ML20245B393
| ML20245B393 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/31/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0896, NUREG-0896-S08, NUREG-896, NUREG-896-S8, NUDOCS 8906230155 | |
| Download: ML20245B393 (94) | |
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AVAILABILITY NOTICE.:
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NUREG-0896 Supplement No. 8 Safety Evaluation Report related to the operation of Seabrook Station, Units 1 and 2 Docket Nos. 50-443 and 50-444 Public Service Campany of New Hampshire, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation May 1989 p o* ** o u, s
o - .- .- - - - - - - - - - - - - - - - - - - - - - - - - - _ _ - - _ _ _ _ _ -
i ABSTRACT This report is Supplement No. 8 to the Safety Evaluation Report (SER) (NUREG-0896, March 1983) for the application filed by the Public Service Company of New Hampshire, et al., for licenses to operate Seabrook Station, Units 1 and 2 (Docket Nos. STN 50-443 and STN 50-444). It has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission and pro-vides recent information on open items identified in the SER. The facility is located in Seabrook, New Hampshire. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can be operated by the applicant without endangering the health and safety of the public.
Seabrook SSER 8 iii
l l- !
l l 1
TABLE OF CONTENTS Page ABSTRACT ............................................................... iii li 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT ..................... 1-1 1.1 Introduction ................................................ 1-1
- 1. 8 Confirmatory Issues..... .................................... 1-2 1.9 License Condition Items...................................... 1-2 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMP 0NENTS ........... 3-1 3.9 Mechanical Systems and Components ........................... 3-1 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures ............... 3-1 3.9.6 Inservice Testing of Pumps and Valves .............. 3-2 3.11 Environmental Qualification of Electrical Equipment Impor-tant to Safety and Safety-Related Mechanical Equipment ...... 3-6 3.11.3 Staff Evaluation .................................... 3-6 3.11.3.3 Service Conditions ........................ 3-6 3.11.3.3.2 Temperature, Pressure, and Humidity Conditions Outside Primary Containment ........... 3-6 5 REACTOR COOLANT SY5 TEM............................................. 5-1 5.2 Integrity of Reactor Coolant Pressure Boundary............... 5-1
- 5. 2.1 Compliance With Code and Code Cases.................. 5-1 5.2.1.1 Compliance With 10 CFR 50.55a.............. 5-1 6 ENGINEERED SAFETY FEATURES......................................... 6-1 6.2 Containment Systems.......................................... 6-1 6.2.2 Containment Heat Removal System ..................... 6-1 6.2.2.2 Conclusion ................................ 6-1 6.2.3 Secondary Containment Systems............. .......... 6-1 6.4 Control Room Habitability.................................... 6-1 Seabrook SSER 8 v
TABLE OF CONTENTS (continued)
P_agg i
7 INSTRUMENTATION AND C0NTROLS....................................... 7-1 1 7.3 Engineered Safety Features System............................ 7-1 l 7.3.2 Specific Findings.................................... 7-1 7.3.2.8 Level Measurement Errors as a Result of Environmental Temperature Effects on Level Instrument Reference legs......... 7-1 7.5 Information Systems Important to Safety...................... 7-1 7.5.2 Specific Findings.................................... 7-1 7.5.2.2 Radiation Data Management System........... 7-1 7.5.2.4 Post-Accident Monitoring Instrumentation... 7-2 9 AUXILIARY SYSTEMS ....... ......................................... '9-1 9.1 Fuel Storage Handling .............. ........................ 9-1 9.1.1 New Fuel Storage ............................. ...... 9-1 9.3 Process Auxiliaries ......................................... 9-2 9.3.4 Chemical and Volume Control System .................. 9-2 9.3.4.2 Evaluation ................................ 9-2 9.5 Othe r Auxi l i a ry Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-3 9.5.1 Fire Protection ... ..... ............ .............. 9-3 9.5.1.4 General Plant Guidelines .................. 9-3 11 RADI0 ACTIVE WASTE MANAGEMENT ...................................... 11-1 11.5 Process and Effluent Radiological Monitoring and Sampling Systems ...................... ..................... 11-1 11.5.2 Evaluation Findings ................................. 11-1 13 CONDUCT OF OPERATIONS ............ . ............................ 13-1 13.1 Organizational Structure and Qualifications ................. 13-1 13.1.1 Management and Technical Support Organization ....... 13-1 13.1.1.1 Corporate Organization ........ ........... 13-1 Seabrook SSER 8 vi
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TABLE OF CONTENTS (continued)
Page 13.1.2 Operating organization .............................. 13-3 13.1.2.2 Operations ................................ 13-3 13.2 Training .................................................... 13-3 13.2.1 Licensed Operator Training Program .................. 13-3 13.2.2 Training for Nonlicensed Plant Staff ................ 13-5 13.3 Emergency Planning .......................................... 13-5 13.3.1 Introduction ........................................ 13-5 13.3.2 Background .......................................... 13-6 13.3.3 Evaluation of the Emergency Plan .................... 13-7 13.3.3.1 Emergency Response Support and Resources ................................. 13-7 13.3.3.2 Notification Methods and Procedures ....... 13-7 13.3.3.3 Emergency Communications .................. 13-8 13.3.3.4 Emergency Facilities and Equipment ........ 13-8 13.3.3.5 Accident Assessment ....................... 13-9 13.3.3.6 Medical Support ........................... 13-9 13.3.3.7 Radiological Emergency Response Training............................ ...... 13-10 13.3.4 Conclusion .......................................... 13-10 13.4 Operational Review ......................................... 13-10 13.4.2 Nuclear Safety Audit and Review Committee .......... 13-10 13.4.3 Independent Safety Engineering Group ............... 13-11 13.6 Physical Security Plan ..................................... 13-11 13.6.1 Introduction ....................................... 13-11 13.6.4 Access Requirements ................................ 13-11 14 INITIAL TEST PROGRAM ........ .................................... 14-1 15 ACCIDENT ANALYSIS ................................................ 15-1 15.1 Increase in Heat Removal by the Secondary System . . . . . . . . . . . 15-1 15.1.5 Steamline Rupture .................................. 15-1 15.8 Anticipated Transients Without Scram ....................... 15-1 15.8.1 Generic Letter 83-28 ............................... 15-1 15.8.1.2 Equipment Classification and Vendor Interface ................................ 15-1 Seabrook SSER 8 vii
1
. TABLE'0F CONTENTS (continued)
' APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW APPENDIX B REFERENCES APPENDIX D ACRONYMS AND~INITIALISMS L APPENDIX F NRC ' STAFF CONTRIBUTORS AND CONSULTANTS APPENDIX H ERRATA TO THE SEABROOK STATION SAFETY EVALUATION REPORT AND LITS SUPPLEMENTS TAPPENDIX W' CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.'2.1 -
EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS-
' APPENDIX X TECHNICAL EVALUATION OF CHANGES TO THE PUMP AND VALVE INSERVICE TESTING PROGRAM, SEABROOK STATION, UNIT 1 j
i Seabrook SSER 8 viii
i 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
~ 1.1 Introduction On March 7, 1983, the Nuclear Regulatory Commission staff (NRC or staff) issued a Safety Evaluation Report (SER), NUREG-0896, on the application of Public Serv-ice Company of New Hampshire (PSNH, hereinafter referred to as the applicant) for licenses to operate Seabrook Station, Units 1 and 2. In April 1983, the NRC issued the first supplement to the SER (SSER 1), in June 1983 the second supple-ment (SSER 2) was issued, in July 1985 the third supplement (SSER 3) was issued, in May 1986 the fourth supplement (SSER 4) was issued, in July 198'i the fifth supplement (SSER 5) was issued, in October 1986 the sixth supplement (SSER 6) was issued, and in October 1987 the seventh supplement (SSER 7) was issued.
This eighth supplement (SSER 8) provides information to update the status of the NRC review.
Each of the sections and appendices to this supplement bears the same designa-tion as the related portion of the SER. The contents of this document are sup-plemental to the initial SER and SSERs 1 through 7, and not in lieu of those documents unless otherwise noted. Appendix A is a continuation of the chronol-ogy of this safety review. Appendix B lists any references other than NRC docu-ments or correspondence between the NRC staff and the applicant cited in this supplement.* Appendix D lists acronyms and initialisms used in this supplement.
Appendix F identifies the principal staff contributors and consultants. Appen-dix H is a continuation of errata to the SER and its supplements. Appendix W is a Technical Evaluation Report prepared for the NRC staff by its contractor, EG&G Idaho, Inc. Appendix W evaluates the applicant's program for classifying all safety-related components other than reactor trip system components as safety related on documents and in information-handling systems that are used to control plant activities that may affect these components. Appendix X is the technical evaluation by EG&G Idaho, Inc., of changes to the pump and valve inservice testing program for Seabrook Station, Unit 1.
Appendices C, E, and I through V have not been changed by this supplement.
The NRC Project Manager for the Seabrook operating license review is Mr. Victor Nerses. He may be reached by telephone at 301 492-1441 or by mail at the follow-ing address:
Mr. Victor Nerses, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
- Availability of reference materials cited is provided on the inside front cover of this report.
Seabrook SSER 8 1-1
1.8 Confirmatory Issues In Section 1.8 of the SER and its supplements, the staff noted that some items have been resolved essentially to its satisfaction but that certain confirmatory information for these. items'has not yet been provided by.the applicant.
This supplement closes two of the confirmatory items. These items and the section of this supplement that presents the results of the staff's evaluation follow.
(52) Postaccident monitoring (7.5.2.4)
(55) Containment enclosure emergency cleanup system (6.2.3)
The remaining and additional confirmatory items and the sections of the SER or its supplements where they are discussed are listed below. -The staff has de-cided that the confirmatory issues listed.below may be resolved after initial operation.
(6) Loose parts monitoring system (4.4.5.3)
(45) Steam generator tube rupture (15.6.3)
(49) Cable tray supports (3.7.3)
(50) Turbine system maintenance program (3.5.1.3)
(51) Inadequate core cooling, TMI Action Plan Item II.F.2 (4.4.5.4)
(54) Tests, operational procedures, and support systems (5.4.7.5)
(56) Radiation data management system (7.5.2.2)
(57) Fire protection (9.5.1.4)
(58) Control room habitability (6.4)
(59) Initial test program (14)
(60) Sampling and analyses of effluents (11.5)
(61) Containment heat removal system (6.2.2) 1.9 License Condition Items In Section 1.9 of the SER, or in its supplements, the staff noted several issues for which a license condition may be desirable to ensure that staff requirements are met during plant operation if those requirements have not been met before the operating license is issued. The license condition may be in the form of a condition in the body of the operating license, or a limiting condition for operation in the Technical Specifications appended to the license. As of this supplement, the remaining license condition is:
(21) Safety parameter display system, TMI Action Plan Item I.D.2 (18.2)
The issues listed below are no longer license conditions. Those that are marked with an asterisk have become confirmatory items.
(13) Emergency planning (13.3) l (16) Implementation and maintenance of the physical security plan (13.6)
(22) Radiation data management system (7.5.2.2)*
(23) Fire protection (9.5.1.4)*
(25) Control room habitability (6.4)*
(26) Sampling and analyses of effluents (11.5)*
(27) Containment systems (6.2)*
Seabrook SSER 8 1-2 b - _ _ - - _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS l
3.9 Mechanical Systems and Components 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures In light of the thermal stratification found in the pressurizer surge line of several pressurized-water reactors (PWRs), the NRC staff issued Bulletin 88-11 on December 20, 1988. Since thermal stratification causes changes in piping stresses, f atigue life, and line deflections from those predicted in the orig-inal design, all licensees and near-term operating license applicants of PWR plants were requested to conduct visual inspection of the surge line, to up-date stress and fatigue analysis for ensuring compliance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
and to monitor thermal conditions'and line deflections. Actions requested.
should be completed within the periods specified in the bulletin, unless the NRC staff considers the changes acceptable.
In response to.the bulletin, the applicant submitted a letter' dated March 7, 1989, to which were attached Westinghouse Topical Reports WCAP-12151 and -12152 and a Westinghouse surge line isometric drawing (Drawing No. SURG-WOO 49, " Pres-surizer Surge Line of Seabrook Plant Unit 1"). The applicant also provided additional information in response to staff questions in April.1989 in Supple-ment 1 to WCAP-12151 and -12152. The submittals-indicated that at Seabrook Unit 1, the applicant had conducted a walkdown af ter hot functional testing, instrumented sensors, and performed a qualitative assessment to show the feasi-bility of compliance with the ASME Code. The following is the staff's evalua-tion of the information presented-in the above submittals.
Section 5.1 in WCAP-12151 and -12152 indicates that no signs of distress in the supports and no indication of any crushed insulation or signs of abnormal pipe movements were found during the walkdown conducted after hot functional testing.
This is positive evidence that clearances around the pipe were adequate to accommodate thermal expansion by stratification.
The staff reviewed the locations of thermal and displacement monitoring points.
Figure 1 in the WCAP-12151 and -12152 shows that there are two temperature mon-itoring locations. However, in its letter of March 7, 1989, and during recent discussions with the NRC staff, the applicant indicated that two additional tem-perature monitoring locations had been instrumented. Furthermore, the revised surge line isometric drawing shows three locations. In addition, the applicant had not described their intended application and had not specified the duration of the monitoring program. The applicant clarified these matters in a conference call by stating that there were three temperature monitoring locations and that this will be recorded in a detailed report to be submitted by June 30, 1989.
The staff reviewed a comparison of various operating parameters and thermal mon-itoring results at Seabrook Unit 1 with those of four similar Westinghouse-designed PWR plants for which plant-specific analyses had been performed Seabrook SSER 8 3-1
l (Tables 1 and 2 in Supplement 1 to WCAP-12151 and -12152). The staff found that the Seabrook operating parameters and monitoring results are bounded by l those of envelope transients used for plant-specific analysis in other plants, i In Section 3.0 in WCAP-12151 and -12152, Westinghouse determined that on the basis of a comoarison of layout and geometry with those of other plants, the !
surge line wou'd meet the ASME Code. The staff reviewed the detailt of the l surge line layout of these plants that were given in the isometric drawing and !
in Supplement 1 to WCAP-12151 and -12152. It found that the applicant's judg-ment was reasonable but qualitative. Quantitative data supporting this judgment are necessary.
Tables 4 to 8 in WCAP-12151 and -12152 present comparisons of moments and stresses at various locations of the surge line piping and supports at Seabrook and the other four plants. Such comparisons, although incomplete, provide a reasonable assurance of surge line integrity for a short duration of plant operation.
