS F Ro#RIETARY WFORM ATioN l l TER-CS257-195 4.3 EQUIPMENT QUALIFIED WITH RESTRICTIONS This section includes equipment items that are acceptable on the basis that (1) all criteria defined in Section 2 of this report are satisfied with the exception of the qualified life criterion; (2) the equipment requires specific modification which, when completed, will establish full qualification with the exception of satisfying the qualified life criterion; or (3) with the exception of satisfying the qualified life criterion, deviations from the criteria presented in Section 2 have been found to be acceptable. 4.3.1 NRC Category II.a EQUIPMENT THAT SATISFIES ALL APPLICABLE REQUIREMENTS OF THE DOR GUIDELINES WITH THE EXCEPTION OF QUALIFIED LIFE The equipment items in this section are fully acceptable on the basis that all applicable criteria defined in the DOR Guidelines are satisfied with the exception of the qualified life criterion. With respect to qualified life, the equipment items have been found to have a qualified life which (1) is limited to a time interval less than plant life, (2) has not been adequately established in terms of calendar time, or (3) has not been evaluated by the Licensee. 4.3.1.1 Equipment Item No. 2 Solenoid valves Located in the Reactor Building ASCO Model NP-8344A70E Drywell Vent and Purge Valves (V-26-16 and V-26-18) (Licensee Reference 2.24) l I ORIGINAL TEXT TAKEN FRCM DRAPT INTERIM TECHNICAL E"/ALCATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These valves are ASCO Model NP-8344A70E and are qualified for LOCA environment. The results and description of the test are given in the ASCO Test Report No. ACS 21678/Tr, Revision A, dated March 1978. I g 4-8 M Frankhn Research Center

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DELETED MATEMIAL t$ PAOPRIETARY INFORMATION TER-C5257-195 FRC EVALUATION: FRC has reviewed Reference 2.24 and has the following comments:

1. During the qualification test program described in the reference, The test results must therefore be regarded as inconclusive until the uncertainties associated with the method of making the wiring interface with the solenoid, both in the plant and in the test, are resolved. The Guidelines state (Section 5.2.5):

               "If a component fails at any time during the test, even in a so called ' fail safe' mode, the test should be considered inconclusive with regard to demonstrating the ability of the component to function for the entire period prior to the failure."

They further state (Section 5.2.6):

               "The equipment mounting and electrical or mechanical seals used during the type test should be representative of the actual installation for the test to be considered conclusive."

However, because the environmental service conditions resulting from a HELB accident do not involve extremely high temperature, large radiation doses, or liquid spray, the deficiencies in the test are not of concern for this equipment item. The environmental parameters of the test program exceed by wide margins the plant-specific l environmental service conditions stated by the Licensee. However, no . justification for the Licensee's stated ambient temperature of only 77*F was given in the Reference 1 SCEW sheet. FRC notes that this value is lower than any other cited on the SCEW sheets, and does not correspond to HELB environmental conditions.

2. The pre-aging simulated in the test program was intended to represent )

an installed life (and hence a qualified life) of ambient temperature. The ambient temperatures at the instelled locations within the plant are lower, and hence the qualified life is longer. The Licensee has not provided any justification for the claimed 40-year qualified life. An explicit, conservative determination of qualified life and replacement schedule (if needed) should be established. 4-9 4 AUJa Franklin o= a av n. Research

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DELETED MATERIAL O PROPRIETARY INFORMATioN TER-CS257-195 FRC CONCLUSION: This equipment is assigned to NRC Category II.a because a substantial period of qualified life and the ability to withstand the Licensee-stated HELB conditions at the installed location have been demonstrated. The Licensee should review the stated environmental conditions and establish a conservative qualified life. A surveillance program to monitor performance and identify any degradation requiring maintenance or replacement should also be implemented. 4.3.1.2 Equipment Item Nos. 49 and 50 (previously designated I6) Electrical Cable Located Within the Drywell 49: General Electric Model S1-58145 Vulkene 50: General Electric Model S1-58073 Vulkene (Original Licensee References 2.7, 2.11, 2.12, and 2.18; Final Licensee References 2.16 and 2.21) ORIGINAL TEXT TAKEN FROM DRAET INTERIM TECHNICAL EVALUATION REPORT (3.3.2.5): Reference 2.7 is discussed in Subsection 3.3.2.3 [4.5.2.17 in this re port] . The connector tests described therein used cables removed from the

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plant. Reference 2.12 is a letter from General Electric stating that the Oyster Creek plant has two types of No. 12 AWG GE Vulkene cable installed "inside the containment" (FRC presumes this to mean within the drywell). This letter further states that the installed cable has an insulation thickness of 0.047 inch and that this is adequately represented by the No.12 AWG GE Vulkene Type SIS cable included in the test program of the electrical penetrations conducted by GE in February 1975. The letter notes that the i cables in the test program had an insulation thickness of 0.031 inch, and l therefore the installed cable, having thicker insulation, "is considered qualified for the LOCA environment." The report of the penetration tests was not provided for review, so this reference must be regarded as irrelevant. Reference 2.18 is a report of a test performed on No. 12 AWG GE Vulkene cables removed from the Pilgrim Unit 1 plant and spliced. FRC comments are:

a. Although it appears that the tested samples are the same as the installed ones, complete documentation to substantiate this has not been provided. The Licensee should submit a listing of the type of cable (manufacturer, construction, materials) used for each item of Class lE equipment within the drywell and provide complete documentation to relate this to valid test reports.

4 4-10 h0J Franklin Research Center A Dneman of The Frannhn wtseame l 1

i DELETIO MATERiAus PROPMETARY INFORMADON TER-C5257-195

b. Neither test report included nuclear radiation exposures or consideration of aging. The thermal environmental parameters during the tests were adequate to represent plant-specific DBE conditions.

LICENSEE RESPONSE: [No response provided.] FRC EVALUATION: The Licensee SCEW sheet identified the cable by specific type and added FIRL Report F-C4497-2 as evidence of qualification. FRC has reviewed the information provided by the Licensee, as well as the additional reference, and has the following conunents:

1. The SCEW sheets 1-6A and 1-6B describe the cable installed in the drywell and relate it to the FIRL Report F-C4497-2 [2.211, resolving comment (a) of the DITER.

2. The cable tested in FIRL Report F-C4497-2 was pre-aged and irradiated to 200 Mrd, resolving comment (b) of the DITER.

FRC CCNCLUSION: This equipment is assigned to NRC Category II.a because qualification has been demonstrated by test, except for qualified life. The Licensee should establish a conservative qualified life (see Section 4.1.3) . 4.3.1.3 Equipment Item No. 53 (previously designated I7) Electrical cable Located Within the Drywell Rockbestos, Model Not Stated (Final Licensee References 2.15 and 2.16) ORIGINAL TEXT TAKEN FRCM DRAFT INTERIM TECHNICAL EVALUATICN REPORT (3.2.2) : Reference 2.15 is a manufacturer's qualification test report for three types of Rockbestos Firewall III cable (single conductor #16, #12, and 46 AWG). The first of these is stated to be instrumentation cable. The samples were thermally aged at 302*F for 1300 hours, which was intended to simulate 40 years " aging" in the plant at 194'F. The pre-aged cables were irradiated to 200 Mrd (gamma), and then exposed to a steam / chemical spray / moist atmosphere 4-11 4

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DELETED M ATERIAL 23 PROPRIETARY IN FORMATION TER-C5257-195 environment. Peak conditions were 346*F/ll3 psig steam for 3.6 hours; the total duration of the test was 140 days (30 days with steam plus 100 days at 200 *F/100 % RH) . The cables were sprayed during the first 24 hours of the steam exposure with a solution of boric acid and sodium hydroxide. These conditions envelop the Licensee's expected MSLB and LOCA profiles by wide margin 2. The use of a different chemical solution in the spray is not regarded as i significant deficiency. Current and voltage loadings of the cable samples were applied during the 30-day steam exposure. FRC concludes that this report establishes the environmental qualification of this equipment item according to the requirements of the Guidelines. This conclusion does not imply concurrence in the Licensee's implied claim that a 40-year qualified life has been established. The Arrhenius plot is based upon mechanical property data, and no information is presented to relate this to long-term electrical performance. The thermal aging exposure and the simulated LOCA exposure are both very severe, however. As a consequence, high confidence can be placed in the performance of the cable, and the qualified life can be expected to be quite long. LICENSEE RESPONSE: [No response provided.) FRC EVALUATION: As the Licensee provided no additional information, the original comments still apply. FRC CONCLUSION: This equipment is assigned to NRC Category II.a because qualification has been demonstrated by test except for qualified life. The Licensee should establish a conservative qualified life (see Section 4.1.3) . 4-12 4dLJ FmnWin Renan:h Center l i

          =>   a w n= r   m emw.                                                         l l

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DELETED MATERIAI.ls PROPRIETARY INFORM AT10N TER-C5257-195 4.3.1.4 Equipment Item Nos. 31A and 32A Solenoid Valves Located in the Steam Tunnel 31A: ASCO Model 206-832-3RU 32A ASCO Model 206-301-3RU MSIV Solenoid Valvas and MSIV Position Indicators (Final Licensee Reference 2.24) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONCE (EQUIPMENT ITEM ADDED IN REFERENCE 1): 31A NS-0 4A-L1, -L2, -L3 32A: NS-0 4B-L1, -L2 The MSIV solenoid valves are used to direct instrument air to hold open the outside containment main steam isolation valves. The MSIV position indication switches are utilized to provide a scram signal when the MSIVs are less than 90% open. A loss of power or air to the MSIV solenoids causes the MSIVs to fail in the safe direction, closed. Also redundant protection is provided by the inside containment isolation valves that would not be affected by the environment created by outside containment breaks. In the event the outside containment MSIV position switch did not provide a scram signal, two scram signals would still be available to ensure the reactor was shut down immediately for a MSLB. These two signals are the MSIV position switch signal from the inside valves and the reactor low water level signal, both of which would not be affected by the harsh environment created during this event. The one-year integrated accident exposure of these components is at least two orders of magnitude below that which would cause any degradation. Based upon the above discussion, it is expected that the main steam isolation function and reactor scram function required to mitigate MSLB outside containment will be accomplished. FRC EVALUATION: FRC has reviewed Reference 2.24 and has the following comments:

1. During the qualification test program described in the reference, 4 4-13 fjdu Franklin a c- e m Research nen a m. C. enter

DELETED MATEmiAL 10 PROPRIETA;Y INFOAMATION TER-CS257-195 The test results must therefore be regarded as inconclusive until the uncertainties associated with the method of making the wiring interface with the solenoid, both in the plant and in the test, are resolved. The Guidelines state (Section 5.2.5) :

                     "If a component fails at    any time during tne test, even in a so called ' fail safe' mode,    the test should be considered inconclusive with regard    to demonstrating the ability of thu component to function for    the entire period prior to the f ailure."

They further state (Section 5.2.6):

                     "The equipment mounting and electrical or mechanical seals used during the type test should be representative of the actual installation for the test to be considered conclusive."

However, because the environmental service conditions resulting from a HELB accident do not involve extremely high pressure, large radiation doses or liquid spray, the deficiencies in the test are not of concern for this equipment item. The environment.al parameters of the test program exceed by wide margins the plant-specific environ-mental service conditions stated by the Licensee.

2. The pre-aging simulated in the test program was intended to represent an installed life (and hence a qualified life) of ambient temperature. The ambient temperatures at the installed locations within the plant are lower, and hence the qualified life is longer. An explicit, conservative determination of qualified life and a replacement schedule (if needed) should be established.

FRC CONCLUSION: This equipment is assigned to NRC Category II.a because a substantial period of qualified life and the ability to withstand the Licensee-stated HELB conditions at the installed location have been demonstrated. The Licensee should review the stated environmental conditions and conservatively establish the qualified life (see Section 4.1.3) . 4 4-14 b Franklin Research Center A Dawesen of The Peervian ansamme

CinTED W. Tamas. s PROPMt(T/.AY INPOMWATION TER-C5257-195 4.3.1.5 Equipment Item Nos. 4A and 34A Motor'-ed Valve Actuators Located in the Reactor Building 4A: ..amitorque Model SMB-000 34A: Limitorque Model SMB-0 Spray and Cleanup Valves (Final Licensee References 2.2, 2.3, 2.4, and 2.5) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPME:ff ITDI ADDED IN REFERENCE 1): 4A: Spray Valves (V-5-167 and V-5-147) Following a worst-case line break (cleanup system line break outside drywell), these valves will remain in a non-harsh environment (95'F/16 psia). Further, these valves are not required to mitigate a cleanup system line break outside containment. 34A: Cleanup va his (V-16-2, -14, -61)

          "Limitorque Qualified" FRC EVALUATION:

1. Reference 2.2 is a letter from Limitorque stating that the test program in Reference 2.3 is applicable to this equipment item. However, with regard to Reference 2.4, Reference 2.2 states:

          "Unfortunately, due to the date of supply,, our records are not completely clear; however, we believe that our Qualification Report B0003 can be used to support the capability of the actuators to withstand irradiation."

2. Reference 2.3  !*, a report of a qualification test program conducted on a The test program consisted of a 12-hour exposure to warm air saturated with water vapor ( and ). The performance of the actuator was monitored by cycling under load during the exposure (plus cycles before and af ter the exposure), and measuring the exposure). Performance was satisfactory, but the (There were no pre-aging, chemical spray, or nuclear radlation exposures in tho test program.)

4-15 4L Franklin Research Center 1 i w e w m .amne

GELETED MATEMIAL 0 PMoPR; ETA'N INFORM ATION TER-C5257-195

3. Limitorque Report B0003 [2.4) describes a qualification test program conducted on an SMB-0 MVA having a Reliance motor with Class B insulation, plus two additional motors. The MVA and motors were thermally aged and imultaneously operated (200 hours at 165'F; operation for 30 seconds in each direction once per hour for 176 hours), and the MVA was then operated for an additional 1817 cycles to simulate wear aging (the two extra motors were operated for in additional 15 minutes while the motors were unloaded) . The MVA then received a nuclear radiation dose of 20 Mrd, and the motors 204 Mrd.

Subsequently, the MVA and motors were seismically tested and subjected to a 16-day steam exposure test. Functional operation was demonstrated prior to and on five occasions during the latter exposure, the last immediately preceding the end of the test. Insulation resistance to ground was measured at each of these times. The MVA malfunctioned once (at 25.8 hours, just after the ambient temperature had been reduced from 250*F to 200*F) . This mal-function was attributed to a "a momentary electrical short due to localized condensate buildup, a malfunction of the reversing contactor, or a combination o f bo th . " The IR readings decreased with time at each of the two temperature plateaus of the steam exposure, but ac current draw was not significantly affected. The manufacturer concluded that "this test generically qualifies Limitorque Valve Actuaturs type SMB/SB for Class lE Service outside primary containment for conditions as defined in this report." However, as noted in paragraph 1 above, Limitorque believes but cannot verify that this reference is applicable to the present evaluation.

4. Refers. ice 2.5 is a letter from Limitorque that"provides a general statement attempting to justify a 40-year qualified life based on the pre-aging exposures that were applied in the test programs. Because the applicabilf 't Reference 2 4 is uncertain, and because there was no pre-aging in the test procram reported in Reference 2.3, this letter appears irrelevant to the present evaluation. The Licensee should evaluate the susceptibility of the materials in the MVA to aging degradation and establish the conservative qualified life (refer to Section 4.1.3 for additional comments).

4 4-16 Ed Franklin

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                   .n          C. enter

onerno warsmaus PmoemeTany neonmation - TER-C5257-195

5. Because the environmental service conditions during an accident do not deviate appreciably from normal non-accident conditions, FRC considers that Reference 2.3 satisfies the Guidelines requirements, except for qualified life.

FRC CONCLUSION: This equipment is assigned to NRC Category II.a because the Guidelines requirements are satisfied except for qualified life. The Licensee should establish a conservative qualified life (see Section 4.1.3) . 4.3.2 NRC Category II.b EQUIPMENT THAT SATISFIES ALL APPLICABLE REQUIREMENTS OF THE DOR GUIDELINES WITH THE EXCEPTION OF QUALIFIED LIFE PROVIDED THAT SPECIFIC MODIFICATIONS ARE MADE

The equipment items in this section will be acceptable and will satisfy l all applicable criteria defined in the DOR Guidelines with the exception of qualified life provided that specific modifications are made on or before the designated date. When the modifications are complete, the equipment can be considered qualified with th( exception of the qualified life criterion. With respect to qualified life, the equipment items have been found to have a qualified life which (1) is limited to a time interval less than plant life, (2) has not been adequately established in terms of calendar time, or (3) has not been evaluated by the Licensee.

For the Oyster Creek Station, no equipment falls within this category. ! 4.3.3 NRC Category II.c EQUIPMENT FOR WHICH DEVIATIONS FROM THE DOR GUIDELINES ARE JUDGED ACCEPTABLE WITH THE EXCEPTION OF QUALIFIED LIFE The equipment items in this section do not satisfy one or more of the l applicable criteria defined in the DOR Guidelines; however, e.ther (1) sufficient bases have been presented to allow a determination that the l specific deviations are judged to be acceptable with the exception of the qualified life criterion, or (2) the specific deviations are judged to be l acceptable with the exception of the qualified life criterion based on a review of the applicable qualification documentation associated with the i 4 4-17 l .2J Franklin Researen Center mame nm I l

DELETEDtATERAL 33 PRoPRIETA%Y WORMATION TER-CS257-195 overall equipment environmental qualification program. With respect to qualified life, the equipment items have been found to have a qualified life which (1) is limited to a time interval less than plant life, (2) has not been adequately established in terms of calendar time, or (3) has not been evaluated by the Licensee. 4.3.3.1 Equipment Item Nos. 3A, 3B, 4B, and 34B Motorized Valve Actuators Located in the Reactor Building 3A and 3B: Limitorque Model SMB-00 Containment Spray Valves 4B: Limitorque Model SMB-000 Containment Spray Valves 34B: Limitorque Model SMB-0 Core Spray Valves (Final Licensee References 2.2, 2.3, 2.4, and 2.5) ORIGINAL TEXT TAKEN FROM DFAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPME!C ITEM ADDED IN REFERENCE 1): 3A Containment Spray valves (V-21-5 and V-21-11) The peak temperatures seen by these valves are 140*F (V-21-5) and 250*F (V-21-ll) following a cleanup system line break outside the drywell. However, these valves are not required to mitigate a line break outside the drywell. If a line break is inside the drywell, the valves located outside the drywell will not experience a high temperature or pressure. The valves will only see a rise in radiation level. However, these valves are normally open and will stay open even if the valve operator is de-energized. Therefore, the ability of the system to be used for drywell and torus cooling will not be affected. 3B: Containment Spray Valves (V-21-1, -3, -7, -9) The peak temperature and pressure seen by these valves following a worst case line break (a MSLB outside drywell) will be 165*F and 15 psia. 4B: Spray Valves (V-21-13 and V-21-17) Valve V-21-17 will not be affected by the break and thus will remain in the non-harsh environment (77'F and 15 psia) . The other valve (V-21-13) will experience a peak temperature of 140*F. However, these valves are not required to mitigate a line break outside drywell. In any case, these valves are normally open and will stay open even if the valve 4 4-18 2 Frankhn

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DELETED MATERIAL 18 PRoPMIETARY INPORMATION TER-C5257-195 operator fails to function. Therefore, the ability of the system to be used for drywell and torus cooling will not be affected. 34B: Shutdown Cooling Valves (V-17-1, -2, -3, -55, -56, -57)

      "Limitorque Qualified" FRC EVALUATION:

FRC has reviewed the references cited by the Licensee and has the - following corronents: ,

1. Reference 2.2 is a letter from Limitorque stating that the test program in Reference 2.3 is applicable to this equipment item. However, with regard to Reference 2.4, Reference 2.2 also states:

      "Unfortunately, due to the date of supply, our records are not completely clearr however, we believe that our Qualification Report B0003 can be used to support the capability of the actuators to withstand irradiation."

2. Reference 2.3 is a report of a qualification test program conducted on a The test program consisted of a 12-hour exposure to warm air saturated with water vapor ( and ). The performance of the actuator was monitored by cycling under load during the exposure (plus cycles before and af ter the exposure), and the exposure). Performance was satisfactory, but the (There were no pre-aging, chemical spray, or nuclear radiation exposures in the test program.)

3. Limitorque Report B0003 [2.4] describes a qualification test program conducted on a SMB-0 MVA having a Reliance motor with Class B insulation, plus two additional motors. The MVA and motors were thermally aged and simulta-neously operated (200 hours at 165'F; operation for 30 seconds in each direction once per hour for 176 hours), and then the MVA was then operated for an additional 1817 cycles to simulate wear aging (the two extra motors were l operated for an additional 15 minutes while the motors were unloaded). The  ;

l MVA then received a nuclear radiation dose of 20 Mrd, and the motors 204 Mrd. 1 Subsequently, the MVA and motors were seismically tested and subjected to a

    &                                           4-19
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                            .  . Ce.nter

DELETED M ATERIAL IS PROPRIETARY WPORM ADON TER-CS257-195 16-day steam exposure test. Functional operation was demonstrated prior to and on five occasions during the latter exposure, the las; immediately preceding the end of the test. Insulation resistance to ground was measured at each of these times. The MVA malfunctioned once (at 25.8 hours, just af ter the ambient temperature had been reduced from 250*F to 200'F) . This mal-function was attributed to a "a momentary electrical short due to localized condensate buildup, a malfunction of the reversing contactor, or a combination of both." The IR readings decreased with time at each of the two temperature plateaus of the steam exposure, but ac current draw was not significantly affected. The manufacturer concluded that "this test generically qualifies Limitorque Valve Actuators type SMB/SB for Class lE Service outside primary containment for conditions as defined in this report." However, as noted in paragraph 1 above, Limitorque believes but cannot verify that this reference is applicable.

4. Reference 2.5 is a letter from Limitorque that provides a general statement attempting to justify a 40-year qualified life based on the pre-aging exposures that are applied in the test programs. Because the applicability of Reference 2.4 13 Uncertain, 3rd because there was no pre-aging in the test program reported in Reference 2.3, this letter appears to be irrelevant to the present evaluation. The Licensee should evaluate the susceptibility of the materials in the MVA to aging degradation and establish the conservative qualified life (refer to Section 4.1.3 for additional comments).

5. FRC considers that all Guidelines requirements except those pertaining to nuclear radiations and aging have been satisfied. Aging was discussed above. With regard to nuclear radiations, it appears that the dose levels are small enough that the Licensee should have no difficulty in establishing qualification by analysis.

FRC CONCLUSION: This equipment is assigned to NRC Category II.c. Although complete qualification decumentation has not been made available to demonstrate A 4-20 W2)Enklin 4 t> n or nResear.ch rma. a m Center

l DEURED WAHIRIAL is Pm0MmnARY INPCAMADON TER-C5257-195 total compliance with the DOR Guidelines, it is expected that the Licensee will be able to demonstrate qualification for all environmental service conditions (including nuclear radiation exposure) and a significant period of qualified life (less than plant life) . 4.3.3.2 Equipment. Item No. 52 (previously designated Il0) Electrical Cable Located Within the Drywell Kerite, Model Not Stated (Final Licensee References 2.16 and 2.23) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT (3.3.2.6): Licensee Reference 2.23 is a test report covering three sets of samples: one for control cable (7/C No. 12 AWG) and two different constructions of power cable (1/C No. 6 AWG) ; plus eight splice samples. FRC's comments on this reference are as follows:

a. The test specimen must be the same as the equipment being qualified.

The Licensee did not present an analysis comparing the impact of deviation between the test specimen's specific design features, materials (specifically, the formulations used in the insulation and jacket), and production procedure, and those of the cables installed in the plant. Therefore, the validity of the test as evidence for qualification has not been established.

b. The test program consisted of steam and boric acid spray exposures , plus cooldown simultaneous with exposure to gamma radiation. The total dose administered to various samples was either The samples were electrically loaded during the simulated LOCA exposures except for periods when electrical measurements were being made. The peak temperature and pressure in the test exceeded the plant-specific accider.t values, but the profile was not completely enveloped. Also, the chemical solution of the spray was different from that in plant. Because of the overall severity of the -test, these deficiencies are judged to be minor and acceptable. The nuclear radiacion exposure was more than adequate.

c. The cable samples were not thermally pre-aged prior to the simulated LOCA exposure, as is required by the Guidelines when it has not been shown that the materials are not subject to aging degradation.

it is particularly It is also important that (i) acceptance criteria be established for these cables, 4-21 4) A. Franklin Research Center acma#n.r- n

DELETED MATEA!AL is PRoPf41ETARY INPCMMATloN TER-CS257-195 considering their plant-specific application, and (ii) a determination be made that the electrical current loadings in the test are adequate. Also, the qualified life should be established. LICENSEE RESPONSE: [No response provided.) FRC EVALUATION: As a result of review of other test reports referenced by Licensee in the EEQ program for SEP plants, FRC has also reviewed FIRL Reports F-C4158, F-C4020-1, and F-C4040-2 (FIRL test on Kerite cable). The cables covered by these reports are: F-C4518: 7/C No. 12 AWG F-C4020-1: 7/C No. 12 AWG F-C4020-2: 7/C No. 12 AWG _ F-C4020-2: 1/C No. 6 AWG For the cables in FIRL Report F-C4020-1, the insulation resistance was noticeably lower after thermal aging, then decreased by a factor of about' 100 af ter irradiation, and by another factor of about 1000 the first 1.5 hour at 346*F/ll3 psig in the test chamber. The report states in the conclusion that the cables were able to maintain load ( ) for 2 days (1 cable), days (1 cable), and days (2 cables) after start of the specified LOCA. For the cables reported in FIRL Report F-C4020-2, the temperature / I pressure conditions of the steam exposure were rapid heating from psig to 346'F/ll3 psig, which was held for 3 hours, followed by cooldown to 140*F in 2 hours. a second rapid heating to 346'F/ll3 psig (held for 3 hours), and then a gradual stepwise drop in temperature to psig, which was held  ; for days. These tests were conducted in 1975 and envelop the Oyster Creek conditions in Appendix A. For the cables reported in FIRL Report F-C4158, the temperature / pressure conditions were hours at , days at (with buffered boric acid spray), then days at 4g 4-22

    .. A Frankhn Research Center
          %.a w n. r-in      .