The effects of thermal stratification on the surge line consist of changes in stresses, line deflections, and fatigue life from those predicted in the orig-inal design. Although the information provided by the applicant at this time to demonstrate compliance with the ASME Code and all the commitments in the Safety Analysis Report is only qualitative, the staff believes that the surge line is likely to maintain its structural integrity during low power plant operation without undue risk to public health and safety. The staff's judgment is based on the following: (1) No discernible distress or damage in the surge line was observed during a walkdown af ter the hot functional tasting, which indicates that the stresses and line deflections were unlikely to reach a criti-cal stage to impair piping functionality; (2) since fatigue is a cumulative time effect, it is unlikely to induce unacceptable damage to the surge line in the relatively short duration of loe power plant operation; and (3) the applicant has committed to submit a detailed quantitative plant-specific surge line analy-sis by June 30, 1989, which should verify the structural adequacy of the surge line.
Conclusion The staff concludes that the information provided by the applicant at this time has qualitatively demonstrated compliance with the ASME Code in regard to the surge line. Since the applicant has committed to submit a detailed quantitative plant-specific surge line analysis by June 30, 1989, no discernible distress or damage in the surge line was obser/ed during a walkdown after the hot functional testing, and the delay requested fcr implementing item 2.a of NRC Bulletin 88-11 is of limited duration, the staff concludes that Seabrook may operate at icw power without undue risk to public health and safety.
3.9,6 Inservice Testing of Pumps and Valves Title 10 of the Code of Federal Regulations, Section 50.55a(g) (10 CFR 50.55a(g)),
requires that inservice testing (IST) of ASME Code Class 1, 2, and 3 pumps and valves be performed in accordance with Section XI of the ASME Code and applica-ble addenda, except if specific written relief has been requested by the licensee and granted by the Commission. In requesting relief under 10 CFR 50.55a(g)(6)(i),
the licensee must demonstrate that conformance with certain requirements of the applicable code edition and addenda is impracticable for its f acility.
Seabrook SSER 8 3-2
The regulation, 10 CFR 50.55(g)(6)(i), authorizes the Commission to grant. relief from these requirements upon making the necessary findings. This section con-tains the NRC. staff's findings with respect to granting or not granting the relief requested as part of the applicant's IST program.
.The NRC staff fourd the Seabrook First Ten Year IST Program, Revision 1, accept-able for implementation in SER Supplement No. 6, dated October 1986. By letters dated June 17, 1987, and January 25 and September 26, 1988, the applicant sub-mitted changes to Revision 1 of the IST program and additional information re-lated to requests for relief from certain code requirements that it determined were impracticable to implement at Seabrook Station. The letters dated Janu-ary 25 and September 26, 1988, were responses to conference calls with repre-sentatives from Public Service Company of New Hampshire and the NRC staff to discuss the changes to the IST arogram that were considered unacceptable or for which further information was necessary.
The major changes contained in the submittals are the addition and revision of three relief requests and the addition of a justification for cold shutdown. The applicant also made minor changes to the IST program that have been evaluated by the staff. The program for the first 10 year interval is based on the require-ments.of the 1983 Edition through Summer 1983 Addenda of the ASME Code; these requirements will rerain in effect through the first 120-mor,th interval of com-mercial operation.
Relief Request P-3 was evaluated by the NRC staff. Appendix X to this supplement contains the technical evaluation report (TER) that provides EG&G's evaluation of the applicant's changes to Revision 1 of the IST program, Relief Requests P-2 and P-4, and cold shutdown justification V-48.
The staff has reviewed the TER and concurs in.its evaluations and conclusions.
The staff's evaluation of Relief Request P-3 is provided below. Table 3.1 sum-marizes the relief request determinations and cold shutdown justifications. The granting of relief is based on the fulfillment of any commitments made by the applicant in its oasis for each relief request and the alternative proposed testing.
Relief Request P The appl 8 cant has requested relief from the allowable ranges of test quantities specified in Section XI, Table IWP-3100-2, for pump differen-tial pressure for service water system pumps SW-P-41A, SW-P-41B, SW-P-41C, SW-P-410, SW-P-110A, and SW-P-1108. The applicant proposes to test the pumps in accordance with ANSI /ASME OM-6, Table 6100-1. In addition to measuring hydraulic parameters, the applicant proposes to monitor vibration in accordance with ANSI /ASME OM-6, Table 6100-1.
Applicant's Basis for Requesting Relief - The requirement to declare a pump in-operative when pump performance exceeds the reference value by 3 percent is impracticable and is not clearly indicative of pump degradation for the follow-ing reasons:
(1) The 3 percent limitation is overly restrictive when compared with the total accuracy of the instrumentation used to gather the test data. To consist-l ently remain below the 3 percent limitation, extremely lower instrument loop accuracies than those required by Table IWP-4110-1 would have to be Seabrook SSER 8 3-3 L_________-___ _ _ _ - - _ _ - - . _ - - - - - - - - - - -
1 Table 3.1 Summary of relief requests, Seabrook Station, Unit 1 i Relief Section XI Equipment Alternative Action request TER/SER requirement identi- method of by )
number section and subject fication testing NRC j l
Pump 1.1.1 IWP-3300 All pumps Take multiple Relief 1 P-2 (TER) Measure pump in IST pump vibration granted j bearing program measurements in 10 CFR 50.55 '
temperature two orthogonal (g)(6)(i) directions quarterly.
Pump 3.9.6 IWP-3210 SW-P-41A, Test the pumps Relief P-3 (SER) Allowable ranges -41B,-41C, in accordance granted for pump differ- -410,-110A, with ANSI /ASME 10 CFR 50.55 ential pressure and -110B OM-6, Table (g)(6)(i) 6100-1, includ-ing vibration monitoring.
Pump 1.2.1 IWP-3100 and SW-P-41A, Calculate ir.let Relief P-4 (TER) -4110 -41B,-41C, pressure from granted Measure pump -41D,-110A, measured water 10 CFR 50.55 inlet pressure and -110B level above (g)(6)(i) instrument pump suction.
accuracy Valve 2.1.1 FW-V99, Exercise during Agree V-48 (TER) -V216, and cold shutdowns.
-V357 and CO-V340 established. This is particularly evident when testing pumps that have high flow rates and low discharge head characteristics such as the pumps listed above.
(2) Power plant operating systems were not designed to provide laboratory-type conditions required to meet the 3 percent limitation. The service water (SW) systems require the use of large butterfly valves using remote manual control to throttle large volumes of water to the reference flow rate.
Normally flow rates can only be established to 100 gpm of the specified reference flow.
(3) Reference values are specific sets of data determined by measuring or ob-serving pump performance when a specific pump is known to be performing its required function acceptably. Merely exceeding the 3 percent limita-tion is not a clear signal of pump degradation. It may signify that the reference value is at the lower side of the statistical scatter of the test data in comparison with other periodic test data.
(4) For the pumps listed, the difference between the differential pressure reference value and the 103 percent required action value is approximately Seabrook SSER 8 3-4
i
'l 2.0 psid. This is extremely restrictive and is easily exceeded by any l combination of statistical scatter, instrument inaccuracy, and minor flow variations. In the past, these pumps have been declared inoperable due to high differential pressure readings exceeding the allowed range by less i than 0.5 psid. Further evaluation of these conditions has shown there has been no pump degradation. Furthermore, since an actual increase in pump differential pressure is not indicative of degraded pump performance, it is not necessary to maintain such a strict upper limit.
(5) The minimum design flow requirements for the SW system pumps are as follows:
Flow Head Service water pumps 8,700 gpm 34.9 psid Service tower pumps 11,360 gpm 47.1 psid These components are currently tested at the following reference conditions:
Minimum / maximum limit Proposed limit Pump Flow Head (0.90/1.03xaP ref) (0.93/1.10xaP ref)
SW-P-41A 11,500 gpm 66.3 psid 59.7/68.3 61.2/72.9 SW-P-41B 11,500 gpm 67.1 psid 60.4/69.1 62.4/73.8 SW-P-41C 11,500 gpm 65.2 psid 58.7/67.2 60.6/71.7 SW-P-41D 11,500 gpm 66.1 psid 59.5/68.1 61.5/72.7 SW-P-110A 13,000 gpm 70.4 psid 63.4/72.5 65.5/77.4 SW-P-110B 13,000 pgm 69.6 psid 62.6/71.7 64.7/76.6 As can be seen from the above figures, there is a wide margin between the design flow requirements and the lower limit, while the difference between the current upper limit of 103 percent and the reference values is very small. Tht proposed upper limit of 110 percent would expand the allowed operating band to minimize unnecessarily declaring the pump inoperable due to a slightly high differential pressure.
(6) Relief from the 3 percent limitation will provide an acceptable level of quality and safety and will not endanger the health and safety of the public.
Evaluation - Section XI, Paragraph IWP-3210, states that if the ranges of Table IWP-3100-2 cannot be met, the owner shall specify in the record of tests the reduced range limits that will allow the pump to fulfill its function. In addi-tion, with these reduced limits, a licensee should be able to demonstrate that significant degradation would be detected. With the limits in Table IWP-3100-2, these safety-related pumps were being declared inoperable when they were known to be in good operating condition; therefore, the applicant has specified new range limits for these pumps as provided for in the ASME Code. The applicant has proposed to use the hydraulic range testing parameters in accordance with ANSI /ASME OM-6, Table 6100-1. The applicant has also proposed to monitor vibration in accordance with ANSI /ASME OM-6, Table 6100-1.
Seabrook SSEk 8 3-5
With its letter of September 26, 1988, the applicant submitted an engineering evaluation of the possibility of a failure mechanism that would be characterized by a degradation in hydraulic performance without an associated rise in com-
[ ponent vibration identifying the developing failure. Further, the evaluation addressed whether the possibility existed for a pump to be operating near the lower end of the performance band allowed by ASME Code,Section XI, when in fact an imminent failure would render it incapable of delivering the minimum design flow required in a worst-case accident scenario.
The applicant's engineering evaluation and discussions with the vendor resulted in the determination that because of the large margin available above the mini-mum design requirements, there is no known hydraulic failure mechanism resulting in a rapid failure of the service water system pumps - rendering them incapable of delivering the minimum design required head and flow - that would not be pre-ceded by symptomatic vibration readings.
Because of the large margin for the service water system pumps above the mini-mum design requirements, the applicant has concluded that vibration monitoring will easily detect a developing failure and render the pump inoperable because of high vibration readings well before the pump approaches the design-required flow or differential pressure.
On the basis of these margins, the applicant has also concluded there is no failure mechanism that would cause degradation in the hydraulic performance of ' I the service water system pumps in question that could result in a pump satisfac-torily passing all surveillance requirements while being unable to achieve mini-mum design flow requirements. The vendor has concurred in these conclusions.
Conclusion On the basis of (1) the impracticality of meeting the ASME Code requirements, (2) the applicant's engineering evaluation, (3) the fact that the applicant has not been able to consistently meet the code-specified differential pressure high limit for these safety-related pumps even though the pumps have been in good working condition, (4) the proposed less-restrictive limits as authorized by the code, and (5) the commitment to additionally measure vibration in accord-ante with ANSI /ASME OM-6, Table 6100-1, the staff recommends that relief be granted from the code requirements as requested. The applicant should follow the corrective actions specified in Paragraph IWP-3230(a) when measured pump differential pressure readings are within the " alert range" specified in ANSI /
ASME OM-6, Table 6100-1, and follow the corrective actions specified in Para- ;
graph IWP-3230(b) when the measured pump differential pressure readings fall !
.into the " required action range" in the same table.
3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment 3.11.3 Staff Evaluation 3.11.3.3 Service Conditions 3.11.3.3.2 Temperature, Pressure, and Humidity Conditions Outside Primary Containment A main steamline break in the main steam and feedwater pipe chase could result in the reopening of the two affected main steam isolation valves (MSIVs) because Seabrook SSER 8 3-6
superheated steam exceeds environmental qualification conditions. It is assumed that superheated steam would result in the consequential reopening of two MSIVs and, therefore, would result in the blowdown of two steam generators. The blow-down of two steam generators is an event that had not been evaluated in Chap-ter 15 of the Seabrook Station Final Safety Analysis Report (FSAR). New Hampshire Yankee (NHY) provided an evaluation in a letter dated June 5, 1986, demonstrating that the potential reactivity added to the core by the consequen-tial opening of two MSIVs is negligible when compared with the negative reac-tivity available by including the actual Cycle 1 shutdown margin of 3.8% Ak/k instead of the standard (1.3% Ak/k) value. NHY committed to an interim modi-fication of the Seabrook Station Technical Specifications to require a 3.8%
Ak/k shutdown margin in modes 1, 2, and 3. NHY also committed to perform a plant-specific analysis of the core response to a main steamline break assuming the consequential reopening of two MSIVs and taking credit for the standard shutdown margin of 1.3% Ak/k.
By letter dated November 4, 1987, the applicant submitted a reanalysis of the steamline break event assuming reopening of the two MSIVs. Included in the submittal was a request to return the shutdown margin specified in the Techni-cal Specifications to the standard margin of 1.3% Ak/k.
The steam release resulting from the rupture of a main steamline would cause an initial increase in steam flow, which would then decrease during the accident as steam pressure fell. With a negative moderator temperature coefficient, the cooldown results in an injection of positive reactivity and the possibility that the core will become critical and return to power. A return to power fol-lowing a steamline rupture is a potential problem mainly because of the high power peaking factors that would exist assuming that the most reactive rod cluster control assembly (RCCA) was stuck in its fully withdrawn position.
The analysis of a main steamline rupture in the Seabrook FSAR demonstrated that no departure from nucleate boiling would occur for any rupture assuming that the most reactive RCCA was stuck in its fully withdrawn position. For any break, in any location, no more than one steam generator would experience an uncontrolled blowdown even if one of the isolation valves failed to close, un-less a consequential failure resulted in the reopening of both MSIVs in the pipe chase. This would cause an additional cooldown of the primary system with a possible increase in the peak return to power. The reanalysis was performed to determine the results of such a failure. The major difference between the reanalysis and the analysis in Section 15.1.5 of the Seabrook FSAR is that a consequential failure of af fected equipment in the pipe chase caused by super-heated steam is assumed.
The four cases analyzed were:
1
- case 1: tubes uncovered late, reactivity calculations based on loop 1
- case 2: tubes uncovered late, reactivity calculations based on loop 2
- case 3: tubes uncovered at predicted time, reactivity calculations based on loop 1
- case 4: tubes uncovered at predicted time, reactivity calculations based on loop 2 l
Seabrook SSER 8 3-7 1
i
\1 h
I o
9 f
Plots showing the behavior of various core and plant parameters during the transient were compared for the four cases analyzed. The results of cases 1 0 and 3 were more limiting than those of cases 2 and 4 because the calculations l-were based on loop 1, which experienced the more severe cooldown. The tran- (;
l sient parameters for all cases analyzed were compared with those of the FSAR
. transient, and it was verified that the FSAR case bounds these four cases.
Therefore, the FSAR analysis remains limiting, and the conclusions of the Sea- ,
brook FSAR remain valid.
On the basis of the preceding evaluation, the staff finds that the reanalysis !
of the main steamline break showed that the results of the FSAR analysis are j still valid with an assumed shutdown margin of 1.3% Ak/k. As a result of this conclusion, the request to remove the interim requirement for 3.8% Ak/k shut-down margin and return to the standard margin of 1.3% Ak/k is acceptable.
Seabrook SSER 8 3-8
J 5 REACTOR COOLANT SYSTEM 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.1 Compliance With Code and Code Cases 5.2.1.1 Compliance With 10 CFR 50.55a On May 6, 1988, the NRC staff issued Bulletin 88-05, which requires holders of I construction permits and operating licenses to submit information regarding materials supplied by Piping Supplies Incorporated (PSI) at Folsom, New Jersey, and West Jersey Manufacturing Company (WJM) at Williamstown, New Jersey, and requests that licensees (1) take action to ensure that materials comply with American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code) and design specification requirements or are suitable for their intended service or (2) replace such materials. The NRC action was precipitated by the discovery that certified material test reports (CMTRs) for material supplied by-PSI and WJM contained false information about material supplied to the nuclear industry. Supplement 2 of the bulletin issued on August 3, 1988, identifies Chews Landing Metal Manufacturers Incorporated (CL) as a supplier of nonconform-ing material. A number of CMTRs apparently were used to certify that commercial-grade. steel met the requirements of ASME Code,Section III, Subarticle NCA-3800, through improper use of a domestic forging company's letterhead.
The applicant responded to Bulletin 88-05 in letter reports dated August 25 and October 14, 1988. The responses described the methodology used to identify, test, and evaluate WJM, PSI, and CL material; the document and procurement review effort; the test program; suppurt activities of the Nuclear Management and Re-sources Council (NUMARC), the Electric Power Research Institute, and Bechtel; the results of the review and testing programs; and an engineering evaluation and analysis of the nonconforming materials (flanges and fittings).
The applicant conducted a multifaceted program to identify and locate materials supplied by PSI, WJM, and CL and an indepth document review and field inspection covering piping systems, piping used as ductwork, and tubing. It also conducted additional confirmatory reviews using data provided by suppliers, vendors, and contractors. The applicant's efforts showed that 369 flanges and fittings were identified as having been supplied by WJM.
The 369 WJM-supplied flanges and fittings were tested, and a total of 30 com-ponents was found to have hardness values below Brinell hardness number (BHN) 137 (ultimate tensile strength [UTS] - 66 ksi). The 30 parts consisted of 12 flanges and 18 socket weld couplings.
Structural evaluation of the 12 nonconforming flanges was based on the assump-tion that the reduced flange cape"ty is linearly dependent on the yield strength of the material. The evaluation ot the socket weld couplings utilized the gen-eric acceptance criteria developed by NUMARC, which include separate qualifica-tions for pressure design and moment loading. In all cases, the flanges and fittings were found to be suitable for their intended service.
Seabrook SSER 8 5-1
l Description and Evaluation of Applicant's Responses The applicant conducted a multifaceted program to identify and locate materials supplied by PSI and WJM. The program was conducted in accordance with New Hampshire Yankee (NHY) Engineering Evaluation 88-020, " Process for Identification /
Location of Nonconforming Materials Supplied by PSI and WJM." The procedures included initial screening of CMTRs, and it was determined that PSI did not supply any material to Seabrook. Subsequently, however, it was discovered that CL may have supplied suspect B31.1 material, and CL-supplied material was included within the scope of the applicant's review.
The applicant conducted an indepth document review and field inspection cover-ing piping systems, piping used as ductwork, and tubing. Document packages were located through ASME N-5 data reports up to the N-stamp date and examined to determine if suspect material was present. Additionally, installation records such as weld process sheets were reviewed to discover materials purchased and installed after the N-stamp date, and work control documents including ASME NIS-2 forms were examined. Piping used as ductwork is designed to American National Standards Institute (ANSI) B31.1 criteria. For piping used as ductwork systems, piping isometrics were used to create a bill of materials similar to those found in ASME N-5 data reports. In some cases, the fittings and flanges were field traced to determine their source.
The results of the above document review and field inspection were independently checked by a review of records at Dravo (primary piping supplier / fabricator) and Gunyon (primary small-bore piping supplier / fabricator). A review of construction and operations storeroom material issue tickets was conducted. This provided additional verification that material issued for specific work activities was included in the work document package review.
The applicant conducted a procurement review to provide additional confirmatory assurance that all suspect flanges and fittings were identified. The procure-ment review effort consisted of (1) a record search by Connex (primary shop fabricator) and Gunyon and Rador to identify purchase order numbers, job numbers, suppliers, material descriptions, quantities of products shipped, heat identifi-cation, and shipment dates for products furnished to Seabrook Station; (2) a comprehensive search of the procurement / inventory data base for any direct pur-chases of PSI, WJM, and CL products by the applicant; and (3) a review of the NUMARC and NRC records review of WJM's purchase order file. As a further con-firmation, the applicant has sent out a vendor and contractor questionnaire regarding the use of any suspect material. The results of the applicant's identification efforts showed that (1) a total of 369 flanges and fittings in safety-related systems was positively identified as being supplied by WJM or could not be identified from traced field markings and therefore included in the evaluation; (2) no flanges and fittings were supplied by PSI; and (3) no flanges and fittings in safety-related systems were supplied by CL.
On the basis of its review of the letter report dated August 25, 1988, the staff finds that the applicant conducted a thorough and comprehensive search to identify and locate nonconforming flanges and fittings supplied by PSI, WJM, and CL in response to the requirements of Bulletin 88-05 and its Supplements 1 and 2. The staff also finds that the applicant was responsive to the action and reporting requirements of Bulletin 88-05 and Supplements 1 and 2 and that there Seabrook SSER 8 5-2
is a very high probability that all nonconforming flanges and fittings have been identified. The staff concludes that the applicant's identification efforts provide an adequate basis to resolve the nonconforming material identification concerns described in Bulletin 88-05 and are acceptable.
The licensee's test program was conducted in accordance with Procedure ES88-0-15,
" Testing, Evaluation and Reporting Methodology." In situ nondestructive tests of carbon steel were performed using Procedure NHY-EHT-1, "Equotip Hardness Testing."
Testing of stainless steel was performed using Procedure NHY-FN-1, " Delta Ferrite Inspection." The equotip hardness numbers were converted to BHNs and ultimate tensile strength using the American Society for Testing and Materials SA-370 conversion table.
Confirmatory independent laboratory tests were performed by J. Dirats and Bechtel.
The applicant has submitted the results of in situ and laboratory tests to NUMARC and Bechtel and has provided 65 additional WJM material samples consisting of l carbon and stainless steel flanges and fittings to Bechtel for laboratory j testing.
The results from the tests performed both in situ and in the laboratory were uti-lized in the NUMARC/Bechtel report listed in Supplement 2 to NRC Bulletin 88-05.
After discussion with the staff, in accordance with the staff position stated below, the applicant performed additional chemical analyses of weld neck flange material to determine if the chemistry of material outside the specification range of hardness values is consistent with the product chemistry expected for SA-105 material.
The applicant reported that 369 flanges and fittings were identified and tested as a result of the records review and field walkdowns of safety-related systems.
A total of 30 carbon steel, SA-105, fittings and flanges tested below a hardness value of BHN 137. For the components whose hardness values fell within the allowable range of hardness for SA-105 material (BHN 137 [UTS - 66 ksi] to BHN 187 [UTS - 90 ksi]), the hardness test was accepted as demonstrating adequate tensile strength. This is supported by the widespread industry testing data and specifically by recent NUMARC testing data on a substantial number of nuclear industry samples from heats of suspected materials. Of the ASME items, 29 had hardness values between BHN 124 and BHN 137. One non-ASME item of B31.1 material also tested below the BHN 137 value and was included in the evaluation.
There were 29 pieces that were tested having a hardness less than BHN 137. They represented 9 heats of materials that were used in 77 installed pieces. Actual tensile tests were performed on samples from all but three of these heats. The tensile test results indicated that even though the hardness tested below BHN 137, the ultimate tensile strength exceeded the 70,000 psi minimum for SA-105.
The remaining three pieces, for which there were no representative tensile tests, were tested for chemical composition and found to meet the SA-105 requirements.
Therefore, all installed flanges either met the BHN 137 hardness requirement, were from a heat that was acceptably tensile tested, or had an acceptable i chemistry test.
The results of the NUMARC and Bechtel program involving 7000 field tests and 1300 laboratory tests of samples, which included 26 samples from Seabrook,
)
Seabrook SSER 8 5-3
p 1
" support the acceptability of the applir. ant's material in that items having !
field-tested hardness values lower tha1 BHN 137 demonstrate acceptable levels ;
of ultimate tensile strength when samples.are tested. 'l
- Independent NRC tests' performed'by Brookhaven Nationel Laboratory on samples- j randomly selected by the staff were also found to be acceptable and consistent ;
with applicant and NUMARC tests on the same material.
s j l
. The staff developed the following guidance that represents an acceptabl'e approach j
'for addressing the suspect components:
) - <
(1) All installed suspect components must be identified, including all flanges, 1 i
fittings, and welded lugs.
(2) All components not meeting SA-105 supplemental hardness requirements
-(BHN 187-137) shall be tested to ensure that their chemistry is con-sistent~with the chemistry of SA-105 grade steel. These components must also be evaluated fcr the applicable service loading conditions and '
the reduced material strength, if applicable, of the components to determine the adequacy of design margins.
(3) A statistically valid sampling plan (such as U.S. Department of Defense MIL-STD-1050, Tables I and iia) shall be implemented to determine if the.
chemistry of. all the suspect parts' is consistent'with SA-105 grade steel.
Each heat of material representing installed material should be included.
(4) Credit may .,a taken for previously tested samples from representative heats
-for the sampling of the nonconforming components.under (2) above when the number of required heats is determined under (3).
1 The applicant's test program also included samples from heats containing blind flanges in order to ensure that potential material deficiencies in blind flanges
. would be identified. Since the NUMARC W ort had identified blind flanges as having the potential to contain suspect material, the applicant took samples from !
each heat'of blind flange material installed at Seabrook, in addition to the i original sample plan, in order to ensure a high confidence level in installed a material. Together with the Bechtel results from the sample tests performed on.
Seabrook material, a chemical analysis from each heat of blind flange material indicated that, with one possible exception, the blind flanges as well as other I product forms of material are acceptable. One heat of blind flange material j was outside the acceptable chemistry. Both blind flanges from this heat will '
l be replaced by the applicant to preclude any question of acceptability.
On the basis of the applicant's material and chemical testing program described above, which was supported by independent NRC-conductr' mechanical property and ;
chemistry tests considered together with the strue.ture, analysis discussed below, !
'the staff concludes that the components in question s e acceptable for the j
. intended use.
All 30 nonconforming WJM parts with BHN less than 137, with one exception, are ;
located in Safety Class 2 and 3 pipinq systems. The specifications called for l SA-105 material, to be manufactured in accordance with ANSI B16.5 or ANSI B16.11 l standards as applicable. The one exception is a small-bore reducing coupling designed and manufactured to the ANSI /ASME B31.1 Code.
Seabrook SSER 8 5-4
Although the NUMARC testing indicated that all suspect Seabrook samples had acceptable tensile strength, the applicant in addition performed a structural
-evaluation of the components with hardness values less than BHN 137. Structural evaluation of the 12 nonconforming flanges was based on the assumption that the- reduced flange capacity is linearly dependent on the yield strength of the material.. The tensile strengths were determined using more conservative ASTM SA-370 conversion from hardness to tensile strength. ANSI B16.5 indicates that flange pressure-temperature ratings are proportional to the yield strength of the material. ASME Code,Section III, Subarticle NC/ND-3658 contains equations that indicate that the maximum flange moment capacity is linearly dependent on the yield strength of the flange material. Table 1 of Enclosure 3 to the letter report dated August 25, 1988, identifies 12 flanges with computed ultimate ten-sile strengths less than the required 70 ksi. Reduced allowable moments and flange pressure ratings are shown in Table 1 along with the design values deter-mined in the original piping analyses. In all cases, the moment loadings were found to be substantially below the reduced allowable values. TLe actual flange design pressures were found to be less than the reduced allowable values.
A structural evaluation of the 18 nonconforming socket weld couplings is provided in Table 2 of Enclosure 3 to the letter report dated August 25, 1988. The eval-uation utilizes the generic acceptance criteria developed for NUMARC by Bechtel (Attachment A to Enclosure 3 to the above letter report, Tables AI and A3).
The generic acceptance criteria include separate qualifications for pressure design and moment loading. The pressure qualification in Table AI is based on the actual stress in the fitting as calculated from equation 3 of ASME Code,Section III, Subarticle NC-3641. The minimum required tensile strength is cal-culated as four times the actual stress with a minimum value of 40 ksi. The minimum cequired tensile strength is compared with the actual tensile strength from in situ testing to determine pressure qualification. Moment loading quali-fication of the fittings is based on the assumption that stress levels at the pipe-to-fitting interface are at the applicable limits allowed by code. The procedure is to determine the applicable margins (ratios) to code allowable values from Table A3 that were developed for a stress intensification factor of the pipe or weld of 2.1. Fittings are generically qualified when their ratios are equal to or less than 1.0. In all cases, the minimum required tensile strengths for pressure qualification were found to be less than the actual in situ tensile strengths of the nonconforming fittings, and the ratios to code allowable values for moment qualification were found to be less than 1.0.
Conclusions 01 the basis of its review of the letter report dated August 25, 1988, the staff finds that the applicant conducted an adequate testing program and struc-tural analysis of the nonconforming flanges and fittings using acceptable and conservative analytical methods and evaluation criteria. The staff also finds that the applicant was responsive to the action and reporting requirements of Bulletin 88-05, including Supplements 1 and 2, and that the applicant has quali-fied all WJM-supplied nonconforming parts as being suitable for the intended service. The staff concludes that the testing and analytical procedures used by the applicant to qualify the nonconforming parts and the results of the test-ing and analyses provide an adequate basis for resolving the concerns with respect to demonstrating adequacy for service. Although the staff does not con-sider thet the nonconforming parts fully satisfy the standards of the ASME Code, Seabrook SSER 8 5-5
1 1
i the use of this material is an acceptable alternative in accordance with '
. 10 CFR 50.55a(a)(3)(ii), because replacement of the suspect materials'with mate-rials fully complying with all specified requirements would result in hstdship or unusual difficulties without a compensating increase in the level of quality or safety.