DEUTED WATEfeAL W MCPANTARY INPORMADON l TER-C5257-195 and finally ambient. This test also envelops the specified conditions for the Oyster Creek Station. All three tests involved simultaneous nuclear radiation and steam / spray exposures. The Licensee submittal does not state whether cables are exposed or in conduit. The tests reviewed involved exposed cables, and the radiation dose cate and total dose exceeded the Oyster Creek requirements. As discussed in Section 4.1.3, FRC does not agree with the manufacturer's claim of a lifetime ir excess of 40 years. FRC CONCLUSION: This equipment is assigned to NRC Category II.c because PRC is aware of test results that qualify the cable. The Licensee should establish a conservative qualified life (see Section 4.1.3) . 4.3.3.3 Equipment Item No. 54 (previously designated Ill) Electrical Splices Located Within the Drywell Raychem Type WCSF (Final Licensee References 2.9 and 2.16) ORIGINAL TEXT TAKEN FRCM DRAFT INTERIM TECHNICAL EVALUATION REPORT (3.3.2.7): Licensae Reference 2.16 is a test report for Raychem splices. Although the test program reported may be adequate, the Licensee has not established that the splices in the plant used the materials and techniques covered in this test program. (The absence of the plant-specific chemical spray is not regarded as a serious deficiency.) LICENSEE RESPONSE: (The Licensee identified the splices as WCSF type, referenced an additional test report (Wyle No. 44114-2), and noted that the radiation level due to an accident is being re-estimated and is_ expected to be lower than the 57 Mrad used in the evaluation.] 4 4-23 , EdJ4 Frankun

           >=.a n n= nResearch.
                         .u. Center                                                          l l

i l l l OELETED MATERIAL 23 PROPRIETARY INFORM ATION TER-CS257-195 l l FRC EVALUATION: After reviewing documentation on splices referenced previously and information supplied by other Licensees for the EEQ program, FRC has the following c' aments:

a. According to information provided to various licensees by Rayenem Corp., fa,ilure in the cable insulation may also result in failure of the splice.

b. Testing reports showed WCSF-type splices to be satisfactory with the Kerite, Rockbestos, and GE Vulkene cables identified in SCEW sheets )

for Oyster Creek, but tests for the Tensolite cable (Equipment Item No. 51) were not reported.

c. The testing conditions enveloped the pressure, temperature, and radiation levels applicable to Oyster Creek. Chemical sprays in the tests (boric acid solutions buf fered to a pH of 9.5-10.5) differed from the Oyster Creek spray. However, as noted in the DITER above, the difference in spray is not considered a serious deficiency.

d. As discussed in Section 4.1.3, FRC does not' agree with the manufacturer's stated 40-year life for this equipment. The Licensee should obtain information to establish a conservative qualified life for splices on all the cables installed in the Oyster Creek Station.

l FRC CONCLUSION: 1 Except when used on Tensolite cables, these splices are assigned to NRC Category II.c because test reports supported compliance with all Guidelines ) criteria except qualified life. WCSF-type splices on Tensolite cables, if l any, would be assigned to NRC Category IV.b because they are likely to be satisfactory but documentation is lacking. The Licensee should establish a  ! conservative qualified life for each cable / splice system and a surveillance l program to monitor performance and identify any degradation which would l indicate the need for maintenance or replacement (see Section 4.1.3).  !

v i i

   . 1 ---

4-24 R /Ja Franklin

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                                      ,            - - - ,       .             4- - ,w

l DELETED MATERIAL IS PHOPfWETARY INPORMATION TER-C5257-195 4.4 NRC Category III EQUIPMENT THAT IS EXEMPT FROM QUALIFICATION l 7 tie equipment items in this section are exempt from qualification on the basis that (1) the equipment does not provide a safety function (i.e., should not have been included in the equipment list submitted by the Licensee), or (2) the specific safety-related function of the equipment can be accomplished by some other designated equipment that is fully qualified. In addition, any failure of the exempt equipment must not degrade the ability of qualified equipment to perform its required safety-related function. 4.4.1  ::quipment Item No. 39 Electric Motors Located in the Reactor Building General Electric Model 5K818841C45 Core Spray Booster Pumps (NZ-03-A through NZ-03-0) (Final Licensee Reference 2.14) ORIGINAL TEXT TAKEd FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSES RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCZ 1): The purpose of the core spray booster pumps is to provide additional pressure increree to the core spray water discharged by the core spray pumps. This ensures rated core spray flow will be established at a l reactor pressure of 110 psig. The core spray system consists of two independent systems, each of which can accomplish its safety function even considering a single active failure. Pumps A and C are in System I and B and D are in System II. The environmental conditions in the area of the B and D pumps are nonharsh when only temperature and pressure are considered. Therefore, there always will be at least one system available to carry out its safety function. The one-year integrated accident exposure to the pumps in System I is on the order of 1 Mrads, and System II pump exposure is on the order of 0.1 Mrads. By evaluation, it has been determined that there will be no detrimental radiation i effects up to radiation exposures of 200 Mrads. l Based on the above considerations, it is expected that even considering the worst-case HELBs, there will be at least one core spray system booster pump available to deliver rate core spray flow to the reactor if that should be required. (Qualified) Per GE report 4 4-25 MO Franidin Research Center

                 = cm w N n.ama mean.

DELETED M ATERIAw W PAOPRfETAOY WPOMM AN TER-CS257-195 TRC EVALUATION: The Licensee has stated that the core spray booster pumps are each able to supply 100% of the core cooling naeds. Accordingly, even if two of the pumps located in the harsh area of the reactor building were rendered inoperable by the MSLB and a single failure prevented one of the mild area , pumps from operating, a 1004 pump located in another area (described by the Licensee as mild) would still remain to furnish the necessary cooling to the core. A MSLB in the reactor building should be of short duration so that the remaining pump would not have to operate for more than a few hours or days to bring the plant to a safe shutdown. On this basis, the pump motor can be considered exempt from qualification. FRC CONCLUSION: The core spray booster pump motors are assigned to NRC Category III because there is sufficient redundancy with equipment located in a mild area to withstand a single failure and still provide the necessary system function capability. 4'jdJ Franklin Research Center 4-26 a w w w.wa -

l DELETED MATENAL is PROPMETAMY INPORMATION I I TER-C5257-195 4.5 EQUIPMENT FOR WHICH DOCUMENTATION CONTAINS DEVIATIONS FROM THE i GUIDELINES THAT ARE JUDGED UNRESOLVED l This section includes equipment items which are deficient on the basis that all criteria defined in the DOR Guidelines are not satisfied. However, the equipment item is either scheduled to be tedted or is judged to have a high likelihood of operability. 4.5.1 NRC Category IV.a EQUIPMENT THAT HAS QUALIFICATION TESTING SCHEDULED BUT NOT COMPLETED The qualification of the equipment items in this section has been judged deficient or inadequate based upon review of the documentation provided by the Licensee; however, the Licensee has stated that the equipment item is scheduled to be tested by a designated date. The results of the testing will j dictate the specific qualification category he equipment item. l ( For the Oyster Crask Station, no equipment falls within this category. 4.5.2 NRC Category IV.b EQUIPMENT FOR WHICH QUALIFICATION DOCUMENTATION IN ACCORDANCE WITH THE GUIDELINES HAS NOT BEEN ESTABLISHED The qualification of the equipment items in this category is deficient or inconclusive based upon review of the documentation provided by the Licensee. l This equipment is judged to have a high likelihood of operability for the specified environmental service conditions; however, complete and auditable records reflecting comprenensive qualificatiori documentation have not been ( made available for review. l 4.5.2.1 Equipment Item No. 1 Pressure Switches Located in the Reactor Building Dresser Model 1539 VX Automatic Depressurization System (ADS) Pressure Switches (IA83A through IA83E) (Licensee reference not cited) CRIGINAL TEXT TAKEN FRCM DRAF'" INTERIM TECHNICAL EVALUATION REPORT: None l ! l l 4-27

           .dJ Frank!!n Research Center ac   awwr      a  a,.

DELETED MATERIAL IS PROPRIETARY INFORMATioN TER-CS257-195 LICENSEE RESPONSE (EQUIPMENT ITDI AFDED IN REFERENCE 1): The ADS provides for a controlled blowdown of the recctor pressure vessel to rapidly reduce pressure during a small pipe break. This permits core spray actuation prior to uncovering the fuel. The pressure switches will open the electromatic relief valves in the ADS on an overpressure condition in the reactor pressure vessel. Each pressure switch is installed at a different location outside the Drywell and a single HELB in the vicinity will net subject all five switches to a peak terperature and pressure. These switches are necessary only for over-pressurization protection and their failure does not affect the ability of the Control Room operator to manually operate ADS valves in order to achieve a controlled cooldown. Even without the relief valves, reactor vessel overpressure protection is provided by 16 safety valves located within the containment. Therefore. they will be unaffected by any HELBs outside containment. FRC EVALUATION: The Licensee has not provided, and FRC has found no other sources of, valid qualification documentation for this equipment. Therefore, qualification has not been established in accordance with the requirements of the Guide-lines. However, some of this equipment is likely to function adequately because its safety function is expected to be performed early in the accident scenario and not all of the pressure switches are expected to be exposed to a harsh environment at the same time. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. FRC CONCLUSION: This equipment is assigned to NRC Category IV.b because, although valid qualification documentation has not been provided, the Licensee has shown that the equipment is likely to function. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to a program of equipment qualification or replacement by June 1982. 4 4-28 NJ Franklin Research Center ADmmadT4Pwwkwamu l

r DELETED MATameAL is PaomesTARY WPORMATION TER-C5257-195

 '4.5.2.2     Equipment Item No. 4C (previously designated IA-2)

Motorized valve Actuators Located Within the Drywell Limitorque Model SMB-000 Main Steam Line Isolation (V-1-106,107) (Original Licensee Reference 2.4; Final Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT (3.3.2.1) : Reference 2.4 is a test report of a qualification test for an SMB-0 actuator. FRC has the following comments with rega'rd to this reference. i a. The test report is for a Limitorque Model SMB-0 actuator with a Reliance motor having Class B insulation. The Guidelines require that the test specimens be the same as the equipment being qualified. The Licensee did not present an analysis comparing the impact of deviations between the test specimen's specific design features, materials, and production procedures and those of the installed equipment. Therefore, an independent conclusion can not be reached regarding the extent to which the tested equipment is similar to that installed in the plant, and the validity of the test, as evidence of qualification, has not been established.

b. The test program included wear / thermal / humidity / seismic aging, vibration test to simulate severe seismic events, and a steam exposure. The environmental parameters, aging considerations, and I other aspects of the test program were intended to demonstrate qualification of equipment located outside of the primary containment; they are not adequate to demonstrate qualification for these equipment items located within the containment drywell.

The Licensee has stated that these equipment items will be replaced with

fully qualified equipment during the 1981 plant outage. FRC is also aware of other test reports, referenced by other Licensees, that demonstrate satis-factory performance for a period of at least a few hours under inside-containment service conditions. FRC recommends that the Licensee centact the manufacturers to obtain access to these reports.

LICENSEE RESPONSE: These valves are inside containment isolation valves for the emergency condenser, shutdown cooling, and cleanup systems. The valve actuators were supplied by Limitorque Corporation and are equipped with Reliance motors having Class B insulations. Our discussions with Limitorque personnel indicated that a test was performed by Franklin Institute Research Laboratories for Westinghouse Company utilizing the same valve 4 4-29 b Franklin Research Ce.nter a cm=a am. r,= .

DELETED MATE 3AL O PROPRIET ARY INFORM ATlON TER-C5257-195 l assembly with motors having Class B insulation. According to the same source, the valve functioned at least 12 hours under conditions expected after a LOCA. The report was identified by the Limitorque personnel as FIRL test F-C2485-01 (dated May 1969) . Several attempts by us to obtain this test report did not succeed since the report is classified as Westinghouse proprietary information. In view of this situation, a decision was made by JCP&L to replace all of these valves with qualified valve assemblies. Accordingly, purchase order No. 28930 was issued on December 20, 1979 and the valve assemblies, along with qualification report, were delivered to Oyster Creek Nuclear Generating Station in June 1980 and are currently kept in our storage room on site. Therefore, the qualified valve assemblies will be installed at the next scheduled shutdown, which will take place in the spring of 1981. FRC EVALUATION: Since the Licensee has not provided valid qualification documentation for this equipment, full qualification has not been established in accordance with the requirements of the Guidelines. Based upon a review of the originally cited reference for similar type valves and the equipment's brief required operating time, this equipment is expected to function adequately.

                                                                                       ~

A review of the Licensee's justification (Chapter 7 of Peference 1) for continued plant operation with this equipment item is given in Appendix D of this report. FRC CONCLUSION: This equipment is assigned to NRC Category IV.b because, although valid qualification documentation has not been provided, test reports for similar type valves have shown that the equipment is likely to function adequately during an accident. The Licensee has committed to replace these valves by the

                                                                          ~

spring of 1981. 4.5.2.3 Equipment Item No. 11 Temperature Detectors Located in the Reactor Building Rochester Instrument System, Model Not Stated Isolation Condenser Area Temperature Detectors f1B-06-E through H) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRArr INTERIM TECHNICAL EVALUATION REPORT: None - 4 4-30 dJfJ %. Re Frankhn,n. search C. r- u. enter

I DELETED MATERIAL is PROPAIETARY INPOeMATION TER-C5257-195 LICENSEE RESPONSE (EQUIPMENT ITDI ADDED IN REFERENCE 1): The isolation condenser area temperature monitors provide indication in the control room of steam leaks in the area. These temperature detectors do not provide any automatic safety functions, but are referred to in the station emergency procedures as one of the parameters that can be used to detect leaks in the isolation condenser system. Since the system is primarily there to detect leaks and not breaks, it is unlikely that the area temperature will reach those levels described in the worst-case break analysis. The one-year worst-case integrated radiation exposure to these instruments is on the order of 1 Mrad for two detectors and on the order of 64 x 104 rads on the other two. While a material list is not available at this time, an evaluation of other temperature switches at the facility shows that radiation exposure up to 1 Mrad is acceptable. The evaluation uses a one-year exposure, and these instruments are used only to verify steam leaks in the area. They would only De utilized by the operators during the first few minutes of any event involving steam leaks in the main steam or isolation condenser system. FRC EVALUATION: FRC has reviewed the operational evaluation above and the information contained on SCEW sheets 44, 45, 46, and 47 and notes the following:

a. The Licensee states that the maximum temperature / pressure to which the equipment is exposed are 280*F/16 psia at the radiation levels noted above.

b. Note B of the SCEW sheets states that this equipment will either be qualified or replaced by July 1, 1982.

c. The equipment is required only for the first 10 minutes of a HELB.

A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. FRC CONCLUSION: This equipment is assigned tn NRC Catiagory IV.b because there is no evidence of qualification, but there is a high likelihwd of operability based on the analysis provided by the Licensee. Although th( Licensee's evaluation of this equipment item has not been completed, the Lice nsee has committed to a program of equipment qualification or replacement by June 1982. gMU Franklin Research Center 4-31 u>=a ame w

DELETED CATERtAt tS PmQPmtETAMY INFQ4MATION TER-CS257-195 4.5.2.4 Equipment Item Nos. 19, 20, 21A, and 22A Solenoid Valves Iccated in the Reactor Building 19: ASCO Model 8344-B27 (V-27-1, -2) 20: ASCO Model 8344-A27 (V-27-3, -4) 21A: ASCO Model 83148 (V-23-13) 22A: ASCO Model WP8300B61RU (V-23-17, -18) Purge Valves and Nitrogen Valves (Final Licensee References 2.7, 2.11 [ Items 20, 21A, 22A], and 2.13 (Item 191) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPME!C ITEM ADDED IN REFERENCE 1): These are normally closed containment isolation valves that will not change position given a f ailure of the solenoid valve. They are in a non-harsh temperature / pressure environment. Our evaluation of the component materials revealed that this component contains thermal aging and radiation-sensitive materials (Buna-N and/or fish paper) . Therefore, the sensitive component materials will be replaced by June 1982. FRC EVALUATION: References 2.11 and 2.13 provide information on the sensitivity of materials to nuclear radiations. Reference 2.7 is not adequately identified and a copy was not provided for review. As noted in Appendix D, this equipment should be qualified for a HELB environment. Also, FRC is not aware of valid qualification documentation for this solenoid valve from other sources. Therefore, qualification has not been established in accordance with the , requirements of the Guidelines. It is expected that this equip: rent will function adequately because the environment is not extremely " harsh." A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. The Licensee should proceed with the preventive maintenance activities on an expedited schedule. The manufacturer should be consulted to obtain recommended replacement schedules for the coils and other non-metallic components used in these valves. h_ 4-32 d.2J Franklin Research Center aw.m.w -

DELETED MATERIAL is POPRIETARY INFORMATION  ! l TER-C5257-195 j FRC CONCLUSION: This equipment is assigned to NRC Category IV.b. Although valid qualification documentation has not been provided, the equipment is likely to function adequately because the environmental conditions are not harsh (except for radiation) for the accident it is intended to mitigate. ?e Licensee has stated that thermal- and radiation-sensitive materials will be replaced by June 1982. 4.5.2.5 Equipment Item No. 22B Solenoid Valves Located in the Reactor Building ASCO Model WP8300561RU ventilation valves (V-23-21, -22; V-28-17, -18, -47) (Licensee References 2.7 and 2.11) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITDt ADDED IN REFERENCE 1): These are containment isolation valves that are normally closed and would not be required for outside-containment HELBs. Our evaluation of the component materials revealed that this component contains thermal aging and radiation-sensitive materials (Buna-N and/or fish paper). Therefore, the sensitive component materials will be replaced by June 1982. FRC EVALUATION: Reference 2.11 provides information on the sensitivity of materials to nuclear radiations. Reference 2.7 is not adequately identified and a copy was not provided for review. FRC is not aware of valid qualification documen-tation for this solenoid valves therefore, qualification has not been established ir. accordance with the requirements of the Guidelines. It is expected that this equipment will function adequately because the only harsh environmental parameter is radiation. However, FRC notes that ASCO has provided recommended replacement schedules for the coils and elastomer parts used in these valves. The Licensee should ensure that the preventive maintenance program includes these recommended replacement schedules. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. 4 4-33 dJJ aFranklin ca a w n.Research r ma w. .,. C. enter

Df uRED M ATER:AL is PRoPRETARY INFOAM AT ON l TER-CS257-195 FRC CONCLUSION: This equipment is assigned to NRC Category IV.b. Although valid qualification documentation has not been provided, the equipment is likely to function adequately because the environmental conditions are not harsh (except for radiation) for the accident it is intended to mitigate. The Licensee has stated that thermal- and radiation-sensitive materials will be replaced by June 1982. 4.5.2.6 Equipment Item No. 26 Solenoid Valves Located in the Reactor Building Atkomatic Model 15-702-B, Type 50R Particulate Monitor System, Oxygen Analyzer System, and Torus Sample System Valves (V-3 8-16, V-38-17, V-38-9, and V-38-10 ) (Final Licensee References 2.6 and 2.7) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These valves are in a non-harsh temperature / pressure environment and are not required to function for HELBs outside containment. Our evaluation of the compontnt materials revealed that this component contains thermal aging and radiation-sensitive materials (Buna-N and/or fish paper) . Therefore, the sensitive component materials will be replaced by June 1982. FRC EVALUATION: Tne references cited by the Licensee are not adequately identified and copies were not provided for review. Also, FRC is not aware of valid qualification documentation for this solenoid valve from other sources. Therefore, qualification has not been established in accordance with the requirements of the Guidelines. However, this equipment is likely to function adequately because the only harsh environr+ntal parameter is radiation. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. 4 4-34 bA"J a o Franklin a w th. Research r- C. enter

DELETED MATElmAL 13 PeopmsETARY INP0mMATION TER-C5257-195 FRC CONCLUSION: This equipment is assigned to NRC Category IV.b. Although valid qualification documentation has not been provided, the equipment is likely to function adequately because the environmental conditions are not harsh (except for radiation) for the accident it is intended to mitigate. The Licensee has stated that thermal- and radiation-sensitive materials will be replaced by June 1982. 4.5.2.7 Equipment Item No. 27 Solenoid Valves Located in the Reactor Building ASCO Model LB82627 Particulate Monitor System, Oxygen Analyzer System, and Torus Sample System Valves (V-38-22 and V-38-23) (Final Licensee Reference 2.7) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITD1 ADDED IN REFERENCE 1): These valves are in a non-harsh temperature / pressure environment and are not required to function for HELBs outside containment. Our evaluation of the component materials revealed that this component contains thermal aging and radiation-sensitive materials (Buna-N and/or fish paper). Therefore, the sensitive component materials will be replaced by June 1982. FRC EVALUATION: The reference cited by the Licensee is not adequately identified and copies were not provided for review. Also, FRC is not aware of valid qualifi-cation documentation for this solenoid' valve fror6 other sources. Therefore, qualification has not been established in accordance with the requirements of the Guidelines. This equipment is likely to function adequately because the only harsh environmental parameter is radiation. However, FRC notes that ASCO has provided recommended replacement schedules for coils and elastomer parts used in these valves. The Licensee should ensure that the preventive maintenance program includes these recommended replacement schedules. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. 4 4-35 )

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DELETED MATERIAL 1s PROPRIETARY INFORMATiON TER-C5257-195 FRC CONCLUSION: This equipment is assigned to NRC Category IV.b. Although valid qualification documentation has not been provided, the Licensee has shown that the equipment is likely to function adequately because the environmental conditions are not harsh except for radiation for the accident it is intended to mitigate. The Licensee has stated that the thermal- and radiation-sensitive materials will be replaced by June 1982. 4.5.2.8 Equipment Item No. 28 Temperature Switches Located in the Steam Tunnel Fenwal Model 17002-40 Reactor Isolation Temperature Switches for Main Steam Line Leak Detection (IB-10 A through P) (Final Licensee Reference 2.11) ORIGINAL TEXT TAKEN FROM DRAM INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These temperature switches are located in the main steam line tunnel outside the drywell to detect a MSLB in the tunnel. However, the detections of a MSLB are provided by other redundant and diverse signals that a:e not affected by the break. Those are reactor water low-level signals, main steam line low-pressure signals, and main steam line high-flow signals, o Que.lification documentation is not available at this time, o This equipmer.' "'ll be either replaced or qualified by June 1, 1982. FRC EVALUATION: The Licensee has neither submitted nor referenced qualification documentation for this item. The Licensee has stated: o The switches are redundant to other safety-related equipment which is r)ot simultaneously exposed to the MSLB harsh enviz;onment. o This equipment is required to operate during a HELB outside containment. o This equipment will be qualificd or replaced by June 1, 1982. 4-36 410 Franidin Research Center w ar m vm e ae

DELETED WATEMAL IS Pm0PmETARY MPOAMATION TER-C5257-195 FRC notes that the components are differential expansion thermoswitches, non-indicating and hermetically sealed, with adjustable setpoint and NEMA and housing which provides a high temperature trip signal to the reactor protec-tion system. FRC has reviewed documentation relevant to this equipment item for the environmental qualification review program and has reached the following conclusions: o A Fenwal model switch was tested. o was performed at dry heat load. The setpoint retained a . o A submergence test was conducted at psig. o A radiation test impcsed a dose of Mrd. o A high temperature test (heated aluminum block) subjected the switch to for . FRC concludes that the heated aluminum block (dry heat) and immersion tests were not equivalent to HELB high temperature all-steam testing. However, the radiation test imposed a greater dose than the required 6.1 x 4 10 rd. It should also be noted that the test specimen model was No. 17023-6, whereas the actual installed equipment model is 17002-43. FRC concludes that this component lacks documentation of operability under HELB environmental service conditions. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with - this equipment item is given in Appendix D of this report. FRC CONCLUSION: This equipment item is assigned to NRC Category IV.b. Although the qualification documentation is deficient with respect to HELB (high temperature all-steam) testing and the specific relationship of the installed switches to the test specimen, the equipment is highly likely to operate. Its design is simple, the adverse environment is within the temperature range in which the unit has performed satisfactorily, and the immersion test provides assurance that steam in-leakage will not be a problem. However, aging and qualified life have not been addressed. Altnough the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to equipment qualification or replacement by June 1982. 4-37 g'?J L Franklin Research Ce.nter a om a w w.mm ia .