I Seabrook SSER 8 5-6
i l
l t
1
-6 ENGINEERED SAFETY FEATURES !
1 6.2 Containment Systems 6.2.2 Containment Heat Removal System -
6.2.2.2 Conclusion In SER Supplement No. 7, the staff stated that it would allow operation with the present containment building spray / residual heat removal (CBS/RHR) pressure !
isolation configuration until the first refueling outage, but would condition !
the operating license to require that the applicant perform certain long-term i actions. Subsequently, in its letter dated March 30, 1989, the applicant stated :
that a design change is being developed that provides for long-term actions to l address the staf f's concerns.regarding the CBS/RHR system interface. This de- !
sign change involves the addition of a check valve in series with each of the l four existing check valves that provide isolation at the CBS/RHR system inter-face. The applicant submitted a complete description of this design change for staff review and approval on May 1, 1989. The applicant also committed, on j receipt of staff approval, to expedite the implementation of this design change )
and to try to improve the previously accepted schedule of installation before startup from the first refueling outage. On the basis of its review and evalu-ation of these commitments, the staff has concluded that the commitments are acceptable and that the license condition proposed in SER Supplement No. 7 is no longer required. The intplementation of these commitments is subject to staff audit.
6.2.3 Secondary Containment Systems In SER Supplement No. 6, the staff noted that the applicant would perform con-firmatory testing before the plant entered mode 4 to demonstrate that the heat-ing, ventilation, and air conditioning system serving the charging pump cubicles would maintain the required negative pressure in the cubicles during a design-basis accident. A requirement to demonstrate this feature before initial entry ,
into mode 4 and periodically thereafter has been incorporated into Sections j 4.0.4 and 4.6.5.1d.4) of the Seabrook Station Technical Specifications issued j with Seabrook Station Operating License No. NPF-67. The staff finds that the j reuirement to demonstrate this feature before initial entry into mode 4 and i periodically thereafter by imposing these Technical Specification requirements )
is much more restrictive than performing the one confirmatory test initially j committed to by the applicant. Therefore, this issue is closed. l l
6.4 Control Room Habitability In Section 6.4 of SER Supplement No. 7, the staff found that the inherent single- !
j failure problem of the control room habitability system restricted operation to no greater than 5 percent of rated power and stated that, before proceeding l above 5 percent of rated power, the applicant should demonstrate to the satis- !
faction of the NRC staff that the control building air (CBA) system provides l 1
i
.Seabrook SSER 8 6-1
i protection in accordance with the provisions of Sections 6.4, 6.5.1, and 9.4.1 .
of the Standard Review Plan (SRP, NUREG-0800) and General Design Criterion (GDC) 19 of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations.
In Supplement No. 7, the staff stated that NRC inspectors had observed that the control building heating, ventilation, and air conditioning (HVAC) (CBA) system was susceptible to the single failure of 4160-V ac bus E5 or E6 in the absence of operator action and that other problems had been noted in system operation and logic that may have negated the design basis in the applicant's safety analy-sis report. If a vital electric bus fails, the damper located on the discharge of the operable makeup air fan must be opened manually with a handwheel override operator to allow the control room to be pressurized. In this situation the control room cannot be isolated because the radiation monitors cannot function l
as desigeod. This realignment process also necessitates that the makeup air purge line valve, corresponding to the contaminated intake, be manually opened.
This allows the contaminated intake to be purged with air from the clean intake when the makeup air fan restarts.
In its letter dated March 2, 1987, the applicant committed to provide the de-tails of modifications to the CBA system for NRC staff review before they were implemented. In its letter dated January 22, 1988, the applicant described the proposed modifications to the system and provided a reanalysis of the post-loss-of-coolant-accident radblogical doses to the control room personnel as a result of the proposed modifications. In its letter dated June 17, 1988, the applicant provided additional information.
New bypass piping and parallel back-draft dampers have been added, in the modi-fication, around the normal makeup air supply fans and associated discharge dampers to guarantee a flow path during emergency modes of operation. Each out-side air intake valve remains in a throttled open position for both the normal and emergency modes of operation. If either bus E5 or E6 should fail, the oper-ating supply fan that would have been powered from the failed bus will shut down and the opposite train damper will close, resulting in a temporary loss of control room pressure. Manual actuation from the control room of the filter train powered from the unaffected bus will reestablish positive pressure using the bypass piping. The unaffected filter train will also automatically start I
on a high radiation or "S" signal. Therefore, control room pressurization will l
not be compromised in the case of a vital bus failure combined with a design-basis accident. The new design eliminates the need for a purging operation, since both intakes remain in a throttled open position; hence, the purge lines will be capped off and their associated purge valves removed. The nominal makeup air flow rate during emergency conditions is 600 cfm. Adjustments will be made during preoperational system balancing to ensure that the excess makeup air flow will not cause unacceptably high control room complex pressures and i that all associated design requirements will be satisfied.
1 In a letter dated March 30, 1989, the applicant stated that proposed technical specifications pertaining to this design change will be submitted by April 30, 1989, and that this design change will be completed by September 30, 1989. The applicant submitted the proposed technical specifications on April 28, 1989.
On the basis of its review of the information in the above submittals, the staff has concluded that the proposed modification, when implemented as described in tne submittals, would satisfactorily eliminate the single-failure problem Seabrook SSER 8 6-2
described in Supplement No. 7. In addition, the staff has concluded, on the basis of its independent analysis of potential control room operator doses, that the doses are within the acceptance criteria of GDC 19. Therefore, when the proposed modification is implemented, the control room habitability system will provide radiological protection in accordance with SRP Sections 6.4, 6.5.1, and 9.4.1 and GOC 19 and, hence, is acceptable. There are no open-items, therefore, regarding the findings in Section 6.4 of Supplement No. 7.
The applicant's commitments dealing with this design modification are subject to staff audit.
In the SER, the staff had determined that the control room habitability system was acceptable with regard to the protection provided against potential toxic gas. However, as a result of a recent offsite fire, by submittal dated June 17, 1988, the applicant committed to update its offsite chemical hazards evaluation by reevaluating the nearby toxic chemical sources and the applicable transpor-tation routes in so far as they affect control room habitability, and submit the findings to the NRC staff. By letter dated March 30, 1989, the applicant stated that this analysis update has been completed and that this updated analysis will be submitted for staff review by April 30, 1989. The applicant submitted the updated analysis on April 28, 1989. The applicant's updated analy-sis is subject to staff audit.
l Seabrook SSER 8 6-3
7 INSTRUMENTATION AND CONTROLS 7.3 Engineered Safety Features System
-7.3.2 Specific Findings 7.3.2.8 Level Measurement Errors as a Result of Environmental Temperature Effects on Level Instrument Reference Legs l
-In SER Supplement No. 5, the staff stated that the applicant had committed to implement the following change in the plant: " Generate a high containment pres-sure alarm (about 2.0 psig) from SI-PT-936 analog input (A0502). The alarm will be set as low as is feasible without causing nuisance alarms (channel has a span of.0 to 60 psig)." For the reasons noted below, the staff has determined that these containment pressure' instrumentation and alarms are not necessary; therefore, they need not be installed. !
The staff completed its review of a letter from the applicant dated June 9, 1988, concerning level measurement error due to reference leg heatup. The staff con-cluded that since the setpoint calculations associated with the replacement steam generator water level transmitter have demonstrated that operator action is no longer required for the feedwater-line-break (FWLB) event inside containment, the 4
. containment pressure alarms and indication that were required to enhance this 'j operator action need not be installed for that service at the Seabrook Station. l The Rosemount transmitters, which replaced the troublesome Veritrak transmitters l at Seabrook, have a proven history of successful operation in similar applica- l tions at other nuclear power plants. Also, the above-mentioned pressure alarms- j and indication were part of a past commitment by the applicant associated with !
an early (1986) applicant study of the FWLB event during which the Veritrak !
transmitters were used for measurements of steam generator water level. i 7.5 Information Systems Important to Safety 7.5.2 Specific Findings 7.5.2.2 Radiation Data Management System j l
Isolation Devices -
l In SER Supplement No. 5, the staff described the interim measures taken by the applicant in regard to the radiation data management system (RDMS), the appli- i cant's commitment to install qualified isolation devices, and the requirement to i submit the qualification documentation to the staff for review before implemen-tation. The staff evaluation of the acceptability of the electrical isolation devices for use in the RDMS at Seabrook is given below.
At Seabrook various radiation monitors of the RDMS and their associated micro-processor computers are designated as Class IE. The RDMS also provides informa- I tion to the safety parameter display system (SPDS), which is classified as non- !
l Class IE. The SPDS must be electrically isolated from the RDMS to prevent f aults Seabrook SSER 8 7-1
in the non-Class IE system from propagating to the Class IE system. One accept-able method for providing this isolation is to use qualified electrical isola-tion devices that testing has shown are able to withstand the maximum credible electrical fault. ,
At Seabrook, General Atomics (GA) RM 80 isolators are used to provide the re-quired isolation between the RDMS and the SPDS. Other isolation devices used l for other inputs to the SPDS were previously reviewed and found acceptable by the staff. The GA isolators were not acceptable to the staff because the test- '
ing that has been done to gualify the devices took credit for the use of fuses between the fault and the 1 solation devices. Because these fuses could be easily replaced with higher amperage fuses that could negate the required pro-tection, the staff found the testing unacceptable.
The applicant took steps to allow plant operation for the first fuel cycle.
These steps included installing an assembly that would not allow inadvertent substitution of a higher amperage rated fuse. The staff found this interim ,
installation acceptable. The applicant also committed to install non-fuse-dependent qualified isolation devices before startup following the first refueling outage. The applicant was required to submit the qualification documentation to the NRC for review before implementation.
In its letter dated November 18, 1988, the applicant submitted the required documentation, which included " Test Report on Electrical Testing of Class IE Communication Isolation Device for Digital Radiation Monitoring System" (GA E-255-1389) and " Test Report on Electrical Testing of Isolation Devices for Digital Radiation Monitoring System" (GA E-255-1333). The staff has reviewed these reports and finds that they have demonstrated that the GA RM 80 communi-cation isolation device is an acceptable isolator for this application. The testing included applying a maximum credible fault of 165 V ac and 30 amperes. ,
During this test the protective fuse normally supplied as part of the mounting l board was bypassed. The testing demonstrated that a fault on the non-Class 1E equipment will not propagate to the Class IE side. The fuse, though normally installed, is not required to provide the protective function and therefore these devices are acceptable.
On the basis of its review and evaluation of the information supplied by the applicant, the staff has concluded that the GA RM 80 communication isolation devices are acceptable for use in interfacing the RDMS and the SPDS. The staff further concludes that the license condition in Section 7.5.2.2 of SER Supple-ment No. 5 has been met in full and is no longer required. The staff notes that the applicant, by letter dated March 30, 1989, committed to have the GA RM 80 communication isolation devices installed and operable by September 30, 1989.
The implementation of this commitment is subject to staff audit.
7.5.2.4 Post-Accident Monitoring Instrumentation As noted in Section 7.5.2.4 of SER Supplement No. 7, the applicant has committed to implement a program for ensuring the capability to monitor plant parameters and system operating status during and following an accident. A requirement to implement this commitment has been incorporated in Section 6.7.4f of the Seabrook Station Technical Specifications issued with Seabrook Station Operating License No. NPP67. Therefore, this issue is closed.
Seabrook SSER 8 7-2
L i
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9 AUXILIARY SYSTEMS- !
l 9.1 Fuel Storage Handling l
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9.1.1 New Fuel Storage ,
1 On December 15, 1985, the staff issued Materials License No. SNM-1963, which j authorized receipt, possession, and use of enriched uranium in the form of fuel !
assemblies at Seabrook Station, Unit 1. When the 10 CFR Part 50 fuel-load li- '
cense was issued, certain conditions of License No. SNM-1963 were_ incorporated j into the 10 CFR Part 50 license. The purpose of this evaluation is to reassess' '
the appropriateness of certain conditions during the amendment of the 10 CFR 3 Part 50 license. l:
Condition 2B(4)a.1 of the fuel-load license currently requires that the appli- '
cant not have more than two fuel assemblies out of storage racks or shipping containers at any one time. Until the applicant develops spacing controls for more fuel assemblies out of storage racks or shipping containers, this is an appropriate and necessary procedure. The staff notes that the applicant has confirmed that spacing controls exist and are part of its written fuel-handling procedures (Station Operating Procedure RS 0722, Rev. 03). This has eliminated the staff's concern regarding the need for Condition 2B(4)a.1.
Condition 2B(4)a.2 of the fuel-load license requires the applicant to maintain 12-inch edge-to-edge spacing between the above two fuel assemblies, between fuel assemblies and shipping containers, and between fuel assemblies and storage con-tainers. This condition'should be revised to require the specified spacing only for fuel assemblies that are out of storage racks or shipping containers. The spent-fuel racks are heavily poisoned and therefore the additional spacing be-tween assemblies in and out of the racks is not necessary. Additionally, the fuel racks in the new-fuel vault are made of stainless steel, which performs as i a moderate neutron poison. More importantly, the new-fuel' vault is maintained l free of water and, hence, k-infinity of the new-fuel vault array is significantly i less than 1.0, which ensures safe subcriticality. Accordingly, the staff recom- I mends that Condition 2B(4)a.2 be revised to read:
PSNH shall maintain a minimum 12-inch edge-to-edge distance between :
fuel assemblies which are out of storage racks or shipping containers.
However, since the applicant has written fuel-handling procedures (Station Oper- ,
ating Procedure RS 0722, Rev. 03) that require maintaining a minimum 12-inch ,
edge-to edge distance between fuel assemblies that are out of storage racks or j shipping containers, this has eliminated the staff's concern regarding the need for Condition 28(4)a.2.
Condition 2B(4)b requires that the applicant not have more than 60 loaded ship- !
ping containers on site. At the time the 10 CFR Part 70 license was issued, the ;
shipping containers in use were limited to 60 containers in a fissile class III l shipment. This container has since been upgraded to fissile class I, which would allow the applicant to have 250 containers in a single shipment or array. By ;
i Seabrook'SSER 8 9-1
spacing the arrays in accordance with U.S. Department of Transportation require-ments, an unlimited number of containers can be safely brought on site. Accord-ingly, the staff recommends that Condition 2B(4)b be deleted.
Condition 2B(4)c requires that the applicant not store loaded shipping containers outdoors for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of receipt. At the time the condition was imposed by Materials License No. SNM-1963, the station was under construction and long-term outdoor storage was of concern to the staff. Because construction is complete, this is no longer a concern. Accordingly, the staff recommends *that Condition 2B(4)c be deleted.
An issue that had concerned the staff was what controls the applicant had placed on the fuel assembly dust wrappers. The plastic dust wrapper on each fuel as-sembly in the vault must be removed from the fuel assembly or must be open at the bottom so that water will not collect in the wrapper. If the storage array were to become flooded, the dust wrappers filled with water, and then the vault drained, the fuel assemblies could be well moderated and effectively coupled to other well-moderated fuel assemblies because the isolating water between the fuel assemblies had drained away. The staff evaluated the condition of full-density water within the fuel assembly and low-density water between the fuel assemblies. There is not enough steel in the storage racks to ensure that the array remains subtritical under this condition. However, in subsequent discus-sions and written correspondence, the applicant has stated that it has removed and will continue to remove the dust wrappers from the fuel assemblies before new fuel is stored in the vault. In addition, the applicant has confirmed that l the removal of the dust wrappers is part of its written fuel-handling procedures I (Station Operating Procedure RS 0722, Rev. 03). This has eliminated the staff's concern.
- 9. 3 Process Auxiliaries 9.3.4 Chemical and Volume Control System 9.3.4.2 Evaluation II.B.3 Postaccident Sampling System Criterion (10)
By letter dated March 30. 1988, the applicant provided additional information en the postaccident sampling system concerning Criterion (10) in Item II.B.3 of NUREG-0737. Criterion (10) states: " Accuracy, range, and sensitivity shall be adequate to provide pertinent dath to the operator in order to describe the radiological and chemical status of the reactor coolant systems." In the letter the applicant stated that it intended to implement a new method for performing postaccident boron analyses, which, on the basis of experience at Seabrook Station, is superior to the method used previously because of its operational simplicity and consistently reliable results. The new method has an analytical range of 1 to 10 parts per million of boron for the postaccident sampling system. It was tested against boron standards in a reactor coolant matrix specified in Item II.B.3 of NUREG-0737. Results of the testing showed that the constituents of the matrix did not affect the accuracy of the boron analyses. The staff concludes that this provision meets Criterion (10) in Item II.B.3 of NUREG-0737 and is, therefore, acceptable.
Seabrook SSER 8 9-2
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9.5 Other Auxiliary Systems-9.5.1 Fire Protection
-9.5.1.4 General Plant Guidelines l Ventilation-In SER Supplement No. 7, the staff stated'that with the submittal of the updated fire protection reports in November 1986, the applicant had met one of the two provisions of the license condition in the fuel-load license and that the license condition would be changed to reflect this. Since the fire hazards analysis dictated the need to modify the plant to incorporate charcoal filter unit detec-tion systems and other modifications'and the applicant has committed by letter I dated March 30, 1989, to complete these actions by September 30, 1989, the staff has concluded that it is not necessary.to condition the license to cover the .'
remaining provision.
The staff accepts the completion date committed to by the applicant, but in no case shall the applicant proceed above 5 percent of rated power without having the heating, ventilation, and air conditioning system charcoal filter unit detec-tion system and any-other modifications dictated by.the fire hazards analysis complete and operable. The implementation of this commitment is subject to' staff audit.
I Seabrook SSER 8 9-3
11 RADI0 ACTIVE WASTE MANAGEMENT 11.5 Process and Effluent Radiological Monitoring and Sampling Systems 11.5.2 Evaluation Findings In SER Supplement No. 5, the staff concluded that the applicant conforms with Attachments 1 and 2 to TMI Action Plan Item II.F.1 (NUREG-0737) except for the capability to obtain representative samples without excessive plateout. There-fore, the staff proposed a license condition that would read:
Before startup following the first refueling outage, the applicant shall demonstrate that the iodine / particulate sampling system is op-erable and will perform its intended function.
In a letter dated March 30, 1989,.the applicant stated that the operability of the iodine / particulate sampling. system has been demonstrated in the particulate and radiciodine transmission modeling and analyses that were performed by the applicant's contractor, Science-Applications International Corporation (SAIC).
The applicant also stated that the SAIC modeling and analyses indicated a need to improve the transmission factor for the wide-range gas monitor high-range sample line. The applicant has developed a design change to improve the trans-mission factor for this line. SAIC has modeled the revised design and deter-mined that it yields acceptable transmission factors. In its letter of Marcr. 30, 1989, the applicant committed to have this design change installed by Septem-ber 30, 1989. In the same letter, the applicant also committed to submit by May'30, 1989, a report, for staff review, of the SAIC modeling and analyses. On the basis of these commitments, the. staff has concluded that a lic'ense condition is not necessary. These commitments'are subject to staff audit.
l l
Seabrook SSER 8 11-1
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13 CONDUCT OF OPERATIONS 13.1 Organizational Structure and Qualifications i 13.1.1 Management and Technical Support Organization _
13.l.1.1 Corporate Organi7.ation By letter dated April 11, 1988, the applicant described changes to the corporate organization. .The structure reporting to the President and Chief Executive Office of New Hampshire Yankee has been realigned as shown in Figure 13.1. The responsibilities and duties of the various organizational elements include those that are necessary for the operation of Seabrook Station and that were previously reviewed and found acceptable by the staff.
Reporting to the President and Chief Executive Officer, New Hampshire Yankee, are the Vice President-Nuclear Production, Executive Director-Emergency Pre-paredness and Community Relations, Comptroller and Chief Administrative Officer, and Vice President-Engineering, Licensing and Quality Programs.
-The Vice President-Nuclear Production has full-time responsibility for the oper-ation of Seabrook Station. Reporting to the Vice President-Nuclear Production are the Regulatory Services Manager, Production Services Manager, Station Man-ager, and Training Manager. The Regulatory Services Manager is responsible for supporting Seabrook Station in the licensing and training areas. The Production Services Manager is responsible for backfit, general administrative, and project support. The Station Manager is responsible for overall management of Unit 1.
The Training Manager is responsible for the training at Seabrook Station, in-cluding the training center.
The Executive Director-Emergency Preparedness and Community Relations has over-l all responsibility for the development and implementation of both the onsite and offsite emergency response plans for Seabrook Station.
The Comptroller and Chief Administrative Officer is responsible for providing accounting and financial services.
The Vice President-Engineering, Licensing and Quality Programs is responsible for quality assurance and quality control and is the focal point of all corpo-rate engineering activitie . Reporting to the Vice President-Engineering, Licensing and Quality Programs are the Manager of Engineering and the Nuclear Quality Manager. The quality assurance program, under the direction of the Nuclear Quality Manager, is described in Section 17 of the SER.
New Hampshire Yankee has contracted with the Yankee Nuclear Services Division !
(YNSD) of Yankee Atomic Electric Company for certain operational and station l support services. New Hampshire Yankee has delegated to YNSD the necessary authority and organizational freedom to accomplish those support services; however, New Hampshire Yankee retains ultimate responsibility for monitoring and approving the work.
Seabrook SSER 8 13-1 m ml
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The staff's review of the modifications and previous staff safety evaluations indicate that the organization is capable of providing support for operation of the Seabrook Station and that the changes meet the appropriate criteria of Section 13.1 of NUREG-0800, the Standard Review Plan.
13.1.2 Operating Organization 13.1.2.2 Operations By letter dated April 11, 1988, the applicant modified the operating organiza-tion as shown in Figure 13.2. The major aspects of this modification were the removal of the Administrative Services Manager and staff, the Training Center Manager and instructors, the Maintenance Service Manager and engineers, and the Compliance Manager from the operating organization; the realignment of the re-porting for some functional entities; and the creation of the new position of Planning, Scheduling, and Outage Manager.
The staff's review of the new organization indicates that there is no substan-tive change in the functional activities carried out by the operating organi-zation except for certain training functions that are now the responsibility of the Corporate Training Manager. The staff concludes that the changes meet the appropriate criteria of Section 13.1 of NUREG-0800.
13.2 Training The applicant's training and requalification programs were evaluated and found acceptable by the staff in the SER. Revisions to the licensed operator requal-ification program were evaluated and found acceptable in Section 13.2.1 of SER Supplement No. 4. Further information on the training for shift technical ad-visors was evaluated and found acceptable in Section 13.2.2.1 of SER Supplement No. 5. Subsequent to those findings, the Commission revised its regulations on operators' licenses (Part 55 of Title 10 of the Code of Federal Regulations)
(10 CFR Part 55), 52 FR 9460, March 25, 1987; and 52 g 49372, December 31, 1987).
By letter dated April 11, 1988, the applicant provided marked-up page changes to Section 13.2 of the Final Safety Analysis Report (FSAR) for Seabrook Station.
These marked-up pages reflected changes that were made to ensure consistency with the revised regulations and with the organizational changes evaluated and found acceptable in Sections 13.1 and 13.4 of this supplement. The applicant stated that the changes on the marked-up pages of the April 11, 1988, submittal will be included in a future amendment to the Seabrook FSAR.
13.2.1 Licensed Operator Training Program The Vice President-Nuclear Production has overall responsibility for the qual-J ification and proficiency of Seabrook Station personnel. He has assigned re-l soonsibility for providing direction and control for the conduct of training at Seabrook Station to the Training Manager. The Operations Training Manager and the Training Support Services Manager report to the Training Manager. The Op-erations Training Manager is responsible for the development and implementation l' cf training conducted in support of the operator training programs, including licensed operator and nonlicensed operator training, technical training for the Seabrook SSER 8 13-3 i
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engineering staff, simulator training, and fire protection training. The Train-ing Support Services Manager is responsible for the administrative support of training activities, including program development, evaluation, and accredita-l tion activities. The applicant also has added a separate section to the program l to describe the licensed operator course on reactor startup experience. The i specific training objectives place additional emphasis on concepts that are l difficult to visualize in a classroom setting. In addition, the applicant has provided a more detailed description of training on detailed systems, mitigating core damage, and transient and accident analysis. The portion of the program that describes simulator training has been rewritten to reflect performance-based training principles as called for in 10 CFR Part 55 and the Institute of Nuclear Power Operations (INP0) accreditation program. Other specific areas of the pro-gram have been similarly modified to ensure that they reflect performance-based training.
The staff has reviewed the changes in the applicant's training program for licensed operators and concludes that these changes meet the applicable require-ments of revised 10 CFR Part 55 and Chapter 13 of NUREG-0800 and are therefore acceptable.
13.2.2 Training for Nonlicensed Plant Staff The Vice President-Nuclear Production has overall responsibility for the quali-fication and proficiency of Seabrook Station personnel. He has assigned re-sponsibility for providing direction and control for the conduct of training at Seabrook Station to the Training Manager. The Operations Training Manager and the Technical Training Manager report to the Training Manager. The Operations Training Manager is responsible for the development and implementation of train-l ing conducted in support of the operator training programs, including licensed operator and nonlicensed operator training, technical training for the e:mineer-ing staff, simulator training, and fire protection training. The Technical Training Manager is responsible for general employee training and for supporting organizational training, including maintenance, chemistry, health physics, in-strumentation and control, quality assurance, and supervisory / management train-ing. The Training Support Services Manager is responsible for the administrative support of training activities, including program development, evaluation, and accreditation activities. The applicant has modified the program description for technical staff and managers to reflect the performance-based training prin-l ciples incorporated in the INP0 accreditation program, which is endorsed in the Commission Policy Statement on Training and Qualification of Nuclear Power Plant Personnel (50 F_R 11147).
The staf f has reviewed the changes in the applicant's training program for nonlicensed plant staff and concludes that these changes meet the applicable requirements of Chapter 13 of NUREG-0800 and are therefore acceptable.
13.3 Emergency Planning 13.3.1 Introduction As set forth in the Federal Register (53 FR 36955, effective October 24, 1988),
the Commission amended its regulations to establish more clearly what emergency planning and preparedness requirements are needed for fuel loading and low-power operation of nuclear power plants. The amended rule specifies that the Seabrook SSER 8 13-5 l
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NRC's review of a licensee's onsite emergency plan would include certain emer-gency planning standards with offsite aspects. These planning standards-are specified in 10 CFR 50.47(d)(1) through (7). The requirement for a public alerting and notification system for the issuance of a license authorizing low power testing was eliminated by the rule change. However, nothing in the amended rule is intended to change the emergency planning standards that must be satisfied before a full power license may be authorized.
13.3.2 Background In SER Supplement No. 4 dated May 1986, the staff concluded that the Seabrook Station Emergency Plan provided an adequate planning basis for an acceptable state of onsite emergency preparedness and met the requirements of 10 CFR Part 50 and Appendix E thereto for issuance of a license authorizing fuel loading and operation up to 5 percent of rated power. This conclusion was based, in part, on the applicant's coordination effort with the Commonwealth of Massachusetts.
However, on September 20, 1986, the Governor of Massachusetts announced that he would not submit emergency plans for the Massachusetts portion of the plume exposure pathway emergency planning zone (EPZ) to the Federal Emergency Manage-ment Agency (FEMA) for review.
Following a revision to the Commission's regulations regarding offsite emergency planning where State and/or local governments decline to participate (52 FR 42078, effective December 8, 1987), the applicant submitted for NRC and FEMA review a utility prepared offsite emergency response plan, the Seabrook Plan for Massa-chusetts Communities (SPMC). In a memorandum and order dated November 25, 1987, the Commission determined that the SPMC was a bona fide plan to be used to com-pensate for the lack of participation in emergency planning by the Commonwealth of Massachusetts and local communities. The SPMC was developed in recognition of, and to compensate for, the fact that the Commonwealth of Massachusetts and local communities in Massachusetts located within the Seabrook plume EPZ are currently not participating in emergency planning. The SPMC is intended to be implemented by New Hampshire Yankee's Offsite Response Organization (NHY-0RO) and does not rely on previously developed State and local government plans for Massachusetts. The Seabrook onsite plan was revised on April 28, 1988, to reflect the development of the SPMC and the NHY-OR0. The SPMC is currently under review by FEMA.
The staff's evaluation of the revised Seabrook onsite plan is provided in Sec-tion 13.3.3 of this supplement. The staff used appropriate portions of the evaluation criteria of NUREG-0654/ FEMA-REP-1, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Revision 1, November 1980, in its review of the re-vised Seabrook plan against the planning standards of 10 CFR 50.47(d)(1) through (7). It should be noted that the term " State and local organizations" as used in the planning standards listed in Section 13.3.3 of this supplement refers to the States of New Hampshire and Maine and to NHY-0RO, the offsite response or-ganization as defined in Supplement No. 1 to NUREG-0654/ FEMA-REP-1, Revision 1, September 1988. The staff's conclusions are provided in Section 13.3.4 of this supplement.