DELETED !dATERIAL O PROPRIETARY lNPORM ATION TER-C5257-195 l I 4.5.2.9 Equipment Item Nos. 31B (previously designated I-A1) and 32B (previously designated I-B1) Solenoid Valves Located Within the Drywell 31B: ASCO Model 206-832-3RU 32B: ASCO Model 206-301-3RU Main Steam Isolation valves I (Original Licensee References 2.2, 2.3, and 2.24; Final Licensee References 2.16 and 2.24) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT (3.2.1): Reference 2.24 is a proprietary test report describing a qualification program conducted for a number of ASCO solenoid valves. DITER References 2.2 and 2.3 are letters from ASCO documenting that the tested and installed equipment models have the same coils, coil enclosures, and valve seats. FRC comments as follows, based on review of these references:

a. Of the valve models tested, those with model numbers that correspond to those of the installed equipment ares o Items I-Al and I-Bl Sample No. 4, Model No.

having a o Item I-Cl: Sample No. 5, Model No. having a , The three references establish conformance between the tested and s installed equipment. i b. The en!ironmental and operational service cond.ition parameters used in the qualification test program exceeded those dictated by plant-specific requirements in all cases except (i) the of the steam temperature / pressure profile and (ii) the use of a boric acid / sodium hydroxide spray solution in lieu of a sodium dichromate

        ,                         solution. These deficiencies are not significant. The Licensee submittal did not explicitly consider the nuclear radiation dose resulting form beta radiations (including the bremsstrahlung radiation it creates while being attenuated). Because the nonmetallic components of the solenoid valves are encased within metallic enclosures, the dose contritution from beta radiation can be expected to be quite small. The test program included a sufficiently large gamma radiation dose (                     ) that the beta dose contribution can be considered to have been accommodated.

c. The pre-aging simulated in the test program was intended to represent an installed life (and hence a qualified life) of ambient temperature. Reference 2.24 states that the coil and seats

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DELETto WATERIAL is PROMIIETARY NWORMATION TER-C5257-195 should be replaced at intervals. Provided that the Licensee has established (i) a replacement schedule consistent with this requirement and (ii) a program to review any in-se vice failures to determine whether they are caused by aging degradation, the equipment is considered to be qualified with a qualified life of 4 years. 3 LICENSEE RESPONSE: [No response provided.] FRC EVALUATION: FRC has reviewed the reference (s) cited by the Licensee and has the following comments:

1. During the qualification test program described in Reference 2.24, The results of the test must therefore be regarded as inconclusive until the uncertainties associated with the method of making the wiring interface with the solenoid, both in the plant and in the test, are resolved. The Guidelines state (Section 5.2.5):

                                 "If a component fails at any time during the test, even in a so called ' fail safe' mode, the test should be considered inconclusive with regard to demonstrating the ability of the component to function for the entire period prior to the failure."

They further state (Section 5.2.6):

                                 "The equipment mounting and electrical or mechanical seals used during the type test should be representative of the actual installation for the test to be considered conclusive."

2. The pre-aging simulated in the test program was intended to represent an installed life (and hence a qualified life) of ambient temperature. The ambient temperatures at the installed locations within the plant are lower, and hence the qualified life is longer. An explicit, conservative determination of qualified life and a replacement schedule (if needed) must be established.

4r.d Franklin Research Center 4-39

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DELETED WATERIAL LS PROPRAETARY INFORM ATION TER-C5257-195 FRC CONCLUSION: This equipment is assigned to NRC Category IV.b. Although the results of the qualification test program are inconclusive, the required function occurs early in the accident scenario, and is therefore highly likely to be performed properly. The Licensee should determine how the electrical connections arc , sealed, establish that moisture infiltration will not cause failure, and establish a conservative qualified life. A surveillance program should be~ implemented to monitor performance and identify any degradation which would indicate the need for maintenance or replacement. 4.5.2.10 Equipment Item Nos. 34C (previously designated I-2B) and 44 (previously designated I-2D) Motorized Valve Actuators Located Within the Drywell 34C: Limitorque Model SMB-0 with Reliance Motor (Class B Insulation) Shutdown Cooling Valves (V-17-19 and V-16-1) 44: Limitorque Model SMB-2 with Reliance Motor (Class B Insulation) Isolation Condenser valves (V-14-36, -37) (Licensee Reference 2.4)

• ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT (3. 3. 2.1) :

Reference 2.4 is a test report of a qualification test for an SMB-0 actuator. FRC has the following comments:

a. The test report is for a Limitorque Model SMB-0 actuator with a Reliance motor having Class B insulation. The Guidelines require that the test specimens be the same as the equipment being qualified.

The Licensee did not present an analysis comparing the impact of deviations between the test specimen's specific design features, materials, and production procedures and those of the installed equipment. Therefore, an independent conclusion cannot be reached regarding the extent to which the tested equipment is similar to that installed in the plant, and the validity of the test, as evidence of qualification, has not been established.

b. The test program included wear / thermal / humidity / seismic aging, vibration tests to simulate severe seismic events, and a steam exposure. The environmental parameters, aging considerations, and other aspects of the test program were intended to demonstrate qualification of equipment located outside of the primary containment; they are not adequate to demonstrate ?nalification for these equipment items located within the containmsat drywell.

The Licensee has stated that these equipment items will be replaced with fully qualified equipment during the 1981 plant outage. FRC is also aware of 4-40 hT2 Franklin Research Center

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OfLITED MATERAL IS PACPMETARY lNPOfwATION TER-CS257-195 other test reports, referenced by other Licensees, that demonstrate satisfac-tory performance for a period of at least a few hours under inside containment service conditions. FRC recommends that the Licensee contact the manufacturers to cbtain access to these reports. LICENSEE RESPONSE: These valves are inside centainment isolation valves for emergency condenser, shutdown cooling, and cleanup systems. The valve actuators were supplied by Limitorque Corporation and are equipped with Reliance motors having Class B insulations. Our discussions with Limitorque personnel indicated that a test was performed by Franklin Institute Research Laboratories for Westinghouse Company utilizing the same valves assembly with motors having Class B insulation. According to the same source, the valve functioned at least 12 hours under conditions exsected after a LOCA. The report was identified by the Limitorque personnel as FIRL test F-C2485-01 (dated May 1969) . Several attempts by us to obtain this test report did not succeed since the report is classified as Westinghouse proprietary information. In view of this situation, a decision was made by JCPEL to replace all of these valves with qualified valve assemblies. Accordingly, purchase order No. 28930 was issued on December 20, 1979 and the valve assemblies, along with qualification report, were delivered to Oyster Creek Nuclear Generating Station in June 1980 and are currently kept in our storage room on site. Therefore, the , qualified valve assemblies will be installed at the next scheduled shutdcwn, which will take place in the spring of 1981. FRC EVALCATION: The Licensee has not provided valid qualification documentation for this equipment. Therefore, full qualification has not been established in accord-ance with the requirements of the Guidelines. Based upon a review of the originally cited reference and the brief time this equipment must operate, the equipment is expected to function adequately. A review of the Licensee's justification (Charter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. , FRC CONCLUSION: This equipment is assigned to NRC Category IV,b because, although valid qualification docamentation has not been provided, the Licensee has shewn that the equipment is likely to function adequately. The Licensee has committed to replace this equipment with fully qualified equipment in the spring of 1981. 4 4-41 di Franklin Research Center w at v enenn ma.

DELETED MATERIAL 0 MIOMilETAM INPORMATION TER-C5257-195 4.5.2.11 Equipment Item Nos. 35 and 36 Solenoid Valves Located in the Reactor Building 35: ASCO Model IN831424 36: ASCO Model WP8300B61V Orywell Isolation valves (Final Licensee References 2.7 and 2.11) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): 35: Reactor Water Sample valves (V-24-30) This valve is the outsira containment isolation valve for the reactor sample line. Although this valve may see a fairly high temperature environment in the event of a cleanup line break, it is normally closed. In addition the redundant valve inside containment is also normally closed. In the event it was open, both the inside and outside containment valve would be closed on diverse containment isolation signals. .Our evaluation of the component materials revealed that this component contains thermal aging and radiation-sensitive materials (Buna-N and/or fish paper) . Therefore, the sensitive component materials will be replaced by June 1982. Based on the above consideration, it is unlikely that containment isolation would not be achieved via the sample line for a cleanup line break. 36: Drywell Sump Discharge Valves (V-22-1, V-22-2, V-22-28, and V-22-29) These valves are the containment isolation valves for the Drywell equipment drain tank and sump. These valves do not see a harsh temperature / pressure environment for any postulated HELBs. Also, it should be noted that these valves are not needed for isolation purposes for breaks outside containment. Our evaluation of the component materials revealed that this component contains thermal aging and radiation-sensitive materials (Buna-N and/or fish paper). Therefore, the sensitive component mata r rials will be replaced by June 1982. Based on the above information, the isolation function of these valves is maintained for all postulated HELBs outside containment. FRC EVALUATION: The references cited by the Licensee are not adequately identified and copies were not provided for review. Also, FRC is not aware of valid i

   &                                          4-42 Lad Frankhn Research Center A Chuemen of The Feensen Insense
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DELETED MATERIAL 13 PROPRIETARY INFORMATION TER-C5257-195 qualification documentation for this solenoid valve from other sources. Therefore, qualification has not been established in accordance with the requirements of the Guidelines. FRC has reviewed the Licensee's justificttion (Chapter 7 of Reference 1) for continued plant operation with this equipment (see Appendix D of this re por t) and is satisfied with the technical discussion except for a remaining concern about the need to open the valves af ter the accident has occurred. FRC also notes that ASCO has provided recommended replacement schedules for coils and elastomeric components used in these valves. The Licensee should ensure that the preventive maintenance program includes these recommended replacement schedules. FRC CONCLUSION: This equipment is assigned to NRC Category IV.b because, although valid qualification documentation has not been provided, the Licensee has shown that the equipment is likely to function adequately during the early stages of an accident. Maintenance and replacement of parts or the entire unit in accordance with the manufacturer's schedule should be followed. The Licensee has stated that thermal- and radiation-sensitive materials will be replaced by June 1982. 4.5.2.12 Equipment Item No. 37 Motorized Valve Actuators Located in the Reactor Building Limitorque Model SMB-1 with Reliance Motor (Class B Insulation) Core Spray Valves (V-20-15, -21, -4 0, -41) (Final Licensee References 2.2, 2.3, 2.4, and 2.5) l l ORIGINAL TEXT TAKEN FRCM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1) : l The core spray system is set up such that V-20-15 and V-20-40 are located l in parallel on one side of the reactor, and V-20-21 and V-20-41 are located on the other side in parallel approximately 180 degrees apart and on two different floors. Also, only one of the valves needs to operate for the system ts perform its function. Only one pair of valves will be subjected to the harsh accident conditions, thereby leaving the other pair in a relatively mild environment environment and able to function. (The Licensee also notes that qualification of these units is not established by References 2.3 and 2.4.] 4 4-43 AW Franklin Research Center

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DELETED M ATE RIAL IS PROPP;2TARY lNFORM A. DON TER-C5257-195 FRC EVALUATION: The Licensee's references have been discussed in connection with Equipment Item Nos. 3A, 3B, 4A, 4B, 34A, and 34B. As the Licensee notes, the cited test reports do not establish qualification in accordance with the requirements of the Gu.tdelines. Based upon a review of the Licensee Response and all known test reports that may apply to this equipment, it appears likely that this equipment will function adequately. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. FRC CONCLUSION: This equipment is assigned to NRC Category IV.b because, although valid qualification documentation has not been provided, the extensive amount of testing conducted on similar equipment provides reasonable assurance that the equipment is likely to function adequately. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to a program of equipment qualification or replacement by June 1982. 4.5.2.13 Equipment Item No. 40 Motorized Valve Actuators Located in the Reactor Building Limitorque Model SMB-2 with Reliance and Peerless Motors (Class B Insulation) Emergency Condenser valves (V-14-30 through -35) (Final Licensee References 2.2, 2.3, 2.4, and 2.5) 3 ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None .

                                                                                     )

LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1) : The isolation condenser system is set up such that v-14-30 and V-14-31 are connected in series; V-14-32 and V-14-33 are also connected in series. According to calculations performed during NEMA standards (Pub. No. mg/1) on motor isolation, V-14-30 and V-14-32, being ac class B motors, can withstand a maximum ambient temperature of 221*F; whereas the other s (V-14-31, V-14-3 3, and V-14-3 4, V-14-3 5) , being de Class B motors, can withstand a maximum ambient temperature of 275'F. According to our l analysis, the maximum accident temperature is 280'F. The above mentioned l motors (V-14's) are only needed for a maximum of 60 seconds. Therefore, the motor will have performed its function 60 seconds into the accident 1 A 4-44

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DELETE 0 MATERIAL is PMoPRIETARY !NPORMATION TER-C5257-195 and will no longer be needed. It is unlikely that the motor windings will heat up (due to the accident temperature) to this critical temperature of 275' within the time that the motors are needed. (The Licensee also notes that qualification of these units is not established by Reference 2.3 and 2.4.] FRC EVALUATION: The Licensee's references have been discussed in connection with Equipment Items JA, 3B, 4A, 4B, 34A, and 34B. As the Licensee notes, the cited test reports do not establish qualification in accordance with the requirements of the Guidelines. Based upon a review of the Licensee Response and all known test reports that may apply to this equipment, it appears likely that this equipment will function adequately. ! A review of the Licensee's justification (Chapter 7 of Reference 1) for l l continued plant operation with this equipment item is given in Appendix D of l this report. . FRC CCNCLUSICN: This equipment is assigned to NRC Category IV.b because, although valid qualification documentation has not been provided, the extensive amount of testing conducted on similar equipment provides reasonable assurance that the equipment is 1!kely to function adequately during the brief required operating time. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to a program of equipment qualification or replacement by June 1982. 4.5.2.14 Equipment Item No. 42' Solenoid valve Located in the Reactor Building ASCO Model Wr8300B61RV Head Cooling System Isolation Valve (V-31-2) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATICM REPORT: l None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFEP2NCE 1) : 1 The purpose of this valve is to provide reactor coolant boundary isolation. This valve is normally closed, and fails closed on a loss of l i 4 4-45

     ...J Franklin.N a c>            %wa ==mmeResearch Center

DELETED WATERIAL IS PROPRIETA3 INFORM ATION TER-CS257-195 air or power. Also, the piping outside of the containment is designed for a higher pressure than the Nuclear Steam Supply System. This valve is used if the head cooling system needed to ensure the Technical Specification limit on the vessel flange to head temperature of 200*F was not violated during a plant cooldown. Based on the above considerations, it is expected that the valve will continue to carry out its safety function of isolating a reactor coolant system boundary even in the event of a HELB inside or outside contain-ment. Our evaluation of the component materials revealed that this component contains thermal aging and radiation-sensitive materials (Buna-N and/or fish paper). Therefore, the sensitive component materials will be replaced by June 1982. FRC EVALUATION: The Licensee has not provided, and FRC has found no other sources of, valid qualification documentation for this solenoid valve. Therefore, qualification has not been established in accordance with the requirements of the Guidelines. FRC's review of the Licensee's justification for (Chapter 7 of Reference

1) for continued plant operation with this equipment item is given in Appendix D of this report. It is expected that this equipment will function adequately because the only harsh environmental condition is radiation. However, FRC notes that ASCO has provided recommended replacement schedules for coils and elastomer components used in these valves. The Licensee should ensure that the preventive maintenance program includes the recommended replacement schedules.

FRC CONCLUSION: This equipment is assigned to NRC Category IV.b because, although valid qualification documentation has not been provided, the equipment is likely to function adequately. . The Licensee has stated that thermal- and radiation-sensitive materials will be replaced by June 1982. 4_ 4-46 nud FranWin Ruearch Center l 4 on .t n r m i I

DELETED MATERIAL is PROPRIETARY IMPORMATION TER-C5257-195 4.5.2.15 Equipment Item No. 43 (previously designated I-1C) Solenoid Valve Located Within the Drywell ASCO Model NP-8320A187E Sample Valve (Original Licensee References 2.2, 2.3, and 2.24; Final References 2.16 and 2.24) ORICINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT (3. 2.1) : Reference 2.24 is a proprietary test report describing a qualification program conducted for a number of ASCO solenoid valves. DITER References 2.2 and 2.3 are letters from ASCO documenting that the tested and installed equipment models have the same coils, coil enclosures, and valve seats. FRC comments as follows, based on review of these references:

a. Of the valve models tested, those with model numbers that correspond to those of the installed equipment are:

o for Items IA-1 and IB-1: Sample No. 4, Model No. o Item IC-1: Sample No. 5, Model No. i The three references establish conformance between the tested and installed equipment.

b. The environmental and operational service condition parameters used in the qualification test program exceeded those dictated by plant-specific requirements in all cases except (i) the of the steam temparature/ pressure profile and (ii) the use of a boric acid / sodium hydroxide spray solution in lieu of a sodium dichromate solution. These deficiencies are not regarded as significant. The Licensee submittal did not explicitly consider the nuclear radiation dose resulting from beta radiations (including the bremsstrahlung radiation it creates while being attenuated) . Because the nonmetallic components of the solenoid valves are encased within metallic enclosures, the dose contribution from beta radiation can be expected to be quite small. The test program included a sufficiently large gamma radiation dose ( ) that the beta dose contribution can be considered to have been accommodated.

c. The pre-aging simulated in the test program was intended to represent an installed life (and hence a qualified life) of amoient temperature. Reference 2.24 states that the coil and seats 4 4-47 E' Franklin Research Center w at wmmen m ma

DELETE 0 M ATERIAL IS PROPRIETARY INFORM ATION TER-C5257-195 should be replaced at intervals. Provided that the Licensee has established (1) a replacement schedule consistent with this requirement and (ii) a program to review any in-service failures to determine whether they are caused by aging degradation, the equipment is considered to be qualified with a qualified life of 4 years. LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): [No response provided.] FRC EVALUATION: FRC has reviewed the references cited by the Licensee and has the following comments:

1. During the qualification test program described in the reference, The results of the test must therefore be regarded as inconclusive until the uncertainties associated with the method of making the wiring interface with the solenoid, both in the plant and in the test, are resolved. The Guidelines state (Section 5.2.5) :

                       "The equipment mounting and electrical or mechanical seals used during the type test should be representative of the actual installation for the test to be considered conclusive."

2. The pre-aging simulated in the test program was intend (d to represent an installed life (and hence a qualified life) of ambient temperature. The ambient temperatures at the installed locations within the plant are lower, and hence the qualified life is longer (see Section 4.1.3) .

FRC CONCLUSION: This equipment is assigned to NRC Category IV.b because, although valid qualification documentation has not been provided, the equipment is expected to function to close and remain closed during the early portion of an accident. To fully quality this equipment, the Licensee should demonstrate that the electrical connection is adequately sealed, and should also I ( 4M: Franklin Research Center 4-48

              % .s w.ama neu.

DELETED MATERIAL iS PRCPRIETARY INFORMATION TER-C5257-195 demonstrate long-term performance. The qualified life should be determined on a more conservative basis. 4.5.2.16 Equipment Item Nos. 46 and 47 (previously designated I-4A, B, C, D) Electrical Connectors Located Within the Drywell Containment ITT-Cannon Models 46: CA-3106E-36A-46P-F80 CA-3100K-36A-46S-F80 47: CA-06RX-36A-10P-A95 CA-310 0RX-3 6A-10S-A9 5 (Final Licensee References 2.7 and 2.16) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION RZPORT (3.3.2.3): Licensee Reference 2.7 is a report on a qualification test performed on connectors that are virtually identical to those installed in the plant. FRC's comments on this test report art as follows:

a. A thorough analysis was made (and reported in Reference 2.7) of the operation service conditions associated with the installed equipment and of thermal aging effects. The Licensee's contractor concluded that the connectors are not subject to aging degradation, but the basis for this claim is not rigorous (i.e. ,' it relied on the claim of 40,000 hours service at 105'F and the "10*C Rule," rather than on specific aging data). Possible long-term effects of humidity and nuclear radiation were not considered. A more conservative approach to qualified life should be taken.

b. The temperature / pressure profile in the test exceeded the plant-specific profile (except for rise time), and the correct chemical spray was used.

c. The analysis in the report concluded that only 4.8 Mrd of nuclear radiation would be required to establish qualification. This is regarded as inadequate for equipment that must provide long-term service within the drywell. Also, it is ,not stated in the report that even this rather modest exposure was administered.

1 LICENSEE RESPONSE:

                                                                                   )

1 (No response provided.) FRC EVALUATION: The Licensee provided no respcese to the DITER; therefore, the original comments apply. The Licensee states in the SCEN sheets that this equipment ge 4-49

   .5 Franklin Research Center
        < cw.an ao m v,.an m

DELETED M ATECIAL O PROPRIETAQY lNFORMATION TER-C5257-195 will be qualified or replaced by July 1, 1982. The major qualification discrepancy is between the 57-Mrd radiation dose required by the SCEW sheet and the test dose of less than 5 Mrd. FRC CONCLUSION: This equipment is assigned to NRC Category IV.b because the equipment is highly likely to perform adequately, but the qualification is not complete for the radiation dose specified by the Licensee. It is noted that the Licensee has committed to qualify or replace the item by June 1982. 4.5.2.17 Equipment Item No. 51 (previously designated I8) Electrical Cable Located Within the Drywell Tensolite, Model Not Stated (previously stated "Tefzel Insulation /Unjacketed") (Final Licensee References 2.16 and 2.22) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT (3.3.4.1): Documentation reflecting qualification for the_following equipment has not been maje available for review. LICENSEE RESPONSE: [No response provided.] FRC EVALUATION: Licensee Reference 2.22 and other test reports available to FRC on Tefzel insulated cables (FIRL Report F-C3859-1) have been reviewed. Based on these reviews, FRC has the following comments:

a. With regard to similarity of test specimen to installed cable, the ,

Licensee submittal has not identified the type, size, or number of ' conductors, or the jacket material of the Tensolite cable,

b. Reference 2.22 notes that 7C AWG No. 12 with a combined and Nomex insulation and Tefzel jacket were tested to recommendations of IEEE Stds 323-74 and 383-74 and had satisfactory insulation L

  &                                            4-50 dbnklin Research Center 4 cm.aa er ne Fr aana ===.

DELETED MATEM8AL is PROPRIETARY INFORh4ATION TER-C5257-195 resistance after test. No data were presented on insulation resistance during LOCA exposure nor does the reference state that the tested cables are the same as those installed at Oyster Creek (see Appendix G).

c. Single conductor and multiconductor cables using Tefzel insulation for another manufacturer did not perform satisfactorily as reported

, in FIRL Report and did not survive the test. It appears that this was due to aging and a 200-Mrd radiation exposure.

d. Pressure, temperature, humidity, and chemical spray of the tests enveloped the LOCA conditions for oyster Creek.