Seabrook SSER 8 13-6 {
13.3.3 Evaluation of the Emergency Plan 13.3.3.1 Emergency Response Support and Resources
!. Planning Standard L Arrangements for requesting and effectively using offsite assistance on site have been made, arrangements to accommodate State and local staff at the licensee's near-site emergency operations facility have been made, and other ;
organizations capable of augmenting the planned onsite response have been identified.
Emergency Plan Evaluation The staff initially addressed this element of the Seabrook plan in SER Supple-ment No. 1 dated April 1983. Further staff evaluation regarding the onsite emergency support responsibilities of offsite organizations was provided in SER Supplement No. 4. The Seabrook plan continues to identify the local organiza-tions relied on to augment the planned onsite emergency response. Letters of agreement with Exeter Hospital for medical support anc' the Seabrook Fire Depart- 1 ment for fire-fighting and medical transportation assistance are included in the I plan. In Supplement No. 4, the staff concluded that the arrangements between the applicant and offsite organizations and the capability of the Seabrook emer-gency response facilities, including the accommodation of offsite organizations at the emergency operations facility (E0F), were adequate to support a response effort in the event of an emergency. The revised Seabrook plan describes the responsibilities of the NHY-0RO, including its concept of ~ operation, coordination with the State of New Hampshire and the Commonwealth of Massachusetts, and inter-action with the applicant's onsite organization. The NHY-0R0 aill occupy and operate the Massachusetts emergency operations center (E0C) at the applicant's E0F in Newington, New Hampshire.
The staff finds that the Seabrook plan provides adequate information on the identification of offsite organizations capable of augmenting the planned on-site response, the arrangements for requesting and effectively using offsite assistance on site, and the accommodation of offsite authorities et the EOF.
( 13.3.3.2 Notification Methods and Procedures Planning Standard Procedures have been established for licensee communications with State and lo-cal response organizations, including initial notification of the declaration of emergency and periodic provision of plant and response status reports.
Emergency Plan Evaluation In SER Supplement No. 1, the staff addressed the applicant's procedures and methods for notifying the Seabrook Station emergency response organization, the State of New Hampshire, and the Commonwealth of Massachusetts. In Supplement No. 4, the staff further addressed the applicant's means for providing early notification to offsite authorities.
'Seabrook SSER 8 13-7
As a result-of the nonparticipation in Seabrook offsite emergency planning by
.the. Commonwealth of Massachusetts and local communities, the applicant has estab-lished the NHY-0R0 as the organization responsible- for providing compensatory action in the Massachusetts emergency planning zone. . Section 11 of.the revised Seabrook plan describes the procedures and methods for initially notifying offsite authorities, including the NHY-0R0. Within 15 minutes of the declaration of an emergency at Seabrook, the Seabrook Station Short-Term Emergency Director will notify the New Hampshire State Police, the NHY-0R0 E0C contact point, and the Massachusetts State Police by means of the nuclear alert system. This system is staffed on a 24-hour basis at both ends of the communication links - the sta-tion, the E0C contact point, and the State Police dispatching points. Once the E0F/E0C is activated, the notification of offsite authorities in New Hampshire and Massachusetts will take place at this facility. Notification of the State of Maine is coordinated by the State of New Hampshire. The revised Seabrook plan also provides for periodic followup reports to offsite authorities on an as-needed basis.
On the basis of its review of the Seabrook plan, the staff finds that the appli-cant has adequately described the procedures for communicating with offsite response organizations.
13.3.3.3 Emergency Communications Planning Standard Provisions exist for prompt communications among principal response organiza-tions to offsite emergency personnel who would be responding on site.
Emergency Plan Evaluation The staff evaluated this element of the Seabrook plan in SER Supplement ~No. 4.
Section 7 of the reviewed Seabrook plan describes the emergency communications i network for notifying and coordinating activities with offsite and onsite emer- i gency response organizations. During the onsite emergency preparedness imple-mentation appraisal (EPIA) conducted from December 9, 1985, to June 13, 1986, the staff found that the provisions that exist for prompt communications to off- I site response personnel who would be responding on site were adequate. In addi-tion, the communication links between Seabrook Station and offsite authorities described in the Seabrook plan have been used during drills and exercises in the past 3 years and have been found adequate. The Seabrook plan provides for the testing of these communication links on a schedule consistent with the guidance in NUREG-0654.
On the basis of its review of the Seabrook plan, the staff concludes that the Seabrook plan adequately describes the provisions for prompt emergency communi-cations to offsite emergency personnel who would be responding on site.
13.3.3.4 Emergency Facilities and Equipment Planning Standard Adequate emergency facilities and equipment to support the emergency response on site are provided and maintained.
Seabrook SSER 8 13-8
Emergency Plan Evaluation
-The staff addressed this element of the Seabrook plan in Supplements Nos. 1 and 4. In Supplement No. 4 the staff concluded that the emergency facilities and equipment were~ adequate to support a response effort on site in the event of an emergency at Seabrook. In addition, the staff reviewed and found accept-able the emergency facilities and equipment during the onsite EPIA conducted from December 9, 1985, to June 13, 1986. Furthermore, the emergency facilities and equipment have been used during emergency drills and exercises in the past 3 years and have been found adequate.
On the basis of its review of the Seabrook plan and the results of the EPIA, the staff finds that adequate emergency facilities and equipment to support an emergency response on site are provided and maintained.
13.3.3.5 Accident Assessment Planning Standard Adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use on site.
Emergency Plan Evaluation The staff addressed this element of the Seabrook plan in Supplement Nos. 1 and 4. In addition, during the onsite EPIA conducted from December 9, 1985, to June 13, 1986, the staff found that the methods, systems, and equipment were in use as described in the Seabrook plan.
On the basis of its evaluations in Supplements Nos. 1 and 4, the review of the revised Seabrook plan, and the results of the EPIA, the staff concludes that the methods, systems, and equipment for monitoring actual or potential offsite consequences of a radiological emergency condition are adequate and are in use on site.
13.3.3.6 Medical Support Planning Standard l Arrangements are made for medical services for contaminated and injured onsite individuals.
Emergency Plan Evaluation The staff addressed this element of the Seabrook plan in Supplements Nos. I and 4. In Supplement No. 4, the staff found that the applicant had adequately responded to the medical support items identified by the staff in Supplement No. 1 regarding onsite training and first-aid facilities.
On the basis of its findings in Supplement No. 4 and a review of the April 28, 1988, revision of the plan, the staff concludes that the Seabrook plan continues to adequately describe the arrangements for medical services for Seabrook SSER 8 13-9
l contaminated and injured onsite individuals. Letters of agreement with medical .;
and transportation services organizations are addressed in Section 13.3.3.7 of 1
<this supplement. l l
.13.3.3.7 Radiological Emergency Response Training !
l
' Planning Standard j
-Radiological'l emergency response training has been made available to those off !
' site who may be called to assist in an emergency on site.
Emergency Plan Evaluation ;
The staff initially addressed the training program in Supplement No.1 and 1 subsequently reviewed and found it acceptable during the onsite EPIA conducted- ;
-from December 9, 1985, to June 13, 1986. The updated letter of agreement (LOA) !
with the Seabrook Fire Department provides for annual training of emergency l response personnel who may be called to assist in an emergency on site. In ;
addition, the updated LOA with Exeter Hospital calls for annual training of- l medical personnel.
l On the basis of the above, the staff concludes that the Seabrook plan adequately provides for.the training of those offsite response personnel who may be called to assist in an emergency on site.
13.3.4 Conclusion The staff concludes that the Seabrook Station Emergency Plan provides an adequate !
planning basis for an acceptable state of onsite emergency preparedness and meets the requirements of 10 CFR Part 50 and Appendix E thereto for issuance of a li- l]
cense authorizing low power testing and training up to 5 percent of rated power. !
i After receiving the findings and determinations made by FEMA in regard to State, j local, and utility emergency response plans and. preparedness, the staff will j provide its overall conclusion on the status of onsite and offsite emergency pre-paredness for the Seabrook Station in a supplement to the SER before a license authorizing operation above 5 percent of rated power is issued.
I 13.4 Operational Review 13.4.2 Nuclear Safety Audit and Review Committee The reporting relationship of the nuclear safety audit and review committee (NSARC) has been changed because of the revised corporate organizational struc-ture. The NSARC charter will be approved by, and the NSARC will report to, the President and Chief Executive Officer of New Hampshire Yankee.
The staff finds that this change meets the acceptance criteria of Section 13.4 )
of NUREG-0800. ;
l i
i I
Seabrook SSER 8 13-10 .
l
L 13.4.3 Independent Safety Engineering Group.
The reporting relationship of the independent safety engineering group (ISEG).
has been changed because of the revised corporate organizational structure.
The ISEG will report to and submit recommendations to the Vice President-Engineering, Licensing and Quality Programs.
The staff finds that this change meets the ac.ceptance criteria of Section 13.4
.of NUREG-0800.
13.6 Physical Security Plan
-13.6.1 introduction The applicant has filed with the NRC revisions to its physical security plan for Seabrook Station, Unit 1.
In the following section, the staff summarizes how the applicant proposes to meet the revised miscellaneous amendments and search requirements of 10 CFR 73.55.
On the basis of its review of the physical security plan, the staff has con-cluded that the proposed changes satisfy the revised miscellaneous amendments and search requirements of-10 CFR 73.55 (5'. FR 27817 and 27822) and the accom-panying record reporting requirements of 10 ETR 73.70. Accordingly, the pro-tection provided will ensure that the public health and safety will not be endangered.
13.6.4 Access Requirements Personnel Search In the physical security plan, the applicant has committed to conduct a search of all individuals entering the protected area except for legitimate Federal,
. State, and local law enforcement personnel on official duty, through the use of equipment designed for the detection of firearms, explosives, and incendiary devices. In addition, the applicant has committed to conduct a physical pat-I down search of an individual whenever it has reason to suspect that the indi-l vidual is trying to bring firearms, explosives, or incendiary devices into the protected areas, or whenever firearm or explosive detection equipment is out of service or not operating satisfactorily.
Vital Area Access 1
- i. The applicant has modified its physical security plan to limit unescorted access to vital areas during nonemergency. situations to individuals who require access l to perform their duties. It has established authorization access lists for each l vital area that are updated and approved by the cognizant manager or supervisor at least once every 31 days. The applicant has ensured that only individuals whose specific duties require access to vital areas during nonemergency situations are included on its site access list. It has also revoked access and retrieved badges and other entry devices before or simultaneously with the notification of termination of an individual's unescorted facility access.
Seabrook SSER 8 13-11 1' _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _
i i
Because of the commitments made by the applicant in revisions to its physical security plan, the staff concludes that the applicant has met the vital access requirements of 10 CFR 73.55(d)(7)(i)(A)(B) and (C).
Locks and Keys The applicant has committed to provide methods for reducing the probability of compromising keys, locks, combinations, and related access control devices used to control access to protected and vital areas. These methods include the rota-tion of keys, locks, combinations, and related access control devices every 12 months or the changing of these devices whenever there is evidence or sus-picion that any key, lock, combination, or related access control device may have been compromised or when an individual who has had access to any of these devices has had his or her access terminated because of a lack of trustworthy-ness or reliability or inadequate work performance. Only persons granted unescorted facility access are issued such entry devices.
Because of the commitments made by the applicant in revisions to its physical security plan, the staff concludes that the applicant has met the requirements of 10 CFR 73.55(d)(9) to reduce the probability of compromising keys, locks, combinations, and related access controi devices.
Emergency Access The applicant has committed to provide for the rapid ingress and egress of in-dividuals during emergencies or situations that could lead to an emergency by ensuring prompt access to vital equipment. The applicant will review annually i',s physical security plans and contingency plans and procedures to evaluate
- t. heir potential effect on the safety of the plant and personnel. Emergency access is granted by a licensed senior operator.
Because of the commitments made by the applicant in revisions to its physical security plan, the staff concludes that the applicant has met the requirements of 10 CFR 73.55(d)(7)(ii)(B) to ensure access to vital equipment during emer-gencies or situations that may lead to an emergency.
Protection of Secondary Power Supplies The applicant has committed to protect the onsite secondary power supply system for alarm annunciator equipment and nonportable communications equipment by including such equipment within a vital area boundary.
Because of the commitments made by the applicant in revisions to its physical security plan, the staff concludes that the applicant has met the requirements of 10 CFR 73.55(e) by protecting the secondary power supplies of alarm annunci-ator equipment and nonportable communications equipment.
Vital Area Entry / Exit Logging The applicant has committed to maintain a log indicating name, badge number, time of entry, and time of exit of all individuals granted access to a vital area except these individuals entering or leaving the reactor control room.
Seabrook SSER 8 13-12
1 l
Because of the commitments made by the applicant in revisions to its physical i security plan, the staff concludes that the applicant has met the requirements of 10 CFR 73.70(d) by maintaining an entry / exit log for individuals granted access to vital areas (except the reactor control room).
l Conclusion On the basis of the above evaluation, the staff finds that the applicant has met the requirements of the revised miscellaneous amendments and search requirements of 10 CFR 73.55 and the record reporting requirements of 10 CFR 73.70.
l
, Seabrook SSER 8 13-13 I
L - --- -__________ --- _
14 INITIAL TEST PROGRAM In the SER, the staff found the applicant's initial test program acceptable. In a letter dated March 30, 1989, the applicant committed to implement the initial test program in such a sequence that "the safety of the plant is never totally dependent on the performance of untested structures, systems, and components."
The staff finds that this commitment is consistent with the recommendations of Regulatory Guide 1.68 and is therefore acceptable. In its letter dated March 30, 1989, the licensee also committed to review any changes to the initial test pro-gram described in Chapter 14 of the FSAR in accordance with the provisions of 10 CFR 50.59 and to report any changes within 1 month of such changes. On the basis of these commitments, the staff has concluded that a license condition is not necessary. These commitments are subject to staff audit.
)
i Seabrook SSER 8 14-1
l 15 ACCIDENT ANALYSIS 15.1 Increase in Heat Removal by the Secondary System 15.1.5 Steamline Rupture In Section 3.11.3.3.2 of SER Supplement No. 5, the staff found the applicant's commitment to modify the Seabrook Technical Specifications to require a 3.8%
Ak/k shutdown margin in operating modes 1, 2, and 3 an acceptable resolution for a potential two-steam generator blowdown accident resulting from the reopening of two main steam isolation valves following a main steamline break. By letter dated November 4, 1987, the applicant submitted a reanalysis of this accident and asked to return the shutdown margin specified in the Seabrook Technical Specifi-cations to the standard margin of 1.3% ok/k. The staff's evaluation of this re-analysis is given in Section 3.11.3.3.2 of this supplement. Ln the basis of that evaluation, the staff finds the reanalysis and the request to ceturn the shutdown margin specified in the Seabrook Technical Specifications to the standard margin of 1.3% ak/k acceptable.