FRC CONCLUSION: This equipment is assigned to NRC Category IV.b because extensive testing shows it is highly likely to operate satisfactorily at Oyster Creek where the maximum exposure is 57 Mrd. It is recommended that this equipment be replaced with cable that fully satisfies the DOR Guidelines. l 1 l 4-51 4 NIJ aFranklin on a a n. Research n.a . C. enter l

DELETED MATER 1AL 0 PROPRIETARY INFORMAT10N TER-C5257-195 4.6 NRC Category V EQUIPMENT THAT IS UNQUALIFIED The DCR Guidelines require that complete and auditable records reflecting a comprehensive qualification methodology and program be referenced and made available for review of all Class 1E equipment. The qualification of equipment items in this section has been judged to be deficient or inadequate, based upon review of the documentation provided by the Licensee. The extent to which the equipment items fail to satisfy the criteria of the DOR Guidelines can be categorized as follows: (1) documen-tation reflecting qualification as specified in the DOR Guidelines has not been made available for review, (2) the documentation is inadequate, or (3) the documentation indicates that the equipment item has not successfully passed required tests. 4.6.1 Equipment Item No. 6 Pressure Transmitters Located in the Reactor Building General Electric Model GE/MAC 551 Reactor Vessel Pressure Transmitter (ID-45A and B) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRArr INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These pressure transmitters are installed to provide only an indication to the Control Room operator. The transmitters do not perform any safety functions. Even if these transmitters failed, the relief valves in ADS or 16 safety valves will relieve the pressure in the versel, and thus the i reactor vessel is well protected from over pressurization for any postulated HELB. l FRC EVALUATION: The Licensee has neither submitted nor referenced qualification documentation for this item. Also, FRC is not aware of qualification A 4-52 M.5nidin Research Center A :hemen af N Fw ansonste

T DELETED MATERIAL 1s PROPRIETARY INFORMAT1oN TER-C5257-195 documentation for this equipment from other sources. Therefore,-qualification has not been established in accordance with the requirements of the Guidelines. The Licensee Response states that the transmitters do not perform a safety function. However, these transmitters do provide the operator an indication of reactor vessel pressure. Since this information is necessary for cold shutdown and to allow the operator to monitor the performance of safety systems (i.e., the automatic depressurization system [ ADS), the transmitters are safety-related. FRC concludes that this equipment lacks documentation demonstrating operability under HELB environmental service conditions. The Licensee has provided justification for interim operation by stating that relief valves in the ADS will relieve vessel pressure and protect against over-pressurization j (see Appendix D of this report) . The Licensee also stated that this equipment will be replaced or qualified by June 1, 1982. FRC CONCLUSION: This equipment item is assigned to NRC Category V because there is no l evidence of qualification. FRC's review of the Licensee's justification I (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to equipment qualification or replacement by June 1982. 4.6.2 Equipment Item No. 7 Pressure Transmitter Located in the Reactor Building i General Electric Model VPF 1438 l Reactor Vessel Pressure Transmitter (lD-46 A and B) l (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAPr INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): , These pressure transmitters are installed to provide only an indication l to the Control Room operator. The transmitters do not perform any safety functions. Even if these transmitters failed, the relief valves in ACS or 16 saf ety valves will relieve the pressure in the vessel, and thus the reactor vessel is well protected from over pressurization for any postulated HELB. 4 4-53 J.i.0 Franklin Research Center a c> aa om. re eue

DELETED MATERIAL IS PROPRIETARY INFORM ATioN TER-C5257-195 FRC EVALUATION: The Licensee has neither submitted nor referenced qualification documen-tation for this item. Also, FRC is not aware of qualification documentation for this equipment from other sources. Therefore, qualification has not been established in accordance with the requirements of the DOR Guidelines. The Licensee response states that the transmitters do not perform a safety function. However, these transmitters provide the operator with an indication of reactor vessel pressure. Since this information is necessary for cold shutdown and to allow the operator to monitor the performance of safety systems (i.e. , the ADS system) , the transmitters are safety-related. FRC concludes that this component lacks documentation demonstrating operability under HELB environmental service conditions. The Licensee has provided justification for interim operation by stating that relief valves in ADS will relieve vessel pressure and protect against over-pressuriration (see Appendix D of this report). The Licensee also stated that this equipment will be replaced or qualified by June 1, 1982. FRC CONCLUSION: This equipment item is assigned to NRC Category V because there is no . evidence of qualification. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to equipment qualification or replacement by June 1982. 4.6.3 Equipment Item Nos. 8A and 8C Transmitters Located in the Reactor Building General Electric Model GE/MAC 553 8A: Emergency Condenser Level (lG-0 6-A-1, -A-2, -B-1, -B-2) 8C: Containment Spray Flow (IP-03A, B) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None

        &                                               4-54 NJ %Franklin         Researer
                    .# w %n a m. nu            '. enter

_ _ ._ _ _ _ , _ _ - . - _ - - _ .l

DELETED MATEMAL is PMOPMETARY INFORMATION TER-C5257-195 LICENSEE RESPONSE (EQUIPMENT ITDI ADDED IN REFERENCE 1): Emergency Condenser Level Transmitter: Each emergency condenser, containing a minimum water volume of 22,730 gallons of condensate on shell side, provides 11,060 gallons above the top of the tube handles. This volume can accommodate the reactor decay heat for up to I hour and 40 minutes without any need for makeup water (both condensate and service) . If one condenser is used, it can accommodate reactor decay heat up to 45 minutes af ter a scram from full

 ,        power before makeup is required. The reactor can also be depressurized by using ADS. Therefore, the operator can manually initiate the ADS actuation with 45 minutes of a scram following an accident. The ADS, which is located inside the Drywell, is not affected by the accident, l          since the worst-case HELB considered is an emergency condenser line break outside the Drywell.

Containment Spray Flow Transmitter: The containment spray flow transmitters are used by the control room operator to verify containment spray system is actually delivering its required flow. The containment spray system would only be used if there I had been an inside containJent LOCA or the torus had to be utilized as a heat sink in order to ach8. ave safe shutdown. In the case of an inside containment LOCA the harsh temperature and pressure environment outside containment would not exist and only radiation effects would have to be considered. For HELB's outside containment only IP-03-B would see a slightly harsh temperature of 140 degrees. There is not documentation of radiation qualification for these components. It should be noted that these instruments provide only indication and do not perform any automatic safety functions. Even considering the loss of this indication the operator has various other backup parameters that will verify adequate system flow. They are containment spray motor amperes, pump j discharge pressure, torus temperature and valve position. Based upon the above justification, it is expected that instruments will function as intended if they were required for core spray system flow verification. FRC EVALUATION: l The Licensee has neither submitted nor referenced qualification documentation for these items. Also, FRC is not aware of qualification documentation for this equipment from other sources. Therefore, qualification ha s' not been established in accordance with the requirements of the Guidelines. 4_ 4-55 l h Franklin Research Center l *on aw N rna a asam

DELETED M ATERIAL IS PROPRIETARY INFORM ATION TER-C5257-195 FRC concludes that this component lacks documentation derronstrating operability under HELB environmental service conditions. The Licensee has provided a justification for interim operation by stating that (1) the emergency condenser minimum volume of water can accommodate reactor decay heat for a short time and the operator can actuate the ADS, and (2) backup instrumentation can indicate spray flow (see Appendix D of this report) . The Licensee states that this equipment will be replaced or qualified by June 1, 1982. FRC CONCLUSION: These equipment items are assigned to NRC Category V because there is no evidence of qualification. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to equipment qualification or replacement by June 1982. 4.6.4 Equipment Item Nos. 8B, 8D, and 8E Transmitters Located in the Reactor Building General Electric Model GE/MAC 553 8B: Reactor Water Level Transmitters (lD-13A, B; 1A-12A, B) BD: Drywell Pressure Transmitter (IP-07) 8E: Containment Spray Differential Pressure Transmitter (IP-05A through D) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAPI INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): 8B: Reactor Water Level Transmitters (lD-13A, B; IA-12A, B) [SCEWS Nos.: 30-33] These level transmitters are installed to provide only an indication to the control room operator and they do not perform any safety functions. As described in our justification for item 25, the reactor will be scramed and isolated regardless of the availability of these transmitters. 4_ 4-56 dL!$ Franklin Research Center 4caa.emr- - . . _-- - . . - - - _ -_. - _ . - _ _ ._.. _ . . - - . .m ,__ _ . _

DELETED MATEMIAL 18 Pm0PPBETARY INFORMATION TER-C5257-195 8D: Drywell Pressure Transmitter (IP-07) [SCEWS No. : 54] The peak temperature and pressure seen by the switch are 230*F and 16 psia, respectively, following an Emergency Condenser line creak outside containment (worst case HELB) . The containment (Drywell) pressure transmitter is provided to transmit containment pressure I indication to Control Room. However, it is not required to mitigate a line break outside containment. 8E: Containment Spray Differential Pressure Transmitter (IP-05A through D) [SCEWS Nos . : 133-136} The purpose of these differential pressure transmitters is to detect tube leaks in the containment spray heat exchangers. These leaks

might provide a potential leakage path to the environment of

( radioactive effluent.- This component does not provide any automatic function and only serves to provide an alarm in the control room. It is not expected that the containment spray heat exchanger tubes would leak, since they were retubed with titanium in the spring of 1980. This material has preved to be highly resistant to corrosion in other similar applications at Oyster Creek. Therefore, based on the above discussion, there is reasonable assurance that the containment spray heat exchanger will not provide an undetected leakage path for radioactive effluent. (With respect to Equipment Items 8D and 8E, the Licensee has grouped this equipment with items for which the following statement made in the introductory paragraphs of Chapter 7, Reference 1, apply.] These components are not required to mitigate a HELB. Even if this equipment were to fail af ter a HELB, the . protection of the reactor is adequately provided by other systems (and the non-asterisked equipment). Therefore, we have evaluated the thermal aging and radiation susceptibility characteristics of the component materials. This evaluation revealed that certain equipment included thermal aging and radiation-sensitive materials (Buna-N and fish paper) . JCP&L will replace this component material with a qualified one by June 1982. FRC EVALUATION: The advs.rse environmental conditions stated by the Licensee for these ccmponents t.re l l l l 4 4-57

           .iu Franklin Research Center a om.aa as N r wn =====

DELETED MATERIAL is PROPRIETARY INFORMATION TER-C5257-195 8B: 230*F/16 psia, 6.1 x 104 rd l 8D: 230*F/16 psia, 3.9 x 10 rd 8E: 77-140*F/15 psia, 7.5 x 105 rd FRC concludes that this equipment is exposed to a harsh environment and therefore must be qualified. FRC notes that the Licensee SCEW sheets state that the equipment will be replaced or qualified by July 1,1982, but the introduction to Chapter 7 {1] states that radiation-sensitive material will be replaced by July 1, 1982. The Licensee has neither submitted nor referenced qualification documen-tation for this equipment. Also, FRC is not aware of qualification documen-tation for this equipment from other sources. Therefore, qualification has not been established in accordance with the requirements of the Guidelines. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D

                                                       ~

of this report. FRC CONCLUSION: These equipment items are assigned to NRC Category V because there is no evidence of qualification. The Licensee has stated that the equipment item will be qualified or replaced with qualified equipment or that radiation- or thermal-sensitive material will be replaced by July 1, 1982. 4.6.5 Equipment Item No. 30 Electric Motors Located in the Reactor Building General Electric Model 5K-818842A103 . Drives Containment Spray Pumps (PM-51-1-1 through -4) (Final Licensee Reference 2.14) ORIGINAL TEXT TAKEN FROM DIAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REF1' TE 1):

     " Qualified" nklin Rese
       ~ w_       arch Center

DELETED MATERIAL la PnQPMIETARY INFORMAT10N TER-C5257-195 FRC EVALUATION: The Licensee has referenced a General Electric test report and has stated that it will be made available for review, but has not yet done so. The containment spray pump motors are located in a harsh area subject to a 165'F temperature and radiatiion dose of 1 Mrd. The Licensee's qualification documentation should verify that the motors' bearings, lubrication, insulation system, and motor lead splices will not be degraded by the harsh environment. The Licensee should obtain and analyze maintenance information to determine if equipment degradation has been abnormal and to help estimate the equipment's qualified life (see Section 4.13) . FRC CONCLUSION: The containment spray pump motors are assigned to NRC Category V because qualification documentation has not been provided. The Licensee should provide the referenced General Electric test report, and establish a

                                                        ~

conservative qualified life. 4.6.6 Equipment Item No. 45 (previously designated I-3A, -B, -C) Electrical Penetration Located Within the Drywell General Electric Models F01, NS02, NS03, and NSO4 (Final Licensee References 2.16, 2.17, 2.18, and 2.19) ORIGINAL TEXT TAKEN FROM DRAPf INTERIM TECHNICAL EVALUATION REPORT (3.3.2.2) : In general, electrical penetrations perform two safety-related functions: (i) provide a leak-tight barrier as part of the overall plant primary containment, minimizing release of radioactive materials, and (ii) carry electric power, control, and instrumentation signals across the containment boundary. With regard to the first function, the design of this equipment item has three implicit failure modes that must be addressed: distortion of the penetration strut.tural members, failure of elastemeric seals on the mounting flange (if present) , and failure of the seals and electrical insulation around individual conductors. With regard to the second function, two failure modes are relevant: breakdown of the electrical insulation, causing a short circuit t;o ground or between conductors (or high leakage currents, in the case of conductors for instrumentation signals), and A 4-59 Ebnidin Research Center a :>ma as nw r==a mesma.

l DELETED MATERIAL G PROPRIETARY INFORM ATioN TER-CS257-195 breakage of the conductor, causing an open circuit. It is important to note that the two functions are related in at least two ways. First, two of the failure modes for the first function are likely to also cause one or both of the possible failure modes associated with the second function (i.e., an insulation or seal failure around a conductor may both impair containment integrity and cause electrical failures). Second, the f act that the conduc-tors carry electrical currents results in higher than ambient temperatures in the seals and insulation and in electromagnetic and thermal-induced forces being imposed on these materials and the conductors. These effects help to induce failure modes, leading to impairment of both basic functions. The environmental service conditions inside containment are more severe than those outside, considering both normal operation and possible accidents. Hence, these constitute the conditions for which qualification must be established. FRC has reviewed the documentation submitted by the Licensee and has found it to be-highly fragmented and deficient in several aspects, as follows:

a. While the Licensee claims that the penetrations supplied and installed in Oyster Creek are *ype tested in DITER Reference 2.6, no supporting documentation has been provided. Also, FRC notes that there is no identification of the type or models tested in this re fer ence. There is likewise no traceability to the unit tested in Reference 2.17. The Guidelines require that the test specimen must be the same as the equipment being qualified. The Licensee did not present an analysis comparing the impact of deviations between the test specimen's specific design features, materials, and production procedures and those of the installed equipment. Therefore, an J independent conclusion cannot be reached regarding the validity of the tests described in the referenced documentation.

b. For materials subject to thermal aging, the Guidelines require that qualified life must be established. The used in the penetrations is not identified in DITER Reference 2.6 and, while humidity aging tests were performed on several epoxies, no thermal aging test was reported on either the used or the penetration as a whole. The Guidelines require that thermal aging (where applicable), radiation exposure, chemical spray, LOCA/HELB testing, and submergence testing (where applicable) be conducted on the same sample (s) .

c. Although the temperature / pressure profile used in the test reported in Reference 2.17 exceeded the plant-specific profile, no spray was used and there were no nuclear radiation tests on  ; it O 4-60 6dEnklin Research Center a w # % rma a v ena.

DELETED MATEMIAL IS PROPMETARY INPQMMATION TER-CS257-195 has not been established that this material is used in the units installed in the plant.

d. The tests reported in Reference 2.17 included some electric current loadingr, but no information has been presented to show that the values used are adequate for the plant, especially considering possible short circuits in high power conductors as the " active single failure."

LICENSEE RESPONSE: [No response provided.] ' FRC EVALUATION: The Licensee has identified the electrical penetrations as Types F01 and/or NS02, NS03, NSO4. Since the Licensee has provided no additional information, the comments :!.n the DITER still apply. The Licensee notes that the penetrations will be either qualified or replaced by July 1, 1982. FRC CONCLUSION: This equipment is assigned to NRC Category V because testing has not demonstrated that the equipment will meet Guidelines requirements. The Licensee has committed to qualify or replace the equipment by July 1982. 4.6.7 Equipment Item No. 48 (previously designated I5) Terminal Blocks Located Within the Drywell General Electric Model EB (Final Licensee References 2.16 and 2.20) ORIGINAL TEXT TAKEN FRCM DRAFT INTERIM TECHNICAL EVALUATION REPCRT (3.3.2.4): Reference 2.20 is a brief report that describes the results of a steam exposure test on an exposed General Electric CR 151B terminal block, plus one from another manufacturer. DITER Reference 2.10 is a General Electric Co. letter stating that there is very little difference between the CR 151B and EB terminal olocks. This letter also claims that the materials have good tolerance to nuclear radiations, but provides no evidence to substantiate either of these claims. DITER Reference 2.17 is a report on steam exposure tests conducted on Genatal Electric EB-25 terminal blocks. FRC's review of these qualification documents has resulted in the following findings: 4-61 M =om orn.r.Resea.rch Center A@nklin = =.

DELETED M ATERIAL G PROPRIETARY INFORMAT10N TER-CS257-195

a. The Guidelines require that the model of the tested unit be the same l as that of the equipment being qualified. The type test is valid I only if the installed equipment and tested unit have the same design and materials and closely similar production procedures and stress levels. The Licensee has neither completely identified the installed equipment nor presented an analysis comparing the impact of deviations between the test specimen's design features, materials, and production procedures and those of the installed equipment.

Therefore, an independent conclusion cannot be reached regarding the extent to which the units are similar, and the validity of the tests as evidence of qualification has not been established,

b. The Guidelines require that the temperature / pressure profile during the test envelop the expected service conditions for a time duration equivalent to the period from the initiation of the accident until the service conditions return to normal values. This requirement is considered to be essentially satisfied, even though there were some deviations.

c. The fact that the terminal blocks installed in the plant are enclosed within vented junction boxes, while those tested in Reference 2.10 were fully exposed, does not eliminate the need to include the chemical spray environment in the test programs. Experience has shown that deposits of chemicals and contaminants in the spray

                                                               ~

solution generally are present following test of terminal blocks that are enclosed and that these deposits sometimes contribute to electrical failures. Also, if the terminal blocks are used for signals from electrical transmitters, the presence of moisture, high temperature, chemical spray solution, and nuclear radiations may degrade the signal's accuracy.

d. Contrary to the statement in Reference 2.10, filled phenolics often are strongly susceptible to both thermal and radiation aging (see Appendix C of the Guidelines). Neither thermal nor radiation aging was addressed in the test programs, nor was the large radiation dose associated with a LOCA event. Also, the contribution to the total dose from beta radiation may be significant. The period of qualified life must be established.

LICENSEE RESPONSE: [No response provided.) FRC EVALUATION: FRC makes the following additional comments to support the conclusion presented below: 4 4-62 i Lb Franklin Research Center A won er n. n. a msnm. I

DELETED MATERIAL IS PROPRIETARY INPORMAT'ON TER-C5257-195

1. It has not been shown, either by test or analysis, that terminal block failure is unlikely under the effects of thermal aging, radiation, and steam / chemical spray environments postulated to follow a LOCA event. Also, the Licensee has not stated whether the blocks are exposed or installed within junction boxes and whether the presence of moisture could affect the accuracy of instrumentation signals carried by the blocks.

2. The Guidelines require that equipment must be qualified to integrated nuclear radiation dose levels that (i) reflect the sum of both the normal operating dose (for the qualified life period as a minimum) and the accident dose level, and (ii) takes into account the effects of beta radiation and the proximity of the installed equipment to the sump or other concentrated sources of radiation. In reviewing terminal block qualification data referenced in connection with the Palisades plant, FRC noted that the Westinghouse statement regarding radiation qualification was quoted out of context, and that the situation is unsatisfactory for the long term following a LOCA.

3. Aging degradation has not been addressed as required by the Guidelines. The Licensee should evaluate the susceptibility of the terminal blocks to degradation as a result of exposure to temperature and nuclear radiation during the installed life in the plant. If significant degradation is expected ' to occur, aging must be addressed in the test program and an explicit determination made of qualified life.

4 FRC has reviewed several references which provide statements concerning aging and irradiation effects on the materials used in terminal blocks. It has been stated that the material (wood-flour-filled phenolic) is capable of withstanding continuous service at 125'C. It has also been stated that extrapolated 40-year life temperature ranges from 105*C to 110*C. Other reports indicate that mechanical properties begin to degrade at 0.5 Mrd and that elongation and impact strength are reduced by 25% at 3 to 8 Mrd. The mechanical and thermal properties of wood-flour-filled phenolics are highly variable as shown in Appendix F'. The data reviewed for the EEQ program demonstrate that data scatter on thermal aging is wide (e.g.,171 hours at 150*C = 40 years,160 hours at 136*C = 40 years, 100 hours at 126*C = 11.4 years). FRC considers that meaningful forecasts of lifetime and uniform standards for aging damage have not been established for the wood-flour-filled phenolics.

5. With regard to spray, FRC has reviewed 24-hour tests in which deposits accumulated along mold lines of terminal blocks and grounded a terminal. Examination of various terminal blocks af ter simulated LCCA with chemical spray has indicated conductive deposits on block surfaces that resulted in reduced insulation resistance without complete grounding or short circuit. The Licensee has not analyzed O 4-63 INJEnklin Research Center 4 emon or n. mn.en ==enee 1

4 DELETED MATERIALi$ PROPRIETARY INFoRMATION TER-C5257-195 the effect of high conductivity on instrument signals. Merely maintaining voltage does not assure reliable transmission of level / pressure information. FRC has also reviewed Sandia Report Number SAND 80-2447A presented at the Eighth Water Reactor Safety Research Information meeting held at the National Bureau of Standards from October 27 to 31, 1980. The following statement is presented verbatim from page 1 of the report: Otmar M. Steutzer Sandia National Laboratories Albuquerque, New Mexico 87185 Wire connections in reactor systems are generally made by means of Terminal Blocks (tbs) , small insulating boards, each accommodating from 6 to 12 screwdown metal terminals. Figure 1 shows the three models of tbs used in the containment of Three Mile Island, Unit 2 (TMI-2). The blocks are shielded from dirt, or direct steam impingement, by protective enclosures or circuit boxes, many of them similar to the standard fuse < boxes. The enclosures are not hermetically sealed and are equipped with breathers or " weep-holes," which at TMI-2 are 6 mm in diameter, but in some other reactors are 25 mm wide. During a steam outbreak, steam can therefore reach the tbs by diffusing through these openings. This makes the insulator surface more conductive. Figure 2 indicates what happens: increased leakage currents (f rom terminal-to-ground or to another terminal), noise in the circuits, and potentially total electrical breakdown. tbs have been suspect for a long time. At the urging of the NRC, tbs in safety-related (IE) circuits were replaced in most reactors by splices. At TMI, 620 terminals were eliminated, but there are still 2700 in the containment. And in the case of an accident even non-safety circuits may be important. The report presents data and a statistical evaluation of results for probability of failure as a function of time and voltage. FRC CONCLUSION. This equipment item is assigned to NRC Category V because there is no assurance that the terminal blocks would perform reliably or transmit reliable instrument signals under LOCA conditions. 1

                                                                                                                  )

l 4 4-64 h/J Funklin Rewan:h Center I A Dannien of The Fransen ansamme

DELITED MATERIAL 18 PROPRIETAAY INFORMAT10N TER-CS257-195 4.6.6 Equipment Item No. 56 Solenoid valve Located in the Reactor Building ASCO, Model Not Stated Nitrogen System Valve (V-23-20) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (MUIPMENT ITEM ADDED IN REFERENCE 1): These are normally closed containment isolation valves that will not change position given a failure of the solenoid valve. They are in a non-harsh temperature / pressure environment and the expected one-year integrated radiation exposure is on the order of 0.1 Mrads. This is below the level at which any detrimental effects will occur. Based upon the above discussion, there is no reason to believe these valves will not stay closed. This equipment will either be qualified or replaced by July 1, 1982. FRC EVALUATION: The Licensee has not provided, anc FRC has found no other source of, valid qualification documentation for this solenoid valve. Therefore, qualification has not been established in accordance with the requirements of the Guidelines. FRC's review of the Licensee's justification for continued plant operation (Chapter 7 of Reference 1) for this equipment item is given in Appendix D of this report. As noted in Appendix D, the Licensee has not ' addressed the need for this valve to function in the long-term, post-accident period. The Licensee has not provided any analyses to support the assertion that expected radiation exposure (2.29 Mrd stated on the SCEW sheets) "is i below the level at which any detrimental effects will occur." The Licensee should proceed with preventive maintenance activities on an expedited schedule. The manufacturer should be consulted to obtain recommended replacements for coils and other non-metallic components used in these valees. 4 4-65 Lihl Franidin Research Center a ca an w w , a =

DELETED MATERIAL IS PROPRIETARY INFORM ATION TER-C5257-195 FRC CONCLUSION: l This equipment is assigned to NRC Category V because valid qualification documentation has not been provided. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to a program of equipment qualification or replacement by June 1982. 4.6.9 Equipment Item Nos. 23, 24, and 25 Solenoid Valve Located in the Reactor Building 23: ASCO Model WPLB83177 (V-23-15) 24: ASCO Model 831424 (V-23-16) 25: ASCO Model X8031A42 (V-23-19) Nitrogen System Valves (Final Licensee References 2.6 and 2.7) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These are normally closed containment isolation valves that will not change position given a failure of the solenoid valve. They are in a non-harsh temperature / pressure environment and the expected one-year integrated radiation exposure is on the order of 0.1 Mrads. This is below the level at which any detrimental effects will occur. Based upon the above discussion, there is no reason to believe these valves will not stay closed. FRC EVALUATION: The references cited by the Licensee are not adequately identified and copies were not provided for review. Also, FRC is not aware of valid qualification documentation for this solenoid valve from other sources. Therefore, qualification has not been established in accordance with the requirements of the Guidelines. FRC's review of the Licensee's justification (Chapter 7 of Reference 1) for this equipment item is given in Appendix D of this report. As noted in l Appendix D, the Licensee has not addressed the need for these valves to function during the long-term, post-LOCA period. The Licensee has not provided any analyses to support the assertion that the expected radiation _nidin Rese_ arch _. Center

DELETC MATER;A. is PROMutTARY INFORMAT10N TER-C5257-195 exposure (0.29 Mrd on the SCEW sheets) "is below the level at which any detrimental effects will occur." The Licensee should proceed with preventive maintenance activities on an expedited schedule. The manufacturer should be consulted for recommended replacements for coils and elastomer parts used in these valves. FRC CONCLUSION: This equipment is assigned to NRC Category V because valid qualification documentation has not been provided. The solenoid valve should either be qualified or replaced as in the case of Equipment Item No. 56. 4.6.10 Equipment Item 12D Pressure Switches Located in the Peactor Building Barton Model 288A Isolation Condenser Pressure Switches (IB-0 5-A1, -A2 ; IB-05-B1, -B2; IB-ll-A1, -A2; IB-ll-B1, -B2) (Final Licensee References 2.7 and 2.10) CRIGINAL TEXT TAKEN FRCM DRAFT INTR. RIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These switches are provided to sense a sudden pressure change in the Emergency Cond.!nser system following a break in the emergency condenser line. The switches, after sensing the pressure change, will also isolate I the Emergency Condenser system (closure of the isolation valves) . The initiation of the isolation valve closure takes place 40 seconds (35 seconds + 5) af ter the line break. The time delay is provided to avoid a t spurious trip due to a pressure surge when the Emergency Condenser is put I into service under a normal operation. Once a signal for valve closure is initiated by the switches, the valve will complete its closure regardless of the availability of these pressure switches. Therefore, a failure of the switches af ter the initial 40 seconds will not prevent a closure of the isolation valves. Cur analysis indicates that the peak temperature at the valve location at 40 seconds following the break is 170*F. The test report obta taed from the switch manufacturer shows that the switch (Barton 288A) did not experience malfunction or physical damage at a test temperature of 212*F. Further, the radiation test given in the test report indicates that the switch operated normally after a radiation exposure of 3 Mrads. Our analysis shows that these switches could experience up to 0.39 Mrads af ter a full year of radiation exposure due to the accident, based on an extremely conservative assumption of 100% fuel failure in the reactor core. gg 4-67

       /... Franklin
             . w w n.Research.
                       %       Center u.

DELETED MATERIAL 38 PROMtsETARY WFORMATioN TER-C5257-195 FRC EVALUATION: The Licensee has referenced qualification documentation, but did not provide copies for review. Because the switch provides an important function, qualification documentation is necessary. The Licensee should provide additional information to demonstrate that the possible failure of the switch will not degrade the associated safety-related electrical circuit when the switch is exposed to the harsh conditions of a FSLB/HELB outside the drywell containment. The Licensee should also estimate the qualified life of the switch and analyze maintenance. surveillance records for any abnormal behavice which might limit qualified life. Qualification of the switch requires that the switch be shown to be operable for one hour plus its normal operating time.

                                                         ~

FRC CONCLUSION: This equipment is assigned to NRC Category V because qualification documentation was not submitted by the Licensee for review. The Licensee should furnish a statement with supporting documentation on qualified life in accordance with Section 4.1.3. 4.6.11 Equipment Item No. 33 Position Switches Located in the Steam Tunnel Snaplock (NAMCO) Model SL3-C58W MSIV Position Indicaters (NS-04A-1, -2; NS-04B-1, -2) (Licensee reference not cited) ORIGINAL TEX

• TAKEN FROM DRAF* INTERIM TECHNICAL EVALUATION REPORT:

None LICENSEE RESPONSE (EQUIPMENT ITDi ADOED IN REFERENCE 1):

       *he MSIV solenoid valves are used to direct instrument air to hold open the outside containment main steam isolation valves. The FSIV position indication switches are utilized to provide a scram signal when the MSIVs are less than 90% open.

pg 4-68

      .... Frankhn Researen a w wwea   w     C.e=nter

DELETED WATERIAus PROMNETARY INPORMATION TER-C5257-195 A loss of power or air to the MSIV solenoids causes the MSIVs to fail in the safe direction, closed. Also, redundant protection is provided by' the inside containment isolation valves that would not be affected by the environment created by outside containment breaks. In the event the outside containment MSIV position switch did not provide a scram signal, two scram signals would still be available to ensure the reactor was shut down insiediately for a main steam line break. These two signals are the MSIV position switch signal from the inside valves and the reactor low water level signal, both of which would not be affected by the harsh environment created during this event. The one-year integrated accident exposure of thee components is at 1. east two orders of magnitude below that which would cause any degradation. Based upon the above discussion, it is expected that the main steam isolation function and reactor scram function required to mitigate outside containment will be accomplished. FRC EVALUATION: Because it provides an important safety-related function, the switch must be qualified for a MSLB in the steam tunnel. The Licensee did not cite any qualification reference that would demonstrate operability of the switch under accident conditions. The limit switch is required to operate for the short-term period of a postulated MS~2 in the steam tunnel. Because in the event of a small break a harsh temperature condition could exist for an extended time period, it is necessary to demonstrate qualification for a minimum of 1 hour. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. FRC CONCLUSICN: 1 These switches are assigned to NRC Category V because no documentation J has been provided to support qualification. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has  ; committed to equipment qualification or replacement by June 1982.  ; A 4-69 r..J Frandn a e- e Researe.n

n. n . C. enter

DELETED MATERIAL is PROMtIETARY MPORMATION i l TER-C5257-195 4.6.12 Equipment Item No. 55 (previously designated Il2) Relief Valve Operator Located Within the Drywell Dresser Model 1525 VX (previously shown as General Electric equipment) Power Operated Relief Valves (Licensee Reference 2.1) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM. TECHNICAL EVALUATION REPORT (3.3.2.8): Licensee Reference 2.1 describes a steam exposure test performed on a Dresser Electromatic Relief Valve. FRC has the following comments with regard to this reference:

a. It is not evident that the equipment installed in the plant is the same as the item tested. The Guidelines require that the test specimen be the same as the equipment being qualified. The Licensee did not present an analysis comparing the impact of deviations between the test specimen's specific design features, materials, and production procedures and those of the installed equipment. Hence, the validity of the cited test as evidence of qualification has not been established.

b. Although the temperature / pressure profile in the test chamber i

enveloped the service conditions for an adequate time duration, the

 !                                                                  test did not include a chemical spray exposure. Because the effects l                                                                 of added moisture and chemical residues may be more damaging than steam alone, FRC concludes that the absence of the chemical spray environment is a potentially serious deficiency.

c. The Guidelines require that radiation exposure should be applied during the test sequence concurrent with, or prior to, the steam exposure, unless it is k-own that the device contains material; that are not subject to degradation by nuclear radiations. The materials used in this icem have not been so identified. FRC concludes that degradation due to irradiation of this item must be addressed, preferably by a test involving simultaneous exposures to steam, chemical spray, and gamma radiation in order that the effects of gamma, heating, and other insulation stresses be accurately simulated.

d. Aging degradation has not been considered, nor has the qualified life been established, nor has a program to ascertain whether any in-service failures during the installed life of the equipment are the result of aging degradation, as are required by the Guidelines.

LICENSEE RESPONSE: Note C of SCEW Sheet I-9 states that this equipment will either be replaced or qualified by July 1, 1982. 4-70 4 h.f) %Franklin o r, Re.. sea.rch Center

           . _ _ _ . _ _    -_ . . _ _ _ _ , _ _ . _ , , _ _ _ _ _ . _ _ _ _ , - , _ , _ . . . _ _ _ _ _ _                   .,,,,m___.,_.        .____, _ _,_ _ . . , _ _ _ _ , . . . . . - , _ _ _ . . _ _ . , , _ _ _ , , _ _ _ _ .

DELETED WATEAIAL 18 MomIETARY #eP0AMATION TER-C5257-195 FRC EVALUATION: The Licensee has provided no additional qualification information. Hence, the cominents of the DITER still apply. FRC CONCLUSION: This equipment is assigned to NRC Category V because evidence of qualification has not been provided. It is noted that the Licensee plans to qualify or replace this equipment by July 1, 1982. 4.6.13 Equipment Item No. 21B Solenoid Valves Located in the Reactor Building' ASCO Model 83148 i Emergency Condenser Makeup Valves (V-11-34 and V-ll-36) (Licensee Reference 2.11) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ACDED IN REFERENCE 1): These valves provide makeup to the isolation conde..sers. With the minimum water level permitted by technical specifications, the emergency condensers will be available to remove heat at their design capacity , without uncuvering the heat exchanger tubes for 1 hour 40 minutes with l both condensers available and 45 minutes if only one condenser is i available. l l The emergency condenser syst'em is one of the methods available to control l reactor pressure and cool down the plant following a HELB. Since the l emergency condenser line break is the break that causes the harsh l environment, it is likely that one of the alternate cooldown methods would be utilized. In the area of the emergency condensers, there are temperature detectors l that will detect leaks in the emergency condenser system and annunciate i this in the control room. By procedure, the control room operator would isolate the affected system before a rupture developed. Therefore, the actual temperature / pressure environment would not reach the levels indicated in the worst-case analysis. The one-year integrated radiation exposure to those components is in the order of 0.5 Mrads and an evaluation shcws that there will be no detrimental effects with exposure of up to 1 Mrads. l gg 4-71

        ...; Franidin Research Center a:>. a.e N n   a==su.

DELETfD MATERialis PROMWETARY INPoAMaTION TER-C5257-195 Based on the above discussion, it is evident that the ability to achieve cold shutdown will not be adversely affected by the potential environmental effects of these components. Our evaluation of the component materials revealed that this component contains thermal aging and radiction-sensitive materials (Buna-N and/or fish paper) . Therefore, the sensitive component materials will be replaced by June 1982. FRC EVALUATION: The reference cited by *.he Licensee is not adequate to demonstrate qualification. Also, FRC is not aware of qualification documentation for this solenoid valve from other sources. Therefore, qualification has not been established in accordance with the requirements of the Guidelines. A review of the Licensee's justification (Chapter 7 of Reference 1) for continued plant operation with this equipment item is given in Appendix D of this report. FRC CONCLUSION: , This equipment is assigned to NRC Category V because valid qualification documentation has not been provided. Although the Licensee's evaluation of this equipment item has not been completed, the Licensee has committed to equipment qualification or replacement by June 1982. 4.6.14 Equipment Item No. 10 Pressure Switches Located in the Reactor Building General Electric MAC Model 552 Core Spray Pressure Switches (RV-26A,B; RV-40A,C) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION. REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): Evaluation of the HELB indicates that two switches (RV-40B and RV-40D) will remain in the ambient temperature and atmospheric pressure conditions (77'F and 15 psia) . Evaluation of the component material shows that the material having the most susceptibility to radia? ion is sheet fiber (fish paper). According to the DOR Guidelines, the radiation I susceptibility threshold value of this material is in the order of 0.1 rad, which is the same order of magnitude that these switches may pg 4-72 a..' Frankhn a w w 7%. Researe.n r, u.Center

DELETED MATERfAL 18 P90MhtTARY IMPORMADON TER-C5257-195 I experience af ter a full year of continuous exposure from the reactor

coolant containing 100% of noble gas, 50% of halogen, and 14 of others.

Therefore, due to the extremely conservative assumptions used in the radiation analysis, we believe that these switches will function following a HELB. Therefore, six of the ten core spray pressure switches provided (see Items 8 and 12) will be available af ter a worst-case HELB enabling core spray system to function properly. [This equipment item comes from the four units which are in harsh environment. The SCEW sheet notos that this equipment will either be replaced or qualified by July 1, 1982.] FRC EVALUATION: Because the switches provide an important function, qualification documentation is necessary and the Licensee should provide information to demonstrate that the possible failure of the switch will not result in the j degradation of the associated safety-related electrical circuit when the ! switch is exposed to the harsh conditions of a MSLB/RELB out>,ide the drywell containment. The Licensee should also estimate qualified life and analyze maintenance surveillance records for any abnormal behavior which might limit qualified life. l The Licensee stated that satisfactory system initiation function would occur with 8 of ;l drywell containment switches remaining operable. Because the Licensee has not provided electrical system diagrams, FRC could not confirm that the tystem response would not be degraded. The Licensee stated that the switch will either be replaced or qualified by July 1, 1982. , FRC CONCLUSION: l This equipment is assigned to NRC Category V because there is no evidence of qualification. The Licensee has committed to qualify or replace this l equipment by July 1, 1982. I l O 4 ', a

         ;d) FranMin
              - ,_ Research._ Center

DELETED MATERIAL 16 PROPRIETARY INFORM ATioN TER-C5257-195 4.6.15 Equipment Item No. 12C Level Switches Located in the Reactor Building Barton Model 288A Reactor Water Level Switches (RE-18-A through -D) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENI' ITEM ADDED IN REFERENCE 1): The RE-18 switches provide a low-low-low (triple low) signal to the automatic depressurization circuit. This signal could be necesssary if there were a small break that required a rapid depressurization in order to permit a core spray injection. The breaks that cause the harsh environment for these switches do not require the use of the automatic depressurization system. It should be noted that, regardless of the condition of the RE-18 switches, the electromatic relief can be manually initiated by the control room operator if he desires to use them for a cooldown.

                                                          ~

The RE-05 switches provide a reactor high-pressure scram signal and control roem water-level indication. They are redundant and physically separated. Another important consideration in evaluating the potential failure of these components due to HELBs is the fact that both the emergency condenser and cleanup line areas are monitored by area temperature detectors. These detectors will warn the control operator of leaks in those systems long before the pipes rupture. This will enable the operato: to isolate the leak before the harsh environment is established. The one-year integrated accident radiation exposure these components might see is about one order of magnitude less than the level that might cause an adverse effect on the most sensitive material used. [The SCEW sheet states that this equipment will be qualified or replaced by July 1, 1982.] FRC EVALUATION: Because the switch provides an important function, qualification documentation is necessary. The Licensee should also provide information to demonstrate that the possible failure of the switch wil'1 not result in the degradation of the associated safety-related electrical circuit when the switch is exposed to the harsh conditions of a MSLB/HELB outside the drywell containment. A 4-74 du Franklin Research Center A Diwanyt af The Fransen eneBRAe

i

                                                                                     )

DELATED MATEMAL is f uoPRtETARY mePORMAnoN

                                                .                                    1 1

TER-C5257-195 l l 1 The Licensee should also estimate qualified life and analyze maintenance surveillance records for any abnormal behavior which might limit qualified 1 life. l l 1 FRC CONCLUSION: This equipment is assigned to NRC Category V because there is no evidence of qualification. The Licensee has committed to qualify or replace this equipment by July 1, 1982. ! 4.6.16 Equipment Item Nos.14A and 15 l Pressure Switches Located in the Reactor Building l 14A Barksdale Model B2T-Al2SS

Core Spray Pressure Switches (RE-17-A through -D) i 15

Barksdale Model E2T-M12SS (Licensee reference not cited) l l

ORIGINAL TEXT TAKEN FRCM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IM REFERENCE 1): These pressure switches monitor reactor prtssure and are interlocked with the core spray auto initiation logic to prevent core spray injection valves feca opening until reactor pressure is below 285 psig. The core spray system consists of two redundant single failure-proof, low pressure core spray systems. The two postulated HELBs that would create I a harsh temperature environment in these areas are cleanup line rupture j and emergency condenser line ruptures. Both of these postulated breaks do not require the initiation of core spray for ECCS purposes. In these l scenarios, the core spray would be utilized by the control room operator as a safety grade safe shutdown system for reactor water makeup if no high pressure means were available. In that case, these switches are not necessary since the remote manual operation of the core spray injection valves will not be affected by the condition of these switches. The integrated one-year exposure of these components under accident conditions is on the order of 10 Krads; this is significantly below the l level of 0.1 Mrads that would cause any detrimental effects on the most i sensitive material used in them. Based on the above considerations, it is expected that the core spray j system will be able to perform its ECCS function for inside containment 1 breaks and also provide a safety grade reactor makeup capability for  ! 4 4-75 l MJ Franklin Research Center s>~m e nmewweemum , l 1 l

DELETED M ATERIAL IS PROPRIETARY INFORM ATION TER-C5257-195 l cleanup or emergency condenser line breaks outside containment. (The SCEW sheet notes that this equipment will either be replaced or qualified by July 1, 1982.] FRC EVALUATION: , The area where the safety-related switches are located is relatively mild, except for radiation expor2re, for the accident conditions that the switch is designed to mitigate. The Licensee has not addressed the question of whether these switches could incorrectly indicate a pressure of less than 285 psig under HELB conditions. It is not clear that such an occurrence is a

 " safety failure."

Because the switch provides an important function, qualification documentation is necessary, and the Licensee should provide information to demonstrate that the possible failure of the switch will not result in the degradation of the associated safety-related electrical circuit when the switch is exposed to the harsh conditions of a MSLB/HELB outside the drywell containment. The Licensee should also conservatively estimate the qualified life and analyze maintenance surveillance records for any abnormal behavior which might limit qualified life. The Licensee stated that the switch wou'..d either be qualified or replaced by July 1, 1982. FRC CONCLUSION: This equipment is assigned to NRC Category V because there is no evidence of qualification. The Licensee has committed to qualify or replace the equipment by July 1, 1982. 4.6.17 Equipment Item No. 18 Level Switches Located in the Reactor Building Yarway Model C2337 Reactor Water Level Switches (RE-02-A through -D) (Final Licensee References 2.11 and 2.12) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None gg 4-76 WO.%Franklin Research t w r ama === . Ce,nter

DELETED Waramaus PacemsTannm% . TER-C5257-145 LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These switches provide an auto start signal to core spray, a containment isolation signal, a reactor isolation signal, and one of the signals required for an automatic containment spray start. These switches are redundant and physically separated. Another important consideration in evaluating the potential failure of these components due to HELBs is the fact that both the emergency condenser and cleanup line areas are monitored by area temperature detectors. These detectors will warn the control room operator of leaks in those systems long before the pipes rupture. This will enable the operator to isolate the leak before the harsh environment is established. The one-ycar integrated accident radiation exposure these components might see is about one order of magnitude less than the level that might cause an adverse effect on the most sensitive material used. Based on the above discussion, it is expected that the safety function required by these switches will be accomplished for the postulated HELBs outside containment that create the harsh environment. Reactor Water Level Switches and Reactor Water Level Switches / Transmitters: RE-18-A through RE-18-D, RE-05-19A, and RE-05-19B The RE-18 switches provide a low-low-low (triple low) signal to the automatic depressurization circuit. This signal could be necesssary if there was a small break that required a rapid depressurization in order to permit a core spray injection. The breaks that cause the harsh environment for these switches do not require the use of the Automatic j Depressurization System. It should be noted that, regardless of the i condition of the RE-18 switches, the electromatic relief can be manually ) initiated by the control room operator if he desires to use them for a cooldown. The RE-05 switches provide a reactor high-pressure scram signal and control room water-level indication. They are redundant and physically separated. Another important consideration in evaluating the potential failure of these components due to HELBs is the fact that both the emergency condenser and cleanup line areas are monitored by area temperature detectors. These detectors will warn the control operator of leaks in those systems long before the pipes rupture. This will enable the operator to isolate the leak before the harsh environment is established. The one-year integrated accident radiation exposure these components might see is about one order of magnitude less than the level that might cause an adverse effect on the most sensitive material used. 4-77 C a Franklin aon,anwn Research.a r==a n,enter

DELETED M ATERtAL IS PROPRf ETA RY IN FORM AT10N TER-C5257-195 ! FRC EVALUATION: The Licensee has referenced qualification documents, but did not make them available for review. The License

• has stated that the documents will be made available to provide evidence of qualification.

Because the switch provides an important function, qualification documentacion is necessary, and the Licensee should provide information to l demonstrate that the possible failure of the switch will not result in the degradation of the associated safety-related electrical circuit when the switch is exposed to the harsh conditions of a MSLB/HELB outside the drywell containment. I The Licensee should also estimate qualified life and analyze maintenance surveillance records for any abnormal behavior which might limit qualified life. The Licensee has stated that the equipment will be replaced er will be qualified by July 1, 1982. It should be noted that' qualification requirements state that the switch should be qualified for at least one hour plus its normal safety-related operational time. FRC CONCLUSION: This equipment is assigned to NRC Category V because evidence of qualification has not been provided. The Licensee has committed to qualify or replace this equipment by July 1,1982. l l l l l 4_ 4-78 M Franklin Research Center A W of The Frename eneceute

DELETED MATEMAL IS PMCPfeETARY INFORMATON TER-CS257-195 4.7 NRC Category VI EQUIPMENT FOR WHICH QUALIFICATION IS DEFERRED The equipment items in this section have been addressed by the Licensee in the equipment environmental qualification submittals; however, the qualification review has been deferred by the NRC in accordance with criteria

presented in Sections 2.2.3 and 2.2.5 of this report.

4.7.1 Equipment Item No. 5 Pressure Switches Located in the Reactor Building Static-O-Ring Model 12NKA Drywell Hi-Pressure Scram Switch (RE-04A through RE-04D) (Final Licensee References 2.6 and 2.7) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None i LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These switches are installed just outside the Drywell wall, and the peak temperature and pressure to be experienced by these switches are 230*F and 16 psia following an emergency condenser break outside the Drywell. However, these switches are installed to tronitor the pressure inside the ( Drywell and are not required to mitigate a HELB outside the Drywell. l l l l FRC EVALUATION: FRC agrees with the Licensee's position that the switches will only be exposed to a mild environment for any accident that they are designed to mitigate. Integrated radiation exposures are as high as 1.5 Mrads and could be significant, but an evaluation could not be made because the Licensee did not submit the references for review. In addition, no qualified life assessment in accordance with Section 4.1.3 was presented for this equipment. FRC CONCLUSION: This pressure switch is assigned to NRC Category VI because it is I believed by the Licensee to be lccated in a nonharsh area for the accident condition that it is designed to mitigate. Its review is therefore deferred i i 4 4-79

                       .01) Franklin Research Center on    .# wwwa naam.

l

DELETE 0 MATE ~JAL G PROPRIETAT4Y INFORM ATION TER-C5257-195 until af ter February 1,1981 as discussed in Section 2.2.3. The Licensee should make its references available for qualification verification. 4.7.2 Equipment Item No. 9 Pressure Switches Located in the Reactor Building Mercoid Model 9-51/ DAW-43-156-R2IE Core Spray Pressure Switches (RV-29A through D, KV-40B,D) (Licensee reference not cited) . ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): RV-29A through D: Evaluation of the HELB indicates that these switches will remain in the ambient temperature and atmospheric pressure conditions (77'F and 15 psia). Evaluation of the component material shows that the material having the most susceptibility to radiation is sheet fiber (fish paper) . According to the DOR Guidelines, the radiation susceptibility threshold value of this material is in the order of 0.1 Mrads, which is the same order of magnitude that these switches may experience after a full year of continuous exposure from the reactor coolant containing 100% of noble gas, 50% of halogen, and 1% of others. Therefore, due to the extremely conservative assumptions used in the radiation analyses, we believe that these switches will function following a HELB. RV-40B, D: Evaluation of the HELB' indicates that two switches (RV-40B and RV-40D) will remain in the ambient temperature and atmospheric pressure conditions (77'F and 15 psia) . Evaluation of the component material shows that the material having the most susceptibility to radiation is sheet fiber (fish paper). According to the DOR Guidelines, the radiation susceptibility threshold value of this material is in the order of 0.1 Mrads, which is the same order of magnitude that these switches may experience af ter a full year of continuous exposure from the reactor coolant containing 100% of noble gas, 50% of halogen, and in of others. Therefore, due to the extremely conservative assumptions used in the radiation analysis, we believe that these switches will function follcwing a HELB. Therefore, six of the ten core spray pressure switches provided (see Items 8 and 12) will be available af ter a worst-case HELB enabling the core spray system to function properly. 4 4-80 MEnkhn Research Center

       % .a n. r = a m.