15.8 Anticipated Transients Without Scram 15.8.1 Generic Letter 83-28 15.8.1.2 Equipment Classification and Vendor Interface In Section 7.2.4 of SER Supplement No. 4, the staff stated that it was review-ing the applicant's responses to Item 2.1 (Part 2) and Item 2.2 (Part 1) of Generic Letter 83-28, which was issued by the NRC on July 8, 1983, to indicate actions to be taken by licensees and applicants based on the generic implications of the Salem anticipated transient without scram (ATWS) events. The staff has completed its review and its evaluation follows.
Item 2.1 (Part 2)
Item 2.1 (Part 2) of the generic letter requires licensees and applicants to con-firm that an interface has been established with the nuclear steam supply system (NSSS) vendor or with the vendors of each of the components of the reactor trip system, which includes
- periodic communication between the licensee or applicant and the NSSS ven-dor or the vendors of each of the components of the reactor trip system l
- a system of positive feedback that confirms receipt by the licensee or applicant of transmittals of vendor technical information The applicant submitted responses to Generic Letter 83-28, Item 2.1 (Part 2),
in letters dated November 4,1983, and May 4,1987. The applicant described the Westinghouse Interface Program for the Seabrook Station reactor trip system.
Seabrook SSER 8 15-1 I
- - - _ _ - - -- _ _ - _ - _ _-___ -_- _ _ _ - _ _ _ _ _ _ _ - a
L As'Seabrook Station's NSSS vendor, Westinghouse provides technical bulletins to update, supplement, or revise maintenance procedures, manuals, and testing re-quirements for Westinghouse-supplied components. Every technical bulletin that affects safety-related equipment requires a return receipt from the applicant to acknowledge that it has been received. Westinghouse also supplies an updated index (transmitted as a technical bulletin) on an annual basis for all technical bulletins transmitted during the previous submittal period.
When a technical bulletin is received at the station, it is assigned to the appropriate deparment(s) for review and implementation. The technical bulletins are distributed and reviewed in accordance with the applicant's Operating Expe- J rience Review Program. This program is reviewed annually to determine its effectiveness. i The applicant is also an active member of the Westinghouse Owners Group (WOG) and as such is included in all correspondence issued by Westinghouse on behalf of i the WOG, including proposals and recommendations pertaining to the reactor trip system components. These Westinghouse proposed technical changes are transmitted !
to the WOG representatives at the station and are then submitted to the appli- ,
cant's Document Control Center for distribution and evaluation. '
On the basis of its review of the applicant's responses, the staff finds that the applicant's statements confirm that an acceptable NSSS vendor interface program exists for components that are required for performance of the reactor trip function. This program meets the requirements of Item 2.1 (Part 2) of Generic Letter 83-28 and is, therefore, acceptable.
Item 2.2 (Part 1) l l
Item 2.2 (Part 1) of the generic letter states that licensees and upplicants shall describe in considerable detail their program for classifying all safety- l related components other than reactor trip system components as safety related j on plant documents and in information-handling systems that are used to control plant activities that may affect these components. Specifically, the licensee's I or applicant's submittal should describe (1) the criteria used to identify these components as safety related; (2) the information-handling system that identifies the components as safety related; (3) the manner in which station personnel use this information-handling system to control activities affecting these components; (4) the management controls that are used to verify that the information-handling system is prepared, maintained, validated, and used in accordance with approved procedures; and (5) the design verification and qualification testing require-ments that are part of the specifications for procurement of safety-related l
j components.
The applicant submitted responses to Generic Letter 83-28, Item 2.2 (Part 1), in letters dated November 4, 1983, and May 4, 1987. The staff and its contractor, EG&G Idaho, Inc., have reviewed these submittals. The results of the EG&G review are presented in Appendix W. The staff has reviewed the EG&G evaluation and agrees with the EG&G findings.
The staff's guidelines and its evaluation of the applicant's responses to the generic letter are provided below.
l l
Seabrook SSER 8 15-2
- l. - - _- -
h i
e
'"A's,Seabrook Station's NSSS vendor, Westinghouse provides technical bulletins to update, supplement, or revise maintenance procedures, manuals, and testing re-- ;
,'quirements for Westinghouse-supplied components. Every technical bulletin that
~
l affects' safety-related equipment requires'a return receipt from the applicant to acknowledge that'it has been received. Westinghouse also supplies an updated index (transmitted as a technical bulletin) on an annual basis for all technical i bulletins transmitted during the previous submittal period.
When a technical bulletin is received at the station, it is. assigned to the appropriate deparment(s) for review and implementation. .The technical bulletins
- are distributed and reviewed in accordance with the applicant's Operating Expe-rience Review Program. This program is reviewed annually'to determine its effectiveness.
~ The applicant is also an active member of the Westinghouse Owners Group (WOG) and as such is included in all correspondence issued by Westinghouse on behalf of the WOG, including proposals and recommendations pertaining to the reactor trip system components. There Westinghouse proposed technical changes are transmitted to the WOG representatives at the station and are.then submitted to the appli-cant's Document Control Center for distribution and evaluation.
On the basis of its review of the applicant's responses, the staff finds that the' applicant's statements confirm that an acceptable NSSS vendor interface progran exists'for components that are required for performance of the reactor trip function. This program meets the requirements of Item 2.1 (Part 2) of Generic Letter 83-28 and is, therefore, acceptable.
Item 2.2 (Part 1)
Item 2.2 (Part 1) of the generic letter states that licensees and upplicants shall describe in considerable detail their program for classifying all safety-related components other than reactor trip system components as safety related on plant documents and in information-handling systems that are used to control plant activities that may affect these components. Specifically, the licensee's or applicant's submittal should describe (1) the criteria used to identify these components as safety related; (2) the information-handling system that identifies the components as safety related; (3) the manner in which station personnel use this information-handling system to control activities affecting these components; (4) the management controls that are used to verify that the information-handling system is prepared, maintained, validated, and used in accordance with approved procedures;' and (5) the design verification and qualification testing require-ments.that are part of the specifications for procurement of safety-related components.
The applicant submitted responses to Generic Letter 83-28, Item 2.2 (Part 1), in f letters dated November 4, 1983, and May 4, 1987. The staff and its contractor, j EG&G Idaho, Inc., have reviewed these submittals. The results of the EG&G review are presented in Appendix W. The staff has reviewed the EG&G evaluation and agrees with the EG&G findings.
The staff's guidelines and its evaluation of the applicant's responses to the
-generic letter are provided below.
Seabrook SSER 8 15-2
(1) Program - The licensee or applicant should confirm that an equipment classification program exists that provides assurance that all safety-related components are designated as safety related on plant documents such as drawings, procedures, system descriptions, test and maintenance instructions, operating procedures, and information-handling systems so that personnel who perform activities that affect such safety-related components are aware that they are working on safety-related components and are guided by safety-related procedures and constraints.
In its submittals, the applicant described its program for identifying and classifying safety related equipment that meets the staff's requirements as indicated in the following evaluations of the specific items pertaining to the program.
i (2) Identification Criteria - The licensee or applicant should describe the criteria used to identify safety-related equipment and components.
l In its submittal, the applicant stated that Section 3.2.2 of the Seabrook l ' Final Safety Analysis Report provides the criteria used to identify equip-ment as safety related. The applicant also stated that the criteria are consistent with the staff position footnoted in Section 2.2 of the generic letter, with Regulatory Guide 1.26, and with American National Standards Institute Standard N28.2A-1975. The applicant's response meets the staff's. requirements for this item and is acceptable.
(3) Information-Handling System - The licensee or applicant should confirm that the equipment classification program includes an information-handling system that is used to identify safety-related equipment and components.
Approved procedures that govern its development, maintenance, and valida-tion should exist.
In its submittal, the applicant stated that the architect-engineer (United Engineers and Constructors) developed the equipment list, line list, Class 1E list, standard instrument schedule, and cable schedule. These lists and schedules were developed during the design, construction, and startup of Seabrook Station and are incorporated into the applicant's equipment classification information-handling system, which is routinely reverified and updated. The applicant's response meets the staff's requirements for this item and is acceptable.
(4) Use of Information-Handling System - The licensee or applicant should confirr that its equipment classification program includes criteria and procedm as that govern the use of the information-handling system to de-termine if an activity is safety related and if safety-related procedures for maintenance, surveillance, parts replacement, and other activities defined in the introduction to 10 CFR Part 50, Appendix B, are applied to safety-related components.
In its submittal, the applicant described the use of its information-handling system to determine whether an activity is safety related and what procedures are used for purchasing, operation, surveillance testing, and maintenance activities. The applicant's submittal meets the staff's requirements for this item and is acceptable.
Seabrook SSER 8 15-3
(5) Management Controls - The licensee or applicant should confirm that man-agement controis used to verify that the procedures for the preparation, validation, and routine use of the information-handling systems have been and are being followed.
In its submittal, the applicant stated that administrative and quality assurance programs, as well as station manual procedures, ensure that station requirements are followed. The applicant also stated that manage-ment controls govern the information-handling system. The applicant's submittal meets the staff's requirements for this item and is acceptable.
(6) Design Verification and Procurement - The licensee or applicant should document that past usage demonstrates that appropriate design verification and qualification testing is specified for the procurement of safety-related components and parts. The specifications should include quali-fication testing for expected safety service conditions and provide support for the licensee's or applicant's receipt of testing documentation that supports the limits of life recommended by the supplier. If such documen-tation is not available, the licensee or applicant should confirm that the present program meets these requirements. ;
In its submittal, the applicant stated that the approved station criteria !
for the preparation and review of material purchase requests ensure that j design verification and qualification testing is specified for the safety-related equipment procured. The applicant's submittal meets the staff's requirements for this item and is acceptable.
l (7) "Important-To-Safety" Com3onents - The licensee's or applicant's equipment !
classification programs s1ould include (in addition to the safety-related components) a broader class of components designated as "important to safety."
Because the generic letter does not require that the licensee or applicant furnish this information as part of its response, the staff's review of this item will not be performed.
On the basis of its review of the EG&G evaluation and its evaluation above, the staff finds that the applicant's statements confirm that a program exists for classifying all safety-related component 5 other than reactor trip system components as safety related on documents and in information-handling systems that are used to control plant activities that may affect these components.
This program meets the requirements of Item 2.2 (Part 1) of Generic Letter 83-28 and is, therefore, acceptable.
Seabrook SSER 8 15-4
l APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW March 2, 1987 Letter from applicant responding to Inspection Report No.
50-443/86-54.
September 18, 1987 Letter from applicant concerning Seabrook Plan for Massachusetts Communities.
September 21, 1987 Letter from applicant concerning its response to Generic Letter 87-12.
September 24, 1987 Letter to applicant concerning setpoint procedure for certain Rosemount process measurement transmitters at Seabrook Station.
October 19, 1987 Letter from ap9 icant concerning changes to Seabrook Station Operational Quality Assurance Program.
October 26, 1987 Letter to applicant from Region I regarding systematic assessment of licensee performance (SALP) report for the period April 1, 1986 to July 31, 1987.
October 29, 1987 Letter from applicant concerning the relocation of seismic monitors.
October 30, 1987 Letter to applicant transmitting 20 copies of Supplement No. 7 to the staff's Safety Evaluation Report (SER) for Seabrook Station.
November 3, 1987 Letter from applicant requesting that questions concerning the Seabrook Plan for Massachusetts Communities be sent to E. A. Brown, President, New Hampshire Yankee.
November 3, 1987 Letter from applicant concerning Licensee Event Report (LER) 87-018-00, " Security Event."
November 4, 1987 Letter from applicant concerning Seabrook Station Security l Training and Qualification Plan, Revision 4.
l November 4, 1987 Letter from applicant concerning main steamline break outside the primary containment.
November 16, 1987 Letter from applicant concerning LER 87-020-00, " Security Event."
November 23, 1987 Letter from applicant concerning NUREG-0737, Task II.D.1,
" Performance Testing of Relief and Safety Valves."
Seabrook SSER 8 1 Appendix A
November 25, 1987- Memorandum and order regarding the Seabrook Plan for Massa-chusetts Communities.
November 30, 1987 Letter from applicant transmitting 10 CFR 50.59 quarterly report.
November 30, 1987 Letter from applicant responding to Office of Inspection and Enforcement (IE)Bulletin 85-03.
December 9, 1987 Letter from applicant concerning LER 87-021-00, " Security Event."
December 18, 1987 Letter from applicant concerning LER 87-022-00, " Security Event." j December 18, 1987 Letter from applicant concerning LER 87-023-00, " Security Event." !
December 18, 1987 Letter from applicant concerning LER 87-024-00, " Security Event."
December 23, 1987 Letter to applicant concerning Seabrook Plan for Massachu-setts Communities - information' required for full power review.
December 28, 1987 Letter to applicant concerning its request to relocate three seismic monitors (TAC No. 66037). ;
l January 6, 1988 Letter from applicant responding to SALP report.
January 13, 1988 Letter to applicant concerning a steamline break with consequential failure due to superheated steam.
January 20, 1988 Letter from applicant concerning LER 87-026-00, " Security Event."
January 20, 1988 Lettei from applicant concerning LER 87-027-00, " Security Event."
January 22, 1988 Letter from applicant concerning proposed modification to control building heating, ventilation, and air condition-ing (HVAC) system.
January 25, 1988 Letter from applicant concerning inservice testing of pumps and valves (changes to Revision 1).
January 29, 1988 Letter from applicant transmitting Security Event Log. 4 February 1, 1988 letter from applicant concerning Security Event Re-port 88-001.
February 2, 1988 Letter from applicant concerning 10 CFR 50.54(cc) notification.
Seabrook SSER 8 2 Appendix A l ,
February 2, 1988 Letter to applicant transmitting SALP Report 50-443/86-99 (SALP conducted on September 17 and October 2, 1987).
February 2,1988 Letter to applicant concerning SALP report.
February 4,'1988 Letter from applicant concerning proposed revision of Sea-brook Station Offsite Dose Calculation Manual (ODCM).
February 5; 1988 Letter to applicant concerning timely support for the review of the Seabrook Plan for Massachusetts Communities.
February 5,1988 Letter to applicant concerning withholding from public.
disclosure the Seabrook Plan for Massachusetts Communities.
February 10, 1988 Letter to applicant concerning protective action for-beach population.
l l February 16, 1988 Letter from' Eastern Utilities _(Executive Committee of Seabrook Joint Owners) concerning update of financial information.
February 17, 1988 Letter from applicant concerning Security Event Re-port 88-002.
February 17, 1988 Letter from applicant concerning response to Generic Letter 88-02.