DELETED MATERtAL is PeOPfmETAAY lNPORMATION TER-C5257-195 FRC EVALUATION: The area where the safety-related switches are located is relatively mild, except for radiation exposure, for the accident condition that the switch is designed to mitigate [1]. The review of the switch's qualification is therefore deferred until after February 1, 1981, as discussed in Section 2.2.3. The Licensee should also estimate qualified life and analyze maintenance records for any abnormal behavior which might limit qualified life. The Licensee has stated that the switch will either be replaced or qualified by July 1, 1982. FRC CONCLUSION: This equipment is assigned to NRC Category VI because it is believed by the Licensee to be located in a nonharsh area for the accident condition that it is intended to mitigate. The review of this equipment is deferred until af ter February 1,1981, as discussed in Section 2.2.3. Also, the Licensee should furnish a statement on qualified life in accordance with Section 4.1.3. 4.7.3 Equipment Item No. 12A Pressure Switches Located in the Reactor Building Barton Model 288A Containment Pressure Switches (lP15-A through -D) (Final Licensee References 2.7 and 2.10) ORIGINAL TEXT TAKEN FRCM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPME!Tr ITEM ADDED IN REFERENCE 1): The peak temperature and pressure seen by these switches are 200*F and 16 psia, respectively, following an emergency condenser line break cutside containment (worst-case HELB). The contaim ent pressure switches are provided to monitor the pressure inside containment and are not required to mitigate a line break outside containment. Drywell Pressure Switch (RV-46-A through -D) (SCEWS Nos.: 55-58] The peak temperature and pressure seen by chese switches are 230*F and 16 psia, respectively, following an emergency cendenser line break outside I containment (worst-case HELB) . The Drywell (contalment) pressure  ! I 4 4-81

     .t.) Franklin a coma w Research.C.

n. F n a . w. enter 1

DELETED M ATERIAL IS PROPRIETARY INFORMATION TER-CS257-195 switches are provided to monitor the pressure inside containment and are not required to mitigate a line break outside containment. FRC EVALUATION: The area where the safety-related switches are located is relatively mild, except for radiation exposure, for the accident condition that the switch is designed to mitigate [1]. The review of the switch's substantiating qualification is therefore deferred until after February 1, 1981 as discussed  ; in Section 2.2.3. The Licensee has referenced qualification documents, but they were not made available for review. The Licensee has stated that the documents will be made available to provide evidence of qualification. The Licensee should also estimate qualified life and analyze maintenance surveillance records for any abnormal behavior which might limit qualified life. There is a concern, however, that failure of these pressure switches could result in the inadverten: actuation of the containment spray syster by the plant operator because of an incorrect signal from an unquelified instrument. Although the use of the spray system could result in a vacuum condition occurring in the drywell, there is no concern as long as the torus vacuum relief valve system is fully qualified, as mentioned in Section 4.1.1. It seems prudent for the Licensee to provide qualified pressure switches to more adequately ensure against possible inadvertent containment spray system actuation. FRC CONCLUSION: This equipment is assigned to NRC Category VI because it is believed by the Licensee to be located in a nonharsh area for the accident condition that it is intended to mitigate. The review of this equipment is deferred until after February 1, 1981, as discussed in Section 2.2.3. The Licensee should furnish a statement on qualified life in accordance with Section 4.1.3. 4-82 4_Franidin TJLU Research Center A Dmempn W The Franda insthme

DEATIO MATIAMt W PRCPRWTARY WPORMADCN { TER-C5257-195 4.7.4 Equipment Item No. 128 Barton Model 288APressure Switches Located in the Reactor Building Peactor Isolation Switches (RE-22-A through -H) (Final Licensee References 2.7 and 2.10) ORIGINAL tene TEXT TAKEN FROM DRAFT I!CERIM UATION REPORT: TECHNICAL EVAL LICENSEE RESPONSE (EQUIPMENT):ITEM ADDED IN REFERENCE Due to their location, these switches will remain i environment with respect to temperature (77'F) n non-harsh following a EELB. and pressure (15 psia) of continuous exposure from the reactor coolant containinThe radiat g 100% of noble r gas, 50% of halogen, and 1% of others is less than 61 Krads . The component material having the most susceptibilit Viton. y to radiation is susceptibility level of 1 Mrads.According to our raference, this material (viton) has a radiation switches will function properly following a HELBTherefore, we believe that these FRC EVALUATION: The area where the safety-related switches are elocat d i s relatively mild, except for radiation exposure, for the accident co di i switch is designed to mitigate (1) . n t on that the qualification in Section 2.2.3. is therefore deferred until after February 1The review of t

                                                                         , 1981, as discussed The Licensee has referenced qualification document them available for review.                                          s, but did not make be made available to provide evidence                      cation.

of qualifiThe Licensee has stated The Licensee should also estimate qualified life surveillance records for any abncemal behavior which life. might li iand analyze maintenan m t qualified FRC CONCLUSION: This equipment tha Licensee to be located in a nonharsh ais assigned etoeved it NRC by Category VI because is intended to mitigate. The review of this equipmentrea for the accident condition that is deferred until 4 4-83 JLUscen FranMin a e % r Researen Center w.en m ue

DELETED MATERIAL. ls P ROPRIETARY INFORM ATION TER-C5257-395 I 3 The Licensee should aftet

        -4uruary 1, 1981, as discussed in Section 2.2. .

furnish a statement on qualified life (see Section 4.1.3) . 4.7.5 Equipment Item No. 14BTressure Switches Located in the Reactor Building Barksdale Model B2T-A12SS Reactor Pressure Switches (RE-03-A through -D) (Licensee reference not cited) ION REPORT: ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUAT None LICENSEE RESPONSE (EQUIPMENI ITEM ADDED IN REFERENCE 1) ide a ttcram signal These pressure switches are the switches used to provThis is not the scram sign on reactor high pressure. utilized to shut down the reactor in the event of a ruptureThes Emergency Condenser or the cleanup system.Also, the most severe temperature conditions and physically separated. For normal are caused by different HELBs for the redundant switches. d turbine trip pressurization transients (MSIV closure, turbine Anothertrip, anthese switches would important without bypass valves) ,immediately (less than 60 seconds) . ponents due function almost in evaluating the potential f ailure of these comser and cleanup line consideration d to EELBs is the fact that both the emergency con en These detectors will areas are monitored by area temperature detectors.s t long before the the pipes warn rupture. the control operator of leaks in those sys emThis will enable the o harsh environment is established. components h The one-year integrated accident radiation exposure h t voslevel that might d might see is about one order of magnitude less than t ecaus h ctor will Based upon the above discussion, it is expected that tency e reacondenser scram the postulated rupture of the cleanup system or emergAlso, the ability to scram system due to a reactor low water signal. vented. reactor on pressurization transients will not be impeded d or replaced or pre (The SCEW sheet notes that this equipment will be qualifie by July 1, 19 82.] FRC EVALUATION: is relatively The area where the safety-related switches are located mild, except for some radiation exposure, for the accident condition that the 4-84 pg ggg.ms3ntu ,

DELETED MATERIAL is PROPmetTARY INPORMATION TER-CS2'a7-195 switch is designed to mitigate [1]. The review of the switch's substantiating qualification is therefore deferred until af ter February 1,1981, as discussed ) in Section 2.2.3. Because the switch provides an important function, qualification is necessary for the environmen : co which the equipment is exposed. The Licensee should provide information to demonstrate that the possible failure of the switch will not result in the degradation of the associated safety-related electrical circuit when the switch is exposed to the harsh conditions of a MSLB/HELB outside the drywell containment. The Licensee should also conservatively estimate qualified life and analyze maintenance surveillance records for any abnormal behavior might limit qualified life. The Licensee has stated that the switch will either be replaced or qualified by July 1, 1982. FRC CONCLUSION: This equipment is assigned to NRC Category VI because it is believed by the Licensee to be located in a nonharsh area for the accident condition which it is intended to mitigate. The review of this equipment is deferred until after February 1, 1981, as discussed in Section 2.2.3. At that time, the Licensee should provide the references necessary to justify qualification of the equipment. Also, the Licensee should furnish a statement on qualified life in accordance.with Section 4.1.3. . 4.7.6 Equipment Item No. 16 Pressure Switches Located in the Reactor Building Barksdale Model B2T Reactor Pressure Switches (RE-15-A through -D) (Licensee reference not cited) ORIGINAL TEXT TAKEN FRCM DRAFT INTERIM TECHNICAL EVALUATION REPORT: Ncne l LICENSEE RESPONSE (ECUIPMENT ITEM ADDED IN REFERENCE 1): l l These pressure switches are utilized to automatically trip all recirculation pumps and initiate emergency condensers on a reactor high pressure signal. These switches are redundant and physically separated, l 4-85 4

         .... Franklin
                  =o  e Researc.h

n. n Center

ceLETED WATEmw.is PmossegTAny e poswation. TER-C5257-195 Also, the :sost severe temperature conditions are caused by different HC.Bs for the redundant switches. For normal pressurizatien transients (MSIV closure, turbine trip, and turbine trip without bypass valves), these switches would carry out their safety function almost issnediately (less than 60 seconds). Another important consideration in evaluating the potential failure of these components due to EELBs is the fact that both emergency condenser and cleanup line area are monitored by area temperature detectors. These detectors will warn the control operaters of the leaks in these systems long before the pipes rupture. This will anable the operator to isolate the leak before the harsh environ =ent is established. The one-year integrated accident radiation exposure these components might sec h about one order of magnitude less than the level that might cause an adverse effect on the rest sensitive material used. Based on the above discussion, it is reasonable to assume these cceponents would function if needed for a pressurization transient. FRT EVALCATION: The area where the safety-related switches are located is relatively mild, except for radiation exposure, for the accident condition the switch is designed to mitigate [1] . The review of the switch's substantiating qualification is therefore deferred until af ter Februa.y 1,1981, as discussed in Section 2.2.3. Becar _ we switch provides an important function, qualification documents. tion is necessary and the Licensee should provide additional information to dezcastrate that the pcssible failure of the switch will net result in the degradation of the associated safety-related electrical circuit when the switch is exposed to the harsh conditions of a MSLB/m* cutside the drywell containment. The Licensee should also estimate qualified life and analyze maintenance surveillance records for any abnormal behavior which might li=it qualified life. The Licensee has stated that the switch will either be replaced ce qualified by July 1, 1982. 1.-86 f

     .... Frenicn
           . w w % Researen a- . C. enter

ceLETe0 MAftRsAL 18 PROPmETARY INFORMATION TER-C5257-195 FNC CONCLUSION: This equipment is assigned to NRC Category VI because it is believed by the Licensee to be located in a nonharsh area for the accident condition it is intended to mitigate. The review of this equipment is deferred until after February 1, 1981, as discussed in Section 2.2.3. At that time, the Licensee should provide the references necessary to justify qualification of the equipment. Also, the Licensee should furnish a statement on qualified life in accordance with Section 4.1.3. 4.7.7 Equipment Item No. 17 Level Switches Located in the Reactor Building Yarway Model 4316E Reactor Water Level Switches (RE-05-A,B) (Final Licensee References 2.11 and 2.12) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): These water level switches, along with RE-05-19A and RE-05-19B, provide the signal to scram the reactor at a low water level. These switches are redundant and physically separated. These switches would carry out their safety function almost immediately (less than 60 seconds) . Another important consideration in evaluating the potential failure of these components due to HELBs is the fact that both the emergency condenser and cleanup line areas are monitored by area temperature detectors. These detectors will warn the control room operator of leaks in those systems long before the pipes rupture. This will enable the operator to isolate the leak before the harsh environment is established. The one-year integrated accident radiation exposure these components might see is about one order of magnitude less than the level that might cause an adverse effect on the most sensitive material used. Based on the above discussion, it is reasonable to assume these { components would function if needed for a pressurization transient. FRC EVALUATION: The area where the safety-related switches are located is relatively mild, except for radiation exposure, for the accident condition that the switch is designed to mitigate [1] . The review of the switch's substantiating qualification is therefore deferred until af ter February 1,1981, as discussed in Section 2.2.3. p 4-87 J.r; Franklin Researen Center a ca a e n. a a .

DELETED W ATERIAL is PROPMIETARY WPOAMATION TER-CS257-195 The Licensee has referenced qualification documents, but did not make them available for review. The Licensee has stated that the documents will be made available to provide evidence of qualification. Because the switch provides an important safety-related function, qualification documentation is necessary, and the Licensee should provide information to demonstrate that the possible failure of the switch will not result in the degradation of the associated safety-related electrical circuit when the switch is exposed to the harsh conditions of a MSLB/HELB outside the drywell containmer.t. The Licensee should also estimate qualified life and analyze the switch's maintenance surveillance records for any abnormal behavior whi:th might limit qualified life. The Licensee has stated that the equipment will either be aplaced or qualified by July 1, 1982. It should be noted that qualification tequirements state that the switch should be qualified for at least one hour plus its normal safety-related operational time. FRC CONCLUSION: This equipment is assigned to NRC Category VI because it is believed by the Licensee to be located in a nonharsh area for the accident condition it is intended to mitigate. The review of this equipment is deferred until af ter February 1, 1981, as discussed in Section 2.2.3. At that time, the Licensee should provide the references nee _essary to justify qualification of the equipment. Also, the Licensee should furnish a statement on qualified life in accordance with Section 4.1.3. 4.7.8 Equipment Item No. 38 Pressure Switch Located in the Reactor Building Meletron Model 4201E-3B Drywell Pressure Containment Isolation Valve Switch (PS-153) (Final Licensee Reference 2.7) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None 4-88 4..r. Frankhn Research Center

• W at he Frenen sneewe

i DELETED MATEMAL IS PROPMETARY INFORMATION TER-C5257-195 , l LICENSET. RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): This is a high drywell pressure switch that closes a selected group of isolation valves. Since the function is to isolate the containment in the event of breaks inside containment, it is not required to mitigate any outside containment breaks. It, therefore, does not need to be qualified for this application. FRC EVALUATION: The area where the safety-related switches are located is relatively mild, except for radiation exposure, for the accident condition that the switch is designed to mitigate [1]. The review of the switch's substantiating qualification is therefore deferred until af ter February 1,1981, as discussed

in Section 2.2.3.

The Licensee has referenced qualification documents, but did not make them available for review. The Licensee has stated that the documents will be made available to provide evidence of qualification. Because the switch provides an important function, qualification documentation is necessary, and the Licensee should provide additional j information to demonstrate that the possible failure of the switch will not l result in the degradation of the associated safety-related electrical circuit when the switch is exposed to the harsh conditions of a MSLB/HELB outside the drywell containment. The Licenses should also estimate qualified life and analyze maintenance surveillance reecrds for any abnormal behavior which might limit qualified life. FRC CONCLUSION: This equipment is assigned to NRC Category VI because it is believed by the Licensee to be located in a nonharsh area for the accident condition it is intended to mitigate. The review of this equipment is deferred until after February 1, 1981, as discussed in Section 2.2.3. At that time, the Licensee should provide the references necessary to justify qualification of the equipment, especially details of radiation analysis. Also, the Licensee should furnish a statement on qualified life in accordance with Secticn 4.1.3. i i f 4-89 L.; Franklin Researen Center a w w % r- mew

DELETED MATERtAL IS PROPMETARY WePORMA7m TER-C5257-195 4.7.9 Equipment Item No. 41 Level Switches Located in the Reactor Building Magnetrol Model SIM3 Group 4 Scram Discharge Valve Level Switches (RD-08-A through RD-08-F) (Licensee reference not cited) ORIGINAL TEXT TAKEN FROM DRAIT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITD1 AEDED IN REFERENCE 1): These switches provide for alarm, rod block, and a reactor scram on a sensed high water level in the instrument Eefam discharge volume. These components are located in an area that does not see a harsh temperature and pressure environment. Also, the switches do not provide a primary safety function in the event of HELB inside or outside containment. They do serve to back up the signal that provides the reactor scram (high drywell pressure or low water level). The only possible adverse effect that the failure of this switch might create is to allow a scram reset with a significant level of water in the instrument volume. This would require a deliberate action by the Control Room operator ir. violation of station emergency procedures. FRC EVALUATION: The area where the safety-related level switches are located is mild (1]; the review of the substantiating qualification is therefore deferred until after February 1, 1981. The Licensee's maintenance records should be reviewed to determine if abnormal difficulties have been experienced with the switch. Because the level switch provides an important function, qualification documentation is necessary. In addition, a statement concerning qualified life should be provided by the Licensee. FRC CONCLUSION: This level switch is assigned to NRC Category VI because it is believed by the Licensee to be located in a nonharsh area. Its review is therefore deferred until after February 1, 1981, as discussed in Section 2.2.3. 4-90

    .b2 Franklin Research Center aweNn           m.

DELATIO MATERIAL tS PROMetTARY leePOAMAT1086 TER-C5257-195 4.7.10 Equipment Item No. 29 Electric Motors Incated in the Reactor Building General Electric Model 5K828848C7 Core Spray Pumps (NZ-01-A through -D) (Final Licensee Reference 2.14) ORIGINAL TEXT TAKEN FROM DRAFF INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1) :

     " Qualified" FRC EVALUATION:

Th Licensee has referenced a General Electric test report (2.14], but the report was not made available for review. The Licensee has stated that the report will be made available to provide evidence of qualification for these pump motors. It is noted that the core spray pump motor is not expected to be exposed to harsh conditions, except for radiation exposure, because the accident temperature is expected to be less than 100*F, the pressure less than 1 psig, and the radiation-integrated exposure less than 0.56 Mrads. For this reason, this motor's qualification review will be deferred until af ter February 1, 1981. However, maintenance and analysis records should be reviewed to determine if equipment degradation has been abnormal and to assist in the determination of the equipment's qualified life. FRC CONCL'USION: This core spray pump motor is assigned to NRC Category VI because the Licensee believes it is located in a nonharsh area. Its review is therefore l deferred until after February 1, 1981, as discussed in Section 2.2.3.

                                                                               )

f a-91

    .. .bnklin Reseatell Center aw w th. r. w -.

DELETED MATERIAL IS PROPMETARY WFORMATION TER-C3257-195 4.7.11 Equipment Item No. 13 Pressure Switches Located in the Reactor Building Meletron Model 372 MSL Iow Pressure Switches (RE 23-A through -D) (Final Licensee Reference 2.7) ORIGINAL TEXT TAKEN FROM DRAFT INTERIM TECHNICAL EVALUATION REPORT: None LICENSEE RESPONSE (EQUIPMENT ITEM ADDED IN REFERENCE 1): Peak temperature and pressure seen by these switches are 218'F and 23 psia, respectively, following a reactor feedwater line break outside containment. The main steam line low pressure switches are provided to monitor a pressure drop in the main steam line due to a MSLB and initiate a closure of the main steam line isolation valve. Therefore, these switches are provided to de+:ect a MSLB and not to detect a break in the feedwater line. Due to tht location (different floor level), these switches are also protecteo from a MSLB. FRC EVALUATION: The Licensee has indicated that these switches are in a mild environment for the accident which' they are intended to mitigate. Note that they should be qualified for the worst-case MSLB environment and the environment was not identified. (The SCEW sheet indicates no change.) The qualification document cited by the Licensee was not made available for review. The I.icensee should estimate qualified life in accordance with Section 4.1.3 and analyze maintenance surveillance records for any abnormal behavior which might limit qualified life. FRC CONCLUSION: This switch is assigned to NRC Category VI because the Licensee believes it is in a nenharsh area. Its review is therefore deferred until after February 1, 1981, as discussed in Section 2.2.3. p 4-92

4. Frankhn

        .           Research Center A Chuen,en af The Feensen wisesme

DSL87EO MATERIAL IS PROMWETARY lhPORMATION , TER-C5257-195 4.8

SUMMARY

OF THE EVALUATION The following tabulations represent a summary of the results of the equipment environmental qualification evaluation conducted by FRC in accordance with the methodology presented in Section 3. Table 4-1 summarizes the number of equipment items assigned to each NRC qualification category as a result of the evaluation. Table 4-2 consists of Equipment Environmental Qualification Summary Forms for each equipment item identifying complf ance with the qualification requirements defined in Section 3. The following designations are used: X = A deficiency with respect to compliance with a Guidelines requirement. Deficiencies result in equipment items being categorized as unqualified or qualification not established. L = A limiting

• actor with respect to qualification in that qualified life and aging have not been properly considered by the Licensee.

O = Assignment to an NRC qualification category. R = Replacement of the equipment is planned by the Licensee.

        #e                                          4-93
         ...J gigsesren Center  ,

DELETED WATERsAL is rWTARY lNPORWM cN TER-C5257-195 Table 4-1 NUMBER OF EQUIPMENT ITDfS IN EACH QUALIFICATION CATEGORY NRC Number of Category No. Category Definitions Equipment Items I.a Equipment Fully Satisfies 0

• All Applicable Requirements for the Life of the Plant I.b Equipment Does Not Meet All 0 Applicable Requirements:

McVever, Deviations Are Judged Acceptable for the Life of the Plant II.a Equipment Satisfies All 8 Applicable Requirements With the Exception of Qualified Life II.b Equipment Satisfies All Applicable 0 Requirements With the Exception of Qualified Life Provided That Specific Modifications Are Made II.c Equipment Does Not Meet All 6 Applicable Requirements: However, Deviations Are Judged Acceptable With the Exception o'! Qualified Life III Equipment is Exempt from 1 - Qualification Requirements IV.a Equipment Has Qualification 0 Testing Scheduled IV.b Equipment Has High Likelihood 26 of Operability; However, Proper Qualification Documentation Has Not Been Made Available for Review V Equipment is Unqualified 27 VI Equipment Qualification is 11 Deferred 73 4 4-94

      .. . Frankhn Researen Center
            %

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i I l ceLETeo MATEmeAL as pRoMBETARY popoRMATiope I i T2R-C5257-195 i l l l l Table 4-2 PRCTAsn muasTan punygang I Ud, 'J l'renkiin Research Center

                                                                                                   "                   ewn        orsita carx            Y l                                                     D d.="*. M-N                               M.                             Jersey c T.e Power &

ScuapwSMT ENvipcMmENTAL Light Comoeay OU# DOCKET NAc TAC OAit/EPeo4 MESA sap ptAnfs 50 219 42514 ,.,3.g, gd, i SUMMA 87 mEV'8W SoulmWeNTITEM MuMeSM i l a lamelAAiAsieci s I e 17 leAiesiec:eojui e isol es tiamias ouiceums peouine.vaurs, iossonAnone: x _oercancy L ummnoconomem I i l I I lt Iil i i I evioenes on ouAumcAnon XI Lit LutiL:XIX XIXIX IX LXIXIXIXIX XIXIX asuncasain ro ristseeemeu  ! l l l l IXl IliiI II i l i i Aoino ceanAoAncu evAtuarto ILILILlLILI I I ii l i i i i l l l l

                                       ,cuAcmec upa sstaeusage                            i   ILILlLILILI I i             l l l l l 1 I I              I  l famosa4u reicennsv Aomo                            I   lLiLiLILitl l I I I I i i i l I I                        I  I
                                       !cuac aca sttAu txposume                           i   I I       I I il i i l i I I I I I I                    I   I i                                       intAx reuptaAruas AcacuArt                         t   i I        I l i I i i i ! I I I i i i                   i  i Inux possauma AcacuArs                             i   l I i l I i l i l i i i i l i i !                        I  i i                                        TusicumAnc= Acacuaru                              I   I i i i i l           Iil i I I I I i 1                 1 1

! ReculatD **cmLI ENvfLcPSD l l l l i i l I il i l l l ! I I I i cuAt acm sueuspoaNCE I I I I I I I I i i i i i i l I i i l I l cuaL PCM ChtweCAL s**AY l I i l l l 1 l l l l l l l l l 1 l l l l !cuAL aca sAotAncN l l lLill (Ll I l l l 1 l I I I I l l 1 l !agrA aAo Atew eensrcemes I l l I I I I I l l I i t I l I I I I I I rastsacuenca l l l 1 1 I i l I I I l l I l l I l I i resf ecarrem n wum - aumenem i I I I 1 1 I I I I I I I I l ! ! ! I I cuauniv op acuipusNr i i i i l i I I I I i I I I I I I I I scuipweNirmspec se4rsrts  ! I I I i 1 1 I l l 1 l l l t 1 I I I cuAumcarien caraccay. <o-carnocnv castoNAncm IIA. cuAL SCR suNrCPE l l l l l I l l 1 l l l l l l l ! I t j f

                                        !s.a.cuAt av;ucotweNT                             I l l I I I I i l l 1 f I l l ! ! ! l ;

hs.A.cuat sem < suurtrs i 101 l lO1 I I I I I i i t 't i t i i I +s.cuAL*tNemowceircarem 1 I I I I I l ! l l l l l f ( I t i I ( ii4. cu AL < m. ANT UP(/ PSC Review I i 10101 101 I I l 1 l l l l i i i t l l ha. txswer secu cu AL l t i i i ! i i i i l i I i t i i i l inv.A.cuAL issiscutouta l i I l l I i l ! I I l t i i l i i i i I 'iv.s. cuaL %cr asraeuswec 101 I I i i 101 1 I I I I I I I i los I . i v sc uis scT va u mso I I I i i i t I (01010101010:03 101 ! 'l

                                        !vi.cuat.s cartmoto                               I 1 i t i I i 101 I I I I e i IO t I !OiO
 ,                                      'at=ucsvasi scatcuLa                              i*: : : I e IR! t***!*i*i*i*t*i*:***i                             !

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                                       #                                                           4-95 I
                                          .d Frankhn ao     a e Re.

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i i , I i l l OELETED WATEm6AL IS PeopasETARY INPORMAT1oM l TER-C5257-195 4 i

!                                                                                                                                                                                      i
,                                                                            Table 4-2 (Cont.)                                                                                         l

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occx T amerAc cArnisessa l seppuurs so-ra 6:526 ,. . ,, . . , ra_

SUMMARY

pfvttw SoufPMENTITEM NUMSER lourceuws peouteswasTs. t2CIE oll 3 llessetl 151 f 6 l 17 I IS I 19120242877238246251241273 ossoaAncas. x -canoamcy.L-ummmo conomcm I i i l i i l I i i i I l i i i i i i i 1 2 levicencs os ouaumcAnow i Xi XIXl XIX;X IX IX IX IXIX IXI XIXI XIX IX IX IXIX lesunonswee te resisesowas l l Ii i i 'l I l l l l l l 1 l I I I IX laciMo CGaAAP*', sW EVALu Arne 4 i i l i I i ! I I I I IXIXIXIXIXIXiXIXIX i I cuaumaouatesTasus so i I I I I I I i i IXIXIXIXIXIXIXIXIXIXIX leaesaAu reneuriev acino I i i i i i  ! I I IXIXIXIXIXIXIXIXIXIXIX , 'cuAL 8op s!DM gxecsues i I I I I I i i i I i i IXl i I I I I IX psA(*IMPREA?upt aQEOUATE I I l I I I I I I I*I I IXl l I ! I I I leuxpaassuasAcacuArs l I I i i l I I i i l i IXl i I I I I i

! jrest umanc= A:s:uArt                                                  I     i     i     l     I     i     i l i l I i IXl I i i l i l i  lascutaso sacats tuvate*t:                                             I     I     i     1     I     I     I I I I i l I XI I I I i i I icuac aca sueuseoswer                                                  i     i     i     i     i     !     I ' I I I I I I i i ! I i i cuaL Pom entwicAt s**.Av                                             i     i     i     l     I     i     l l I I I I I I I I i i i 1                                             -

cumu rca mAmanc= I I I I I I 1 1 IXIXIXIXIXIXI I IXIXI issta nAoAnca censicanas 1 I I i i i i l I I i i i i i l i i 8 l

  %sise:uswes                                                            !     I     I     i     !     ?     I I I I I I I I I I I I I I_

asteuaAnca n -oum aumenem i I I l l1I I I i l I i l l I I I i i  ! cumuTnv oa s:uiewsw? I i i i i i i i l l l 6 I I i i i i i l f tcusewswTimsasersc arstre i l I i l i l l I I i l i I i i i i i i louauncanon cArnocav. to - cATusomy :sse nAnc m Ita. cuat ace ptAm? ust i I i i i l I i i i i i l l I I i l i l 16a. VAL evaucosuew? I ! I i l I I i i i l i i l i I I I I I

  !s-a. cuat som < *uNTurt                                               i I I I I I I I I I I I l I 8 I I l l 3                                                                       ,

i In-4 CdaL **JeDNo WCC'8'cATiCN I I I I I 1 I I I I I I I I I I I I I I ' in.c. OuaL <

• tant uSEf rac ogvtgw I i l i l I I I I l l i l l l ! I I l i lin. txtupT racw ouaL i i l i i 1 1 I i i l i I I I l l i ! I IN-A.cua: rest sc> scuts l I I I I I I I i i i i i i i! 1 l I is-a ctat uctsstaauswen i I I i I i I i I 6010101 10 1 1 1 1010i0 Iv scuss %ctcvat ::: 101Ol 101 l01 1 101 I I 601 to10lcI t I tvi. cuat is carsames 1 1 101 101 1010' I i l i l I i i i l i t *EstACEVENT sCME0ut! IMfkl 1414 t* ** e n f M liti M 4
• tid i M f
• t i IMIMiM

   #2- ::                                                                                       4-96 J.2 Franidin Research Center a:>          .# N a            a -

1

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

oELETED W ATERIAL IS Pm0PMIETARY lNPORMATION TER-C5257-195 i i Table 4-2 (Cont.) FacTasn me;g;c PLANT NAME

             ,&,p, Frenidin Research Center
                                                          #                   sw7        oystem caggx            3 "Df ".TI.". ,'.*."1"'Tll"           * % set                                    unuTY Jersey Central Power &

323 cs25701 SQuaPMtNT &MViACNMSMTAL Light cuAumCAnON cocKg7 OA7xjgggiggxg " Nnc 7AC sap PLANT 3 $0-219 42526 y.,or.,, pg

SUMMARY

RaviEW goulpMENTITEW NuVS$R 29l 3013643553353333 l34434430C 35134 j 37l 344 39440144 l 42t 43 } 44 ouiosuME RGculREMENTS. (CEslGNAnCNs; X =QEMC ENCY.L=uMmNGCcNoCCN) I i i l l l l l l l l l l 1 i i I

      .svioenes ca cuAumcAneN                 IXIXtL XILIXIXil LIXIXlXIXIXl                          X IX X XIX lasLArteNsNie to rtsr spectueN          lil               i l i Ill IXl i I l l XI I X  i Ao No esonAcAnCN EVALu Afto           I ! Ill lll I lLill l                         l I l I l I                ,

cuAurec urs estasusNeo I I ILI ILI I LitlXIX XIXl i IXl i I IX l

      !amorsnAM roiceNnPv 4GINo               !i Itt iti I LlLIXIXIXiXI I IXl i i IX touAt aca stueu ppesues                 iI I i i I I i i IXl I I I                       I IXl      I I IX pontsunt= Arums AcacuArs               1 I I I I I I I I I I I i i                     l i I      I I I sca peessuns ecacuata                 l i I I I i i i l I i i l I                     I I I      i l I resf eu=AncN AescuArt                  i I I i l I i ! I i l I i l                     I I I           i I meoutaso aacmts tNvstesso              I i I i i t i i l i I I I I                      I I 1          I i lcuat men susMemosNcs                   I i i l i I I I I I I I l i                     I I I      I  i 1 leuAt aca cNewcAi.spaar                 I I i IXl IXl l I I I I I I                      I I I     I  IXl leuAL PCR MACIAnCN                      I I I I I I I I ILlXIXIXIXl                      I I i      I  I IX issTA maclAncNc Nsicameo                l 1 I i i l I I I I I I I I                      I I I     I  I I Tasisscuencs                           l I I I I I I l l l l 1 I l                      l 1 1     I  I I resr eumaneN n %um . nuNenem          i I I IXl IXl I I i i i IXl                      I l t     i  IXl
      !cuaNniv er seuipuaNr                   i I I I I i l I t i I i i                         i I I     i  i I lacuiaMeNresspec as Arsits              1 I i l I i l 4 (*l l l 1 I i i i                          I  i l        !

cuAUFCAnCN cATsocRY. (0 -CATtGC AY OE31CNAT;CN) lA. cu AL PCR PLAN? uPW I  ! l l l l 1 I I I I l l l l l l l l l i

      !i a. cuat ev ;uccaveNr                 i    l     i l   I i  l    i l i    l i  I I I I        i I i I          l
      !il-A. cu At aca < nuNruFs              1    I     101   IOi  i    101 I   i i   i i l I            1 i i lu cuAt asNeiNo wccimeAr'cN             I    I     I I   i l  i   I l I    l'I    I I I i       l  i   l  i lac. :uat <

• TANT ureisac aevtew i i l I i l i l 608 I i i l i i i  !  ! l lm. csuev ascu cu2L i i 1 l 1 i i l i i i ! I i 101 I t i I f rv.A. cuAL east scNs:uts i i  ! I I I I l l l l l l l l l l l l l

       !rv-a. :uAL =cits?Ases so              I i l 101 10101 1 101010 101 ! 101 101010 iv ecuto NcreuAer'es                   i 101 1 I I i i i l i l I I I i i 1 I I
       'vi. 0uat is egesase:                  101 1 ! l 1 i i l l 1 l l to; l loi j ;

I i i l i l I IRl* +i  !-M i  :+i i ;4 l* Ent.AcEMENi scNE0ut!  ! I

      #e                                                      4-97 i.J Franen Researen Center ao aw N neana -

DELETED MATERIAL tS PROPRIETARY lNFORMATION TER-C5257-195 Table 4-2 (Cont.) FRC TASK W PLANT NAME

                                                      .EIIS UM Franklin Research Center                                    3           Ot3TER CREEK a ow The F'=* 'a===               omosE EQutPMENT ENVIRCNMENTA6 dm7e                                       utruTV Jersey Central Power &

tithe Cu AuPIC & TION DOCKET NRC TAC OATE/ ENGINEER SEPPLANTs 50-219 42524 dau.se A

SUMMARY

mEviEW EQulPMENTITEM NuMSER 4 S 1461471481491501s l l 52153154155156 l l l l } l l l GulOEuNE REculmEMENTs, (des 3GNATIONS: X - oEFICIENCv. L = uMONG CONOmCN) l l l l l l tI lIi l i i !l I i 1 l EvtoENet or cu AuricAnCN IX XIXIXlLILIXlLILILIXIXl l l ! l l 1 3 REuncussis to TEsisPECIMEN IX llIil l l IXl l l 1I III 2GiNG oEGRADAnON EVALuATEo IX l i IXl i IXI I IXlXl l i I I i l IcuauriEcusEEstasus Ec I X!X IXIXl L iL IX lt IL lt IXlX l i i I l 1 I leacGnAu Te ioENnav AGING IX IXIXIXit'LIXIL !LILiXIXl l I I I I I i

     !cuAt aom STEAM Execsuat                i I I I Xl i I I I              I i l i i i i            l  I i PEAK TEMPERATURE ACEOu ATE             I l l i l I l l i               I I I l I l I            i  l i PEAK *mEssumE ACE 0uATE                l 1 l l l l l l l               l l l l l l l            l  l   8 l TEST umAncN AoEcuATE                  i i li i l l l l l l l l l l l l                        1  1  I RECulmED **OFILE ENVELOPED             1 1 1 l l l i l l I i ) i l I i l                        l  l 1 CuAL PCa suSMERGENCE                   I i il l l l l l ! l l l l 1 i i                        I  i  !

leurt rem eNEMicAt sam 4v IXl I IX l i i 1 I I I I I I I I I I i i leuALeom=4cianeN I IXIXIXl i IXl I I IXIXl l I I I I l i

     !scA =4ctAnoN CONstOERED                l l l l l l l              l l l l l 1 l l l             l  1 !

iTEs sEcufNcE 1 I I I l l l l l l 1 l l l l l 1 1 I TEST eumAnoN n woun . ruNencNi l l I l l I i l l I i i l l l l l l l cuANTITY CF ECulPMENT l l l 1 l I l l 1 l I l I l ! l l l l l IECutPVENTINsPEO*ED ATs!TE l l 1 l l l l 1 l l l l l l l l l l l l lcuauricAnou c.TEGomy. to -cATEGemy DEsMINAnCN) liA.cuAL FCm DLANTuRE I l 1 l l l l l l l l l j l l l l l l l h-e. cu4L av aucGEMENT I I i I I i i I i. I I I i. I I I l i i l in A.cuAL rom < *uN urE I l 1 i 10601 ! IO ! ! I I I I i 1 l i I

     !n-a. cu AL *ENCING MOcimeAnCN          I l l l l II I l !I l I ! l I I I I !

l in c. cu AL < PLANTuFErrRC REVIEW i I I I l l i 10 1 10 1 1 1 I i l I i l !

     !m. EX EMPT *mCM Cu AL                  l I l l l ! I l I l i l I l l l 1 i i l liv-A. Qu AL1 EsisCME3uLE               l l l          l l 1 l l l l l } l l l l l l l liv.a. euAL Not EsTAsuswEc              I 1010 l I I Ol I i 101 i i l i i i l i Iv. Ecuio Nercu4u=E:                    101 i !Ol i l i I i i 101 ! I I I I i l Ivt 0cALis OEFEmpE:                     l l l          1 l l l l l l l l l l l l l l j imEPLAcEMENT scut:ULE                   t u t

• l *t i i I I I 6 l M Ht-i i 6 i i i e i g_3 4-98 OU Franklin Resear:h Center 4 Ome.on of The Franses enseewe

DELETED MATERIAL IS PROMeETAM MFORMATION TER-C5257-195

5. CONCLUSIONS The tabulations preser.ted in Section 4.8 represent a summary of the results of the equipment environmental qualification (EEQ) assessment conducted by FRC in accordance with the methodology prese .ted in Section 3.

The evaluations are based ,.n the available qualification documentation provided by the Licensee, supplemented in several cases by other relevant technical information. The major deficiencies that have been identified are shown in the Equipment Environmental Qualification Summary Forms (Table 4-2) . The review has shown that qualification documentation for many equipment items is inadequate a'r non-exiatent, and that additional information is essential. The DOR GJidelines require the Licensee to have ongoing programs to review surveillance and maintenance records in order to assure that safety-related equipment that exhibits age-related d'egradation be identified and, if necessAry, replaced. No evidence of such programs was included in the Licensee submittal. The Licensee has offered several system-related arguments to exempt certain equipment items from qualification review. Most of these arguments i fall into two categories: (1) the backup system redundancy can adequately accomplish the function, or (2) the equipment need only survive for a few minutes in order to accomplish its intended function. The FRC conclusions regarding these arguments are given in Section 4 for each equipment item, and a more detailed analysis is presented in Appendix D. The present assessment of Ete status of environmental qualification of the safety-related electrical equipment installed in Oyster Creek involves only equipment located in the " harsh environment" areas and needed to ensure hot shutdown of the plant. The EEQ review of equipment items located in

              " mild" areas and of equipment needed for TMI Action Plan compliance has been deferred by the Licensee until af ter Febr'tary 1,1981.
                   # N-                                                                *-1
                    ...; Franklin Research Ce ster
                          <o    n a w..n.n m. .w

I DELETED MATERIAL 18 PRCPImETARY INPORMATION TER-C5257-195

6. REFERENCES I

i

1. I.R. Finfrock, Jr. (JCP&L)

Letter to D.M. Crutchfielc' (NBC),

Subject:

Environmental l l Qualification of Safety-Relsted Electrical Equipment, with Attachments Jersey Central Power & Light, 28-Oct-80 1 2. Qualfication Documentation References identified in Reference 1: 2.1 Project Engineering Test Report: Dresser Relief Valve Actuator General Electric PEP 42963, Proprietary 2.2 J.B. Drab (Limitorque) Letter to R. Pruthi (GPU Service Corp.),

Subject:

Qualifica-tion Information for Oyst.ar Creek Nuclear Generating Station Limitorque Corp., 21-Aug-80 2.3 S.P. Carfagno, L.E. Witcher, and W.H. Steigelmann Qualification of Limitorque Valve Actuator in a Steam Environment FIRL, 00-Feb-72 Report No. F-C3271, Proprietary 2.4 T. Hess, Jr. Qualification Type Test Report: Limitorque Valve Actuators for Class lE Service Outside Primary l Containment in Nuclear Power Station Service Limitorque Corp., 28-May-75 Report No. B0003 2.5 J.B. Drab (Limitorque) Letter to R. Pruthi (GPU Service Corp.),

Subject:

Qualification Information for Oyster Creek Nuclear Generating Station Limitorque Corp. ,17-Nov-80 2.6 Letter Gilbert Associates, Inc., 04-Jun-80 . 2.7 S tudy Wyle Labs, Proprietary 2.8 IE Bulletin 79-01B USNPC , 14-Jan-80 l

        -                                       6-1
   ...d Franklin Research Center ac   a w m r,  w.

DELETED MATERIAL.13 PROPRIETA3 tNFORMATION TER-C5257-195 2.9 Reports Cable Splice Assemblies, with EWR Owners Group Summary Report 2. QSP-084-H-01 and 02 for Vulkene Cable Hyle Labs,15-Sep-79 44144-2, Proprietary 2.10 Document 9999.1217.2 and Report R3-288A-1 ITT Barton, Proprietary 2.11 Radiation Ef fects Handbook IEEE Nucleonics Committee , 01-Jun-63 2.12 Test Report 5628-3509 with Letters Dated Aug. 6,1980 and Sept. 15, 1980 Yarway, Proprietary 2.13 Study of the Effects of mclear Radiation on the Mechanical Properties of Acetol, Resins, Delsin and Celcon (USAF Nclear Aerospace Research Facility) USAF, 31-Mar-64 2.14 Intter and Test Report General Electric,16-Oct-80 G-EN-O-164 - 2.15 E.J. D' Aquanno (Ibckbestos Co. ) Letter to R.J. Pruthi (GPU Service Corp.),

Subject:

Qualifica-tion Reports on Firewall EP Cable and Firewall III Cable Rockbestos Co. , 24-Oct-80 i 2.16 Tests Conducted at Oyster Creek NGS, Ef fects of Chromate  ! Solutions on Come Elastomers and Metals i 2.17 R.h. Schuster Report Qualification 7sst for F01 Electrical Pertetration Assembly General Electric, 30-Apr-71, Proprietary 2.18 Summary Data, Section 1.5 General Electric, 20-May-72, Proprietary 1 1 2.19 Test Report: Shielded Signal Thermocycling Test 21-Apr-68 EPAQ-Oll, Proprietary

2. 20 R.M. Schuster Report: Terminal Block LOCA Test for Electrical Penetration Assemblies Genera 1L Electric, 06-Nov-73, Proprietary TJh Franklin Research Center 6-2

           % e n. wa www.

l

l l l l OELETED MATERIAL IS PMoPRIETARY INFORMAT1oM 1 TER-C5257-195  ; 2.21 L.E. Witcher and D.V.Paulson Technical Emport Qualification Tests of Electric Cables Under Simulated Reactor Containment Service conditions Including Ioss-of-Coolant Accident FIRL, 00-Mar-77 Report No. F-C4497-2, Proprietary 2.22 D. Chalk (Tensolite Co.) ta tte r to R. K. Pruthi (General Public Utilities),

Subject:

Transmittal of Two Reports Regarding Insulation with h fzel 200 ano 280 Tensolite Co., 24-Oct-80 2.23 S.P. Carfagno, L.E. Witcher, and W.H. Steigelmann Technical Report: Qualification hsts of Electrical Cables Under Simulated Post-Accident Reactor Containment Service Conditions FIRL, 00-Oct-70 Report No. F-C2770, Proprietary 2.24 Intter ,

Subject:

ASCO Test Beport, AQS 216781 TR, Rev. A; #SCO Catalog No. NP-L (Previously 2.1 & 2.3) AS CO, 26-Sep-8 0 Proprietary - 2.25 Intter and hst Report General Electric,10-Cct-80 G-EN-O-163

3. I. R. Finfrock (JCP&L)

Iatter to D.M. Crutchfield (NRC),

Subject:

Environmental Qualification of Electrical Equipment; and Supply Info in Intters cated 4/11/80 and 5/7/80, and Meeting of 10/9/80 Jersey Central P&L,10-Dec-78 l 4. Report: Environmental Effects on Safety Grade Electrical Equipment Due to LOCA and High Energy Pipe Rupture, Prepared for JCP&L EDS Nuclear, Inc., 01-Apr-80 Report te. 02-0370-1045 l S. Latter to NRC,

Subject:

Responses to NRC Request for Addtl. Information, SEP Topic III-5.B, Pipe Break Outside Containment, Oyster Creek Jersey Central P&L, 03-Oct-80 bi.2 Franklin Research Center 6-3 r on an a. v, aman +mue

DELETED MATERIAL l$ PROPRIETARY INFORMATioN TER-C5257-195

6. G. Lainas (NRC)

Iatter to A. Schwencer (NRC) ,

Subject:

Electrical Muipment Environmental Qualification, with Attachments Containing DOR Guidelines CS NRC , 19-Fe b-80

7. Draft Interim hchnical Evaluation Report on Equipment Environmental Qualification for Oyster Creek Nuclear Generating Station FRC, 23-Oct-80

8. N.C. Moseley (NRC)

Letter to B.H. Grier (hFC),

Subject:

IE Supplement No. 2 to Bulletin 79-OlB, Environmental Qualification of Class lE Muipment NRC, 29-Sep-80

9. N.C. Moseley (NRC)

I.etter to B.H. Grier (NRC) et al.,

Subject:

Supplement No. 3 to Bulletin 79-OlB, Environmental Qualification of Class lE Equipment USNRC, 24-Oct-80

10. S.J. Chilk (NRC) -

Memorancum and Order Pursuant to Union of Concerned Scientists Petition for Dnergency and Remedial Relief USNRC, 23-May-80 CLI-80-21

11. J. Archer (IRC)

Memo of 'Islephone Conversation with S. Brown (NRC),

Subject:

Oyster Creek Nuclear Plant Drywell Containment

                   'Iemperature/ Pressure Profiles FRC, 19-Oct-80

12. S. Brown (hRC)

Memo to D. Crutchfield (NRC),

Subject:

Mark I Iong Term Temperature Transient for Environmental Qualification USNRC, 28-Mar-80

13. S.P. Carfagno and R.J. Gibson A Review of Muipment Aging 'Dieory and Technology Electric Power Res. Inst. , 00-Sep-80 hT-1558 O 6-4 d$5nklin Research Center a om a.t w rr.n a .

I l DEL ~TED MATERIAL ts PROPRIETARY INFORMAT10N TER-CS257-195 l APPENDIX A - ENVIRONMENTAL SERVICE CONDITIONS The Licensee provided information concerning " harsh" environmental service conditions in various locations of the plant where safety-related equipment is installed: the containment drywell, the reactor building, and the steam tunnel. The EEQ review of equipment needed to achieve cold shutdown status has been deferred in accordance with Section 2.2.5. In addition, the Licensee has deferred the EEQ review of equipment located in " mild areas," as discussed in Section 2.2.3. Therefore, only the " harsh" environments were discussed in Referetice 1, and considered in this report. In Table 1 of Reference 1, the Iicensee presented the worst-case temperature / pressure / radiation service conditions that each safety-related equipment item located outside of the containment drywell could experience for different types of postulated accidents. According 'to the Licensee, the maximum duration of the pressure / temperature excursion is 1200 seconds before

                                                                             ~

conditions return to normal. Figures A-1, A-2, and A-3 define the results of the containment drywell MSLB analysis, showing the expected temperature and pressure excursions af ter the worst-case postulated accident. Environment 1 - Within Reactor Containment Drvwell Normal Oceration . Temperature 70*-135'? (120*F nominal) Pressure 15.7 psia Humidity 60% (nominal) Radiation (Not stated, included in accident dose) l 1 A-1 4jhij Franklin Research Center a em or tw rw mesem

DELETED MATERIAL O T ROPRIETA%Y lNFORM AhoN 1 TER-CS257-195 Accident Conditions  ! For BWR plants, Section 4.1 of the DOR Guidelines states that the , temperature component of the environmental service conditions within the j I drywell for the loss-of-coolant accident (LOCA) will be 340*F for 6 hours. This exposure is intended as a bounding condition to reflect the superheated steam release associated with the most severe main steam line break (MSLB) accident. Supplement 2 to IE Bulletin 79-OlB states that a plant-specific analysis may be used in lieu of 340*F for 6 hours. The Licensee has provided plant-specific analyses for both LOCA and MSIB events. The latter is more severe than the former, and therefore is the basis for establishing the status of qualification. The Licensee has investigated a wide spectrum of postulated break sizes, break locations, and single failures associated with a LOCA or MSLB accident. The NRC has acknowledged that the MSLB accident appears to be the limiting environmental service condition for which equipment located within the containment drywell is to be evaluated. The environmental parameters associated with the LOCA and MSLB events used for the assessment of qualification of equipment inside the containment are: Temperature Figures A-1, A-3 (Re f. 1) Pressure Figure A-2 (Ref. 1) Radiation 57 Mrd (1 year, gamma radiation only; includes 40-year normal operation) Humidity 100% (assumed) Flood Level Not stated Spray Demineralized water containing sodium dichromate Environment 2 -- Within Reactor Building; Outside of Containment Drvwell Normal Oceration - Temperature Not stated Pressure Not stated Radiation Not stated Humidity Not stated _nklin Rese_ arch ._ Center

DELETED MATERIAL IS PRCPfWETARY INFORMATION TER-C5257-195 Accident Conditions Temperature See Table 1 Pressure See Table 1 Radiation See Table l' Humidity 100% (Assumed) Flood Level Elevation Deoth (Re f. 1)

                                                    -19*6"             2 1/4" 23'6"             2 7/8" 51'3"             2 7/16" 75'3"             1 3/8" 95'3"               9/16"

_Er vironment 3 -- Steam Tunnel Normal Ooeration Temperature Not stated Pressure Not stated e Radiation Not stated Humidity Not stated f Accident Conditions Temperature See Table 1 Pressure See Table 1 Radiation See Table 1 Humidity 100% (Assumed)

Flood Level 0, Water will drain to Turbine Eldg.

l Sump Pump at -8' elev. A-3 Jdb Franklin Research Center a w.anan N rr a m

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ugua3 yanasay uimueJd '~"" t-V WW IWI; SnSE3t. 3ErJ.W3dh2; SNCIIIONCD 3DIAH2S 'IY;NIWNCEIAN3 2 i DRYWELL VAP. TEMP.F 3 134 174 214 254 294 33*

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           .                                   DRYWELL LIO. TEMP.F t

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                                           -           i-           M         0000    12000     Im00      20000  riOOO      20000   32000 M ELOPSED IIME.SEC                                Z Figure A-2. Pressure Versus Time for Postulated Main Steam Line Dreak in u e nryweil til FIGURE SUPPLIED
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1 l l l DeLaTED MATERIAL. Is PROPmETARY INFORMADON l TER-CS257-195 i l APPENDIX B - LISTING OF SAFETY-RELATED ELECTRICAL EQUIPMENT l The following table lists the groupings of safety-related electrical equipment items for the Oyster Creek Nuclear Generating Station. Equipment item numbers provided in the table are used in the Equipment Environmental Qualification Summary Forms and in the equipment qualification discussions presented in Section 4. This table was generated from the lists of equipment items provided by , the Licensee in Reference 1. FRC has listed plant equipment items by I I manufacturer and model number, plant location, and time required to function l as identified by the Licensee. l 4 B-1

                  .Juda Franidin w w % r   Resear.ch am u.          Center

                                                                                                                                )

i DELETEO t"ATERIAL O PROPRIETA%Y INFORM ATION TER-CS257-195  ; EQUIPMENT ITEM ITEM QUALIFICATION TIME NO. DESCRIPTION LOCATION SCEWS NO. REFERENCES REQUIRED 1 Pressure Switches Reactor 1-5 None Intermediate Dresser Building (* 3 h) 1593 VX (Automatic Depres-surization) 2 Solenoid Valves Reactor 6, 7 2.24 Long ASCO Building (% 30 days) NP-8344A70E 3A Motorized Valve Reactor 8, 9 2.2, 2.3, Iong Actuators Building 2.4, 2.5 (30 days) Limitorque SMB-00 (Containment Spray) 3B Motorized Valve Reactor 14-17 2.2, 2.3, Intermediate Actuators Building 2.4, 2.5 (4 h) Limitorque SMB-00

 .              (Containment Spray) 4A      Motorized Valve                       Reactor      10, 11                2.2, 2.3,                Long Actuators                          Building                           2.4, 2.5                  (30 days)

Limitorque SMB-000 (Drywell Isolation) 4B Motorized valve Reactor 12, 13 2.2, 2.3, Intermediate Actuators Building 2.4, 2.5 (4 h) Limitorque SMB-000 (Containment Spray) 4C Motorized Valve Drywell I-2A None Short Actuators (2 min.) Limitorque SMB-000 (MSIV) 5 Pressure Switch Reactor 18-21 2.6, 2.7 Iong Static-O-Ring Building (30 days) 12NKA (Drywell Pressure Scram)

          &b Franklin Research Center                            B-2 A Dmason af The Fw inumewee

. ~ , , . . _ _ . - . _ _- __ _, _ _ . . .. - . - - _ _ , _ _

DELETED MATERIAL IS PROPRIETARY INFORMATION TER-C5257-195 EQUIPMENT ITEM ITEM QUALIFICATION TIME NO. DESCRIPTION LOCATION SCEWS NO. REFERENCES REQUIRED 6 Pressure Trans- Reactor 22, 23 ,None Long mitters Building (30 days) GE/MAC 551 (Reactor Vessel Pressure) , 7 Pressure Trans- Reactor 24, 25 None Long mitters Building (30 days) GE/MAC 551 (Reactor Vessel Pressure) 8A Level Transmitters Reactor 26 29 None Intermediate General Electric Building (4 h) GE/MAC 553 (Isolation Condenser Level) 8B Level Transmitters Reactor 30-33 None Long General Electric Building - (30 days) GE/MAC 553 (Reactor Water Level) 8C Flow Transmitter Re actor 46,49 2.25 Short General Electric Building (4 h) GE/MAC 553 (Containment Spray Flow) l ! 8D Pressure Reactor 54 2.25 Long Transmitter Building (30 days)

General Electric GE/MAC 553 (Drywell Pressure) 8E Pressure Reactor 133-136 2.25 Long Transmitter Building (30 days)

General Electric GE/MAC 553 (Containment Spray Differential) 9 Pressure Switches Reactor 34-37, None Long Mercoid Building 39,41 (30 days) 9-51/ DAW-43-156-R2IE (Core Spray) 4 B-3 db Frankhn Research Center a w w w - u. h

DELETED "iATERIAL 0 PROPRIETA";Y INFORMATION TER-CS257-135 EQUIPME!C ITEM ITEM QUALIFICATION TIME NO. DESCRIPTION LOCATION SCEWS NO. REFEPENCES REQUIRED 10A Pressure Switches Reactor 42, 43 None Iong General Electric Building (30 days) GE/MAC 552 (Core Spray) 10B Pressure Switches Reactor 38, 40 None Long Mercold Building (30 days) 9-51/ DAW-43-156-R21E (Core Spray) 11 Temperature Reactor 44-47 None Short Detectors Building (10 min.) Rochester Instru-ments No Model No. (Isolation Condenser Area Leak Detection) 12A Pressure Switches Reactor 50-53, 2.7,'2.10 Iong Barton Building 55-58 (30 days) 288A (Containment Pressure) 12B Reactor Isolation Reactor 63-70 2.7, 2.10 Long Switches Building (30 days) Barton 288A . (Reactor Isolation) 12C Level Switches Reactor 178-181 None Long Barton Building (30 days) 288A (Reactor Vessel Level) 12D Level Switches Reactor 71-78 lbne Intermediate Barton Building (4 h) 288A (Isolation Condenser Delta P) 13 Pressure Switches Reactor 59-62 2.7 Long Meletron Building (30 days) 372 (MSL Low Pressure) B-4

   %' aFranklin JJ.L w a# ww n. an. Research C. enter
                                        ---._,e.

DELETED MATERIAL t'e #ROPMIETARY INPORMATION TER-C5257-195 EQUIPMENT ITEM ITEM QUALIFICATION TIME NO. DESCRIPTION LOCATION SCEWS NO. REFERENCES REQUIRED 14A Pressure Switches Reactor 79, 80 None Long Barksdale 'uilding (30 days) B2T-A12SS (Core Spray) 14B Pressure Switches Reactor 87-90 None Short Barksdale Building (10 min.) B2T-Al2SS (Reactor Pressure) 15 Pressure Switches Reactor 81, 82 None Long Barksdale Building (30 days) E2T-M12SS (Core Spray) 16 Pressure Switches Reactor 83-86 None Short Barksdale Building (3 h) B2T (Reactor Vessel Pressure) 17 Level Switches Reactor 91, 92 2.11, 2.12 Short Yarway Building (10 min.) 4316E (Reactor Water Level) 18 Level Switches Re actor 93-96, 2.11, 2.12 Long Yarway Building 182, 183 (30 days) C2337 (Reactor Water Level) 19 Solenoid Valves Reactor 97, 98 2.7, 2.13 Long ASCO Building (30 days) 8344-B27 (Drywell Isolation) 20 Solenoid Valves Reactor 99, 100, 2.7, 2.11 Long ASCO Building 10 2 (30 days) 8344-A27 (Drywell Isolation) 21A Solenoid Valves Reactor 101 2.7, 2.11 Long ASCO Building (30 days) 83148 (Drywell Isolation) 9-5 4 NL rank!in Research Center ac e m r - . ann

DELETED MATERIAL is PROPRIETARY lNFORM ATION TER-CS257-195 EQUIPMENT ITEM ITEM QUALIFICATION TIME NO. DESCRIPTION LOCATION SCEWS NO. REFERENCES REQUIRED 21B Solenoid Valves Reactor 176, 177 2.11 Internediate ASCO Building (4 h) 83148 (Isolation Condenser) 22A Solenoid Valves ' Reactor 103, 104, 2.7,2.11 Long ASCO Building (30 days) WP8300B61RU (Drywell Isolation) 22:3 Solenoid Valves keactor 115-119 2.7,2.11 Iong ASCO Building (30 days) WP8300B61RU (Drywell Isolation) 23 Solenoid Valve Reactor 105 2.6,2.7 Long ASCO Building (30 days) WPLB83177 (Drywell Isolation) 24 Solenoid Valve Reactor 106 2.6,2.7 Long ASCO Building (30 days) 831424 (Drywell Isolation) 25 Solenoid Valve Reactor 107 2.6,2.7 Long ASCO Building (30 days) X8031A42 (Drywell Isolation) 26 Solenoid Valves Reactor 109-112 2.6,2.7 Long Atkomatic Building (30 days) 15-702-B (50R) (Drywell Isolation) 27 Solenoid Valves Reactor 113, 114 2.7 Long ASCO Building (30 days) LB82627 (Drywell Isolation) 28 Temperature Switches Main Steam 120-124 2.11 Short Fenwal Tunnel (1 min.) 17002-40 (MSL Leak Detection) g B-6 ddOJ Franklin Research Center 4 Dmmon of The Frenamn megame

DELETED MATERIAL IS PROPRIETARY INFORMATION TER-C5257-195 EQUIPMENT ITEM ITEM QUALIFICATION TIME NO. DESCRIPTION LOCATION SCEWS NO. REFERENCES REQUIRED 29 Electric Motors Reactor 125-128 2.14 Long General Electric Building (30 days) SK-828848C7 (Core Spray Pumps) l  ! 30 Electric Motors Reactor 129-132 2.14 Intermediate  ; General Electric Building (4 h)  ! SK-818642A10 3 .  ; (Containment Spray Pumps) 31A Solenoid Valves Reactor 137-139 2.24 Short ASCO Building (1 min.) 206-832-3RU (MSIV) 31B Solenoid Valves Drywell I-1A 2.16, 2.24 Short ASCO (2 min.) 206-832-3RU (MSIV) l 32A Solenoid valves Reactor 140-141 2.24 Short ASCO Building (1 min.) 206-301-3R (MSIV) 32B Solenoid Valves Drywell I-1B 2.16, 2.24 Short ASCO (2 min.) 206-301-3RU LMSIV) 33 Position Switches Reactor 143-146 2.11 Short Snaplock (NAMCO) Building (1 min.) SL3-C58W (MSIV Position Indication) 34A Motorized Valve Reactor 147-149 232, 2.3, Short Actuators Building 2.4, 2.5 (1 min.) Limitorque SMB-0 (Drywell Isolation) l A_ B-7 j$) Franklin Research Center 4 % orm r,.enw=n

i DELFTED WATE.11AL G PROPRIETARY INFORM ATION TER-C5257-195 EQUIPMENT ITEM ITEM QUALIFICATION TIME NO. DESCRIPTION LOCATION SCEWS NO. REFERENCES REQUIRED 34B Motorized Valve Reactor 151-156 2.2, 2.3, Long Actuators Building 2.4, 2.5 (30 days) Limitorque SMB-0 (Drywell Isolation) 34C Hotorized Valve Drywell I-2B 2.16 Short Actuators (2 min.)

       - Limitorque SMB-0 (Shutdown Cooling) 35       Solenoid valve                          Reactor                    150           2.7, 2.11          Long ASCO                                    Building                                                    (30 days)

LM831424 (Drywell Isolation) 36 Solenoid Valves Reactor 157-160 2.7 Long ASCO Building , (30 days) WP8300B61U (Drywell Isolation) 37 Motorized Valve Reactor 161, 162, 2.2,2.3, Iong Actuators Building 168, 169 2.4,2.5 (30 days) Limitorque SMB-1 (Core Spray) 38 Isolation valve Reactor 163 2.7 Long Switch Building (30 days) Meletron 4201E-3B (Drywell Pressure) 39 Electric Motors Reactor 164-167 2.14 Long General Electric Building (30 days) SK-818841C45 (Core Spray Booster Pump) 40 Motorized Valve Re actor 170-175 2.2, 2.3, tong Actuators Building 2.4, 2.5 (30 dayr) Limitorque SMB-2 (Isolation Condenser) i 4 B-8 OUJ Franklin Research Center A !> aeon of The Frenen sneemme {

DELETED WATERIAL IS PROPRIETARY INFORMATION TER-C5257-195 EQUIPMENT ITEM ITEM QUALIFICATION TIME NO. DESCRIPTION LOCATION SCEWS NO. REFERENCES REQUIRED 41 Level Switch Reactor 184-187 None Short Magnetrol Building (5 min.) SLM3 (Scram Discharge Volume Level) . 42 Solenoid Valve Peactor 188 None Short ASCO Building (5 min.) Wr8300B61RV (Drywell Isolation) 43 Solenoid valve Drywell I-lc 2.16, 2.24 Short ASCO (2 min.) NP-8320A187E (Sample Valve) 44 Motorized valve Drywell I-2D 2.16 Short Actuator (2 min.) Limitorque SMB-2 (Isolation Condenser) 45 Electrical ' Drywell I-3 2.17, 2.18, Long tration 2.19 (30 days) General Eles.gic F01 46 Electrical Connectors Drywell I-4B 2.7, 2.16 Long ITT-Cannon . (10 days) CA-3106E-36A-46P-F80 CA-3100K-36A-46S-F80 47 Electrical Connectors Drywell I-4D 2.7, 2.16 Long ITT-Cannon (30 days) CA 06RX-36A-10P-A95 CA 3100RX-36A-10S-A95 48 Terminal Blocks Drywell I-5 2.16, 2.20 Long ) General Electric (30 days) EB . I 1 49 Electrical Cable Drywell I-6A 2.16, 2.21 Long I General Electric (30 days) l Vulkene SI-58145 4 B-9 dSJ$nidin Research Center  !

          =ca aw w ma aw. .                                                                       I

DELETED MATERIAL IS PROPRIETARY INFORMATK)N TER-C5257-195 EQUIPMENT ITEM ITEM QUALIFICATION TIME MO. DESCRIPTION LOCATION SCEWS NO. REFERENCES REQUIRED 50 Electrical Cable Drywell I-6A 2.16, 2.21 Iong General Electric (30 days) SI-58073 51 Electrical Cable Drywell I-6B 2.lG, 2.22 Long Tensolite Co. (30 days) No Model No. 62 Electrical Cable Drywell I-6C 2.16, 2.23 Long Kerite (30 days) No Model No. 53 Electrical Cable Drywell I-7A, 2.15, 2.16 Long Rockbestos I-7B (30 days) 54 Electrical Cable Drywell I-8 2.9, 2.16 Iong Splices (30 days) Raychem Corp. WCSF 55 Solenoid valve Drywell I-9 2.1, 2.16 Intermediate Dresser (4 h) 1525 VX (PORV) 56 Solenoid Valve Reactor 108 None Long ASCO Building (30 days) No Model No. (Drywell Isolation) A B-10 JhJEnklin Research Center awew u. _ _ , _ _ _ _ _ _ _ . , , . ~ , ._ ..

DELETED WATEMIAL 18 PMDenIETAAY 18ePopasATXps TER-C5257-195 APPENDIX C - SAFh"f SYSTEMS AND DISPIAY INSTRUMENTATION FOR WHICH ENVIRONMENTAL QUALIFICATION IS TO BE ADDRESSED The NRC transmitted to the Licensees for the SEP plants, Indian Point Units 2 and 3, and Zton Units 1 and 2 the DOR Guidelines for evaluating Class lE equipment qualification and the " Guidelines for Identification of That Safety Equipment of SEP Operating Reactors fot Which Environmental Qualification is To Be Addressed." Based on these documents, the Licensee sucaitted a list of safety-related systems that must function in order to mitigate the consequences of a design basis accident. As a result of discussions between the Licensee and the NRC, the following list represents systems and display instruments for which the Licensee and the NRC have determined that qualification is to be addressed. A. , Safe Shutdown Svstems - Reactor Protection System

• Isolation Condenser
• Domineralized Water Transfer +

Service Water Radiation Monitoring ++ Sampling ++ Emergency Diesel AC Power + 125 V DC Power

• Emergency Power Distribution
• B. Accident Mitigating Systems (LOCA, MSLB, FWLB)

Safeguards Actuation System Reactor Depressurization System Core Spray Main Steam Isolation Containment Isolation Containment Spray Standby Gas Treatment + Combustible Gas Control

• Systems used for both safe shutdown and accident mitigation.

      + Review of this equipment deferred until after February 1,1981, as referenced            ,

in Section 2.2.3.

     ++To be added as TMI-Lessons Learned requirement.

4MJ Franklin Research Center C-1 w arnevo m

DELETED M ATERf AL 68 PROMWETARY sNFOAMATION TER-C5257-195 C. Accident Mitigating and Safe Shutdown Instruments (LOCA, MSLB, FWLB) Reactor Water Level Reactor Steam Pressure Containment Drywell Pressure ** Containment Torus Water Level ** Containment Spray Flow'* Isolation Condenser Shell-Side Water Level Energency Service Water Pump Discharge Pressure Containment Spray Pump 3uction Pressure ** Domineralized Water Pump Discharge Pressure l

 **Instrumerits needed for accident mitigation purposes only.

l gh C-2 dd) Acza FranMn Research. Center

                      .# N r==ma  .

oststao uaramaus peopasTany wommanow TER-C5257-195 l APPENDIX D - EVALUATION OF LICENSEE JUSTIFICATIONS FOR CONTINUED OPERATION The Licensee *s documentation contained justification for interim plant I l operation where qualification had not been demonstrated for certain equipment items. At the request of the NRC, FRC conducted a technical evaluation of l these justifications based upon a review of technical information made available by the Licensee. ( In Chapter 7 of the' Licensee's final submittal (1], the Licensee presents l

   " justifications for continued operation" for equipment items that presently lack complete qualification.

l l FRC has performed technical evaluations of each of the positions which the Licensee presents in Chapter 7. FRC finds no technical deficiencies in these positions with the exception of five minor concerns that are expressed below. FRC's concerns involve the Licensee's assertions that these equipment items are not required to mitigate the consequences of an HELB or used for safe shutdoun of the plant. t Paragraph 17, Chapter 7. The Licensee has indicated that the main steam line low pressure switches are exposed to a peak temperature and pressure resulting from a feedwater line break and are protected frem the consequences of a MSLB. The actual temperature and pressure resulting from a MSLB have not been identified. FRC believes that these conditions should be established and that environmental qualification should be addressed for the main steam line low pressure switches (Equipment Item No. 13). i ! Paragraph 27, Chapter 7 It is not clear that the need to periodically i purge the containment throughout the long-term cooling period following an HELS outside containment is totally unnecessary such that the purge valves (Equipment Item Nos. 19, 20, 21A, and 22A) may be allowed to become inoperative. FRC believes that these valves should be qualified for their post-accident environment. Paragraph 28, Chapter 7. Similarly, it is not clear that there is no need to open the nitrogen system purge valves at some time following a LOCA. Therefore, the solenoid valves controlling these pneumatic valves (Equipment Items Nos. 23, 24, 25, and 56) should be qualified to operate under the environmental service conditions to which they may be subjected. i l I D-1 42 Franklin Research Center a cm a e n. w u.

DELETED t' ATE RfAL C PROMMTACY 18ePOAndATION TER-C5257-195 i Paragraph 37, Chapter 7 If sample valve V-24-30 (Equipment Item No. 35) is open coincident with an HELB, the inside conta'nment isolation valve cannot be relied upon to perform ith isolation function in view of the single active failure criterion. FRC believes valve V-24-30 should be qualified for the poet-accident environment to which it is exposed. Furthermore, the qualification of this valve is required in order to obtain post-accident samples in accordance with the recommendations of the 79tI-2 Lessons Learned Task Force. Paragraph 39, Chapter 7 FRC believes that drywell sump discharge valves (Equipment Item No. 36) should be qualified for their post-accident environment for long-term service because they will eventually need to be opened in order to remove contained fluids. 4 4Mu Franklin Research Center D-2 A Quiessen of The Feeruen visemme

1 i DELETt0 MAftmeAL IS PROPmETARY FWORMADON j TER-C5257-195 i APPENDIX E - CORRELATION OF EQUIPMENT ITDI NUMBERS WITH REPORT SECTIONS OF DRAFT IN!ERIM AND FINAL TECHNICAL EVALUATION REPORTS DRAFT INTERIM TECHNICAL FINAL TECHNICAL EVALUATION REPORT EVALUATION REPORT EQUIPMEfff ITEM NO. SECTION SECTION l 1 tbne 4.5.2.1 2 None 4.3.1.1 3A None 4.3.3.1 3B None 4.3.3.1 4A None 4.3.1.5 4B None 4.3.3.1 4C 3.3.2.1 4.5.2.2 5 None 4.7.1 6 None 4.6.1 7 Nong 4.6.2 8A None ~ 4.6.3 8B None 4.6.4 8C Nons 4.6.3 8D None 4.6.4 SE None 4.6.4 9 None 4.7.2 10 None 4.6.14 11 None 4.5.2.3 12A None 4.7.3 12B None 4.7.4 12C None 4.6.15 12D None 4.6.10 13 None. 4.7.11 14A None 4.6.16 14 B. , None 4.7.5 . 15 None 4.6.16 16 None 4.7.6 17 None 4.7.7 19 None 4.6.17 19 None 4.5.2.4 20 None 4.5.2.4 21A None 4.5.2.4 21B None 4.6.13 22A None 4.5.2.4 22B None 4.5.2.5 23 None 4.6.9 24 None 4.6.9 25 None 4.6.9 26 None 4.5.2.6 4Jg' j Franidin Research Center E-1 i a cmma at m vramm mammae

DELETED W ATERIAL IS PROPRIETAF (INFORMATION TER-C5257-195 CORRELATION OF EQUIPMENT ITEM NUMBERS WITH REPORT SECTIONS OF DRAFT INTERIM AND FINAL TECHNICAL EVALUATION REPORTS (Cont.) DRAFT INTERIM TECHNICAL FINAL TECHNICAL EVALUATION REPORT EVALUATION REPORT EQUIPMENT ITEM NO. SECTION SECTION 27 None 4.5.2.7 28 None 4.5.2.8 29 None 4.7.10 30 None 4.6.5 31A None 4.3.1.4 31B 3.2.1 4.5.2.9 32A None 4.3.1.4 32B 3.2.1 4.5.2.9 33 None 4.6.11 34A None 4.3.1.5 34B None 4.3.3.1 34C 3.3.2.1 _ 4.5.2.10 35 None 4.5.2.11 36 None 4.5.2.11 37 None 4.5.2.12 38 None 4.7.8 39 None 4.4.1 40 None 4.5.2.13 41 None 4.7.9 42 None 4.5.2.14 43 3.2.1 4.5.2.15 44 3.3.2.1 4.5.2.10 45 3.3.2.2 4.6.6 4C 3.3.2.3 4,5.2.16 47 3.3.2.3 4.5.2.16 48 3.3.2.4 4.6.7 49 3.3.2.5 4.3.1.2 50 3.3.2.5 4.3.1.2 51 3.3.4.1 4.5.2.17 52 3.3.2.6 4.3.3.2 53 3.2.2 4.3.1.3 54 3.3.2.7 4.3.3.3 55 3.3.2.8 4.6.12 56 None 4.6.8 E-2 4'j uTJ Fis.nklin Reserch Center s w ernev m maue

I DELETED WATEMA4.18 PROMMTARY INFORMATION TER-CS257-195 APPENDIX F - PROPERTIES OF CAST PHENOLIC RESINS rurstCAr. rnomms l l neraal normal Conductivity Espensine 'Je ter l Specifts Ipecifia (e.g.a. :oefft. Absorption

• crevier seat uits) (per *:) (=si a 10** z 10 5 cast 'esta 1.23-L.32 0.a-0.3 3-3 3-1 2-20 Moultime waterial

    'Jond-flaur-filled         1.3 1.4           0.35-0.36                6-12             34              70-130

hopped-cettse-faaric-filled 1.3 1.4 0.30-L 33 3-5  :-6 200-400 Mineral-filled 1.5-2.4 0.23 -1.33 8-20 2-. 10-100 L at.,ated s torial Faser-filled 1.0-1.6 0.3-0.4 3-4 2-3 13-300 Taaria-ft 1ed 1. 3- L. 6 0. 3-3. 4 5-4 2-3 200-300 essestoe-filled 1.3 2.0 3.;3-0.33 S 20 3 100-100 MECIANICAL F907t1T:I3 Oltimate Citimate 01stante Modulue si Modulus of Teesile Seeding shear Compreestoa 11asticity 11:141:7 Itrateth Strergen Strength Streesta (in tension) (La tarsias) !acect

(*bf/ind) (Ibf/ tad) (Lbf/ta2) (1bf/ini) (Ibf/ind) (1bflis3) Strength

• 3 x 10  : LJ 3

s 10 3 s 10 3 x 10 x 10 3

    ".aee 'esta                3-10             *-L3          6-4              10 30         300-1.000               0.1-3.3
    %ultise    w aterist
    *;ood-flour-filled         5-4             4-L3           5.L3             13-40      1. 300-L.300    300-500    0.1-1,3 i    Chopped-cottan-fabric-filled          5-4              5-L3         10-L3             20-35         ?co-L.*00    300-500    0.3 3.J Mineral-filled             4-4              5-13          6-13             10-13      1.000-2.300                3.1-L.0 taminar u M cerial Paper-fillee               3- 12          13-30           3-12             00-40     1.000-3.300                 1.2-;.J Fatetc-filled              5-20           13-30           3 12             30-45         500-L.300                  L-5 naeostoo-filled            ?-L2           10-13           &-4              30 50         300-2.000               0.2-L.0 Nethod af 5.3. 771 fst east resia sad souldias seterials: 3.3. 972 for laminated materiala.
    %fersece 2toru evicz.         1..M. and F.3. 11tc.e. M eno11: ? seine.10NDON ILITF1 3cosa '.td.,1967.

P00R BRIGINAl. O- F-1 d.C Franklin Research Center a cm a as ne r=ven wesee.

DELETED MATERIAL IS PROPMETARY INFORMATION TER-C5257-195 APPENDIX G - EFFECTS OF NUCLEAR RADIATION DOSE RATE ON CABLE PERFORMANCE DURING A LOCA More than 50 separate test reports on electrical cables were reviewed during the equipment environmental qualification evaluation. The major insulation materials used in the cable test samples were cross-linked polyethyleSa chlorosulfonated polyethylene ethylene propylene rubber Neoprene butyl rubber silicone rubber. (Proprietary flame-retardant additives and layered combinations of insulating materials and shields have also been used by various manufacturers to provide j special features required by Licensaes and their engineering contractors.) Testing typically involved irradiation up to 230 Mrd at dose rates l between 0.1 and 2.1 Mrd/h. Measurements of insulation resistance during the tests indicated that cable insulation resistance decreases with increasing dose cate, and that insulation resistance recovers af ter the exposure ceases. Typical reductions in insulation resistance aret from 10 11 to 10 ohms at the low (0.1-0.25 Mrd/h) dose rates 8 from 10 to 10 ohms at the higher (1-2 Mrd/h) dose rates. l There are insufficient test data to determine the mathematical relationship between insulation resistance and dose rate. There is, however, test evidence that the dose rate effect combines with the pressure, temperature, humidity, and spray conditions to further reduce insulation resistance. For very high dose rates (i.e. , greater than about 2 Med/h) during simulated LOCA conditions, insulation resistances in the range of 1000 to 10,000 ohms for 30 f t of cable (measured at 10 V de) have been experienced. l During LOCA, the dose rates calculated in accordance with conservative NRC recommendations are typically 1 to 3 Med/h gamma and 10 Med/h beta during the . I first 10 hours of the LOCA. (These data are for a nominal 1000 MW(e) plant.) l It can be seen that the dose rates for insulation subject to beta radiation ( exceed most test radiation dose rates by an order of magnittde. 4 G-1 l MJ FranMn a w a .e n.Resear.c.h n.a. Center

DELETED MATERIAL O PROPRtETCN INFORMATION TEP.-C5257-195 There is concern, therefore, that exposed cables (i.e., cables not protected from beta radiation by cable tray covers or conduit) will not retain high enough insulation resistance to transmit reliable control and instrumentation signals without attenuation and distortion during the early stages (the first 10 hours) of a LOCA. The Licensees of plants with exposed cables should carefully evaluate the possible effects of combined gamma and beta radiation dose rates, plus elevated temperature and moisture, on the ability of the cables to perform their functions. The evaluation should be based on available test data for the cables, or test data should be generated so that analysis can be performed. i 4 G-2 0d) Franklin Research Center 4 c an .e Th. nwen m-

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