March 1, 1988 Letter from applicant concerning 10 CFR 50.59 quarterly report.
March 11, 1988 Letter from applicant concerning newspaper articles associated with Congressman Edward J. Markey.
March 18, 1988 Letter to applicant requesting additional information on the Seabrook Emergency Plan (vehicular alert and notifica-tion system).
March 30, 1988 Letter from applicant concerning NUREG-0737, Task II.B.3, Criterion (10).
March 30, 1988 Letter from applicant concerning response to NRC Bulletin 88-01.
March 30, 1988 Letter to applicant concerning NRC Bulletin 87-01, " Pipe Wall Thinning."
April 1, 1988 Letter from applicant concerning magnetic tape of validated meteorological data for Seabrook Station.
April 4, 1988 Letter to applicant requesting additional information on the First Ten-Year Interval Inservice Inspection (ISI)
Program for Seabrook Station. l Seabrook SSER 8 3 Appendix A
Y..
1
- m. 4 .,
1 April 7, 1988 Order issued by the Secretary of the Commission extending the time in which the Commission may review ALAB-883 to April 28, 1988.
April 11, 1988- Letter from applicant concerning Security Event Re-port 88-003.
April 11, 1988 Letter from applicant concerning changes to Sections 13.1, 13.2,13.3, and 13.4 of the Final Safety Analysis Report (FSAR) for Seabrook Station.
April 11, 1988 Letter from applicant concerning clarification of changes to the Seabrook Station Operational Quality Assurance' ,
Program.
- April 18, 1988 Letter from applicant transmitting Annual Environmental l Operating Report. -i i
April 21, 1988 Letter from applicant concerning comments on and clarifica- i tion of SER Supplement No. 7.
April 27, 1988 Letter from applicant concerning Security Event Log.
May 11, 1988 Letter to applicant enclosing an exemption from the require- )
ment of 10 CFR 50.54(w) to increase'the amount of required l property insurance from $620 million to $1.06 billion until !
Seabrook Station receives an operating license, which will i ellow the reactor to go critical or operate at any power 4 level. ,
May 16, 1988 Letter from applicant concerning request for comments and additional information regarding Congressman Edward J. ;
Markey's investigative report entitled " Drug and Alcohol Use at the Seabrook Nuclear Power Plant."
May 17, 1988 Letter to applicant concerning extension of construction completion date for Construction Permit CPPR-136 for Sea-brook Station, Unit 2.
May 31, 1988 Letter from applicant transmitting 10 CFR 50.59 quarterly report.
June 3, 1988 Letter from applicant responding to request for additional information on the ISI Program.
June 6, 1988 Letter to applicant concerning full-scale preparedness exercise schedule for Seabrook Station.
June 7, 1988 Letter to applicant concerning Generic Letter 88-03,
" Steam Binding of Auxiliary Feedwater Pumps."
June 9, 1988 Letter from applicant concerning level measurement error due to reference leg heatup.
Seabrook SSER 8 4 Appendix A
June 9, 1988 Letter'from applicant providing additional information in response to NRC staff's comments on the Seabrook Station Physical Security Plan.
June 15, 1988- Letter from applicant concerning anticipated transient without scram (ATWS).
June 17, 1988 Letter from applicant concerning NRC staff's request for additional information on proposed modification to control building HVAC system.
June 27, 1988 Letter to applicant concerning proposed revision to Sea-brook Station ODCM (TAC No. 67390).
July 8, 1988 Letter from applicant transmitting a response to NRC Bul-letin 88-04.
July 8, 1988 Representatives from NRC, Brookhaven National Laboratory, PSNH-NHY, and Altran Corporation met at Altran Corporation office, Boston, Massachusetts, to discuss the concrete cracking issues at Seabrook Station. (Meeting summary -
issued July 19, 1988.)
July 8, 1988 Letter from appliant concerning projected reference tem-perature values for pressurized thermal shock events.
July 8, 1988 Letter from applicant concerning changes to Seabrook Station Technical Specifications.
July 8, 1988 Letter from applicant requesting license amendment.
July 12, 1988 Letter from applicant concerning the environmental quali-fication of RG-58 coaxial cable.
July 18, 1988 Letter from applicant responding to NRC staff's request for additional information in regard to IE Bulletin 85-03.
July 21, 1988 Letter from applicant transmitting a response to NRC Bul-letin 87-02, Supplements 1 and 2.
July 22, 1988 Letter from applicant concerning NRC request to use site l property for contingency telecommunications.
July 22, 1988 Letter from applicant concerning Security Event Log.
July 22, 1988 Letter from applicant concerning diesel generator special report.
July 22, 1988 Letter from applicant requesting license amendment for changes consistent with the guidance of Generic Letter 87-09.
July 22, 1988 Letter to applicant requesting additional information on Seabrook Emergency Plan.
Seabrook SSER 8 5 Appendix A
_ - - _ _ _ _ _ _ _ _ _ _ __-__-_________--_____D
\
l l 1 July 25,1988 Letter to applicant concerning ATWS mitigation system.
July 26,1988 Letter to applicant concerning the witholding of informa- i tion from public disclosure (Figure 2-2 of the Federal l Emergency Management Agency (FEMA) Rep-10 design report).
August 4, 1988 Letter from applicant concerning environmental qualifica- 1 tion of RG-58 coaxial cable.
August 8, 1988 Letter from applicant requesting license amendment (admin-istrative controls), supplemental information.
August 8, 1988 Letter from applicant requesting license amendment (Sec-tions 3.0 and 4.0), supplemental information.
August 8, 1988 Letter from applicant requesting license amendment (Rose-mount transmitters), supplemental information.
August 10, 1988 Letter from applicant submitting the Seabrook Station ODCM, Revision 4.
August 11, 1988 Letter to applicant concerning f kancial coverage for the cost of low power operation, request for additional information.
August 22, 1988 Letter to applicant concerning Seabrook Unit 1 Technical Specifications.
August 23, 15o8 Letter to applicant concerning response to Generic Let-ter 88-05 (TAC Nos. 68949 and 68950).
August 24, 1988 Letter to applicant concerning control room habitability system.
August 25, 1988 Letter from applicant transmitting a response to NRC Bulle-tin 88-05.
August 31, 1988 Letter from applicant transmitting 10 CFR 50.59 quarterly report.
August 31, 1988 Letter from applicant replying to NRC request for additional financial information.
September 15, 1988 Letter to applicant transmitting FEMA's report on Seabrook exercise.
September 22, 1988 letter from applicant concerning the expiration of Seabrook .
Unit 2 construction permit on October 31, 1988. I September 26, 1988 Letter from applicant concerning inservice testing (IST) of pumps and valves (changes to Revision 1).
September 27, 1988 Letter to applicant transmitting Amendment 1 to Facility Operating License NPF-56.
Seabrook SSER 8 6 Appendix A
\
[.
l
? September 28, 1988 Letter to applicant transmitting an environmental assessment.
and finding of no significant impact.for temporary exemption from the schedular requirements of the Property Insurance Rule effective October 4, 1988 (10 CFR 50.54(w)(5)(i)).
September 30, 1988 Letter from applicant responding to NRC Bulletin 88-08.
October 4, 1988 Letter to applicant transmitting an exemption from the final rule amending 10 CFR 50.54(w), which requires $1.06 billion of property insurance. Seabrook has a previous exemption to carry $620 million until a power-level license is issued or the reactor is to go critical.
October 6, 1988 Letter from applicant concerning future FSAR submittals.
October 14, 1988 Letter from applicant transmitting additional supplemental information on NRC Bulletin 88-05.
October 14, 1988 Letter from applicant transmitting certification of Seabrook Technical Specifications.
October 14, 1988 Letter from applicant concerning completion of corrective action _regarding control circuit cable lengths (CDR.84-06-13).
October 19, 1988 Letter from applicant transmitting new addresses for NRC correspondence to Seabrook Station.
October 19, 1988 Letter from applicant concerning changes to Seabrook Station Operational Quality Assurance Program.
October 20, 1988 Letter from applicant responding to CLI-88-07.
October 25, 1988 Letter to applicant concerning Seabrook Updated Final Safety Analysis Report.
October 26, 1988 Letter from applicant concerning Security Event Log.
October 27, 1988 Letter to applicant concerning Seabrook Station bankruptcy proceeding.
October 27, 1988 Letter to applicant concerning the withholding of proprietary information from public disclosure.
October- 28, 1988 Letter from applicant responding to' request for additional information regarding "The Plan in Response to NRC Order CLI-88-07."
November 3, 1988 Representatives from NRC and PSNH-NHY met in Rockville, Maryland, to discuss applicant's readiness to operate Sea-brook Station, Unit 1, at low power levels (up to 5 percent of full power). (Meeting summary issued November 8, 1988.)
November 4. 1988 Letter from applicant concerning Operational Readiness Self Assessment Report.
Seabrook SSER 8 7 Appendix A
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I November.10, 1988 Letter from applicant responding to NRC letter dated f October 25, 1988, and its questions concerning bankruptcy j
! proceedings and negotiations between PSNH and Massachusetts 1 Municipal Wholesale Electric Company.
L November 18, 1988 Letter from applicant transmitting errata to meeting sum- ,
mary. transcripts. The meeting was held on November 3, i 1988, to discuss plant readiness for issuance .of a low- !
l power (5 percent) license. "
November 18, 1988 Letter from applicant concerning radiation data management system isolation qualification documentation - request for i withholding from public disclosure.
November 30, 1988 Letter from applicant concerning 10 CFR 50.59 quarterly report.
November 30, 1988 Letter from applicant transmitting response to Generic Let-ter 88-11.
December'2, 1988 Letter.from applicant concerning receipt and inspection of new fuel.
December 2, 1988 Letter to applicant concerning loss of decay heat removal.
December 15, 1988 Letter to applicant concerning response to NRC Bulletin 88-05.
December 15, 1988 . Letter from applicant concerning ATWS mitigation system.
December 27, 1988 Letter from applicant concerning interim response to NRC Bul-letin 88-11: pressurizer surge line thermal stratification.
December 27, 1988 Letter from applicant transmitting additional information on NRC Bulletin 88-08.
December 30, 1988 Letter from applicant transmitting supplemental response to NRC Bulletin 88-04.
December 30, 1988 Letter from applicant transmitting interim response to NRC Bulletin 88-11: pressurizer surge line thermal stratification.
January 3, 1989 Letter from applicant transmitting response to Generic Let-ter 88-17.
January 5, 1989 Letter to applicant concerning FEMA's consolidated report on Seabrook offsite planning and preparedness.
January 13, 1989 Letter from applicant transmitting additional information on NRC Bulletin 88-08.
January 13, 1989 Letter to applicant concerning its response to NRC Bul-letin 88-04.
Seabrook SSER 8 8 Appendix A
i January 17, 1989 Letter from applicant concerning clarification of licensed i operator status.
January 27, 1989 Letter from applicant concerning Security Event Log.
February 1, 1989 Letter to applicant concerning the withholding of informa-tion from public disclosure (Figure 2-4 of Addendum 1 of the FEMA-Rep-10 design report).
February 2, 1989 Letter to applicant concerning preliminary review of the Seabrook Station alert and notification system.
February 3, 1989 Letter from applicant responding to Generic Letter 88-17, which requests information pursuant to 10 CFR 50.54(f) regarding operation during conditions under which the nu-clear steam supply system (NSSS) is being or would normally be cooled by the residual heat removal system.
February 8, 1989 Letter from applicant concerning notification of change in licensed operator position.
February 21, 1989 Letter from applicant submitting a response to Generic Let- i 88-14 concerning verification of design and operation of in-strument air systems and a discussion of programs for main-taining proper instrument air quality.
March 7, 1989 Letter from applicant responding to NRC Bulletin 88-11:
pressurizer surge line thermal stratification.
March 13, 1989 Letter to applicant concerning completion of licensing issues and confirmation of commitments.
March 15, 1989 Letter from applicant concerning ATWS mitigation system.
March 20, 1989 Letter from applicant concerning decommissioning funding assurance in response to CLI 88-10.
March 20, 1989 Letter to applicant forwarding comments on its response to Generic Letter 88-17 for Seabrook Unit 1 pertaining to ex-peditious actions in regard to loss of decay heat removal (TAC No. 69776).
March 20, 1989 Letter to applicant concerning Westinghouse Topical Report WCAP-12151 and application for withholding proprietary in-formation from public disclosure.
l March 21, 1989 Letter to applicant concerning Seabrook radiation data '
management system isolation device qualification.
l March 30, 1989 Letter from applicant concerning licensing items status summary.
March 30, 1989 Letter from applicant concerning onsite property damage insurance.
Seabrook SSER 8 9 Appendix A
March 31, 1989 Letter to applicant concerning decommissioning funding assurance.
March 31, 1989 Letter from applicant transmitting a response to NRC Bul-letin 88-10: nonconforming molded case circuit breakers.
April 7, 1989 Letter to applicant forwarding Federal Emergency Management Agency correspondence.
April 7, 1989 Letter from applicant concerning licensed simulator instructor.
April 10, 1989 Letter from applicant transmitting a response to NRC Bul-letin 88-11: pressurizer surge line ther. mal stratification.
Enclosure includes proprietary report WCAP-1251 and with-holding from public disclosure requested and nonproprietary version.
April 10, 1989 Letter to R. E. Sweeney, Manager, Bethesda Licensing Office of PSNH, concerning draft document on decommissioning funding assurance dated March 20, 1989.
April 14, 1989 Letter to applicant requesting a report of the status of )
implementation of TMI Action Plan items at Seabrook. ]
April 17, 1989 Letter from applicant concerning diesel generator special report.
April 18, 1989 Letter from applicant transmitting the status of TMI Action Plan items.
April 20, 1989 Representatives from NRC and PSNH met in Rockville, Maryland, to discuss the NRC staff's comments of March 31, 1989, on PSNH's decommissioning funding assurance package of March 20, 1989. (Meeting summary issued April 26, 1989.)
April 24, 1989 Letter to applicant concerning safety evaluation for onsite emergency planning and in response to NRC Bulletin 88-11.
April 25, 1989 Letter to applicant concerning supplemental decommissioning funding financial assurance arrangements.
April 27, 1989 Letter to applicant notifying the NRC of employees' potential safety issues.
April 27, 1989 Letter to applicant transmitting responses to NRC staff com-ments on decommissioning funding assurance.
April 28, 1989 Letter from applicant transmitting a license amendment and Technical Specification change to the control room emergency makeup air and filtration subsystem.
April 28, 1989 Letter from applicant transmitting Security Event Log cover-ing the period January 1 through March 31, 1989.
Seabrook SSER 8 10 Appendix A
April 28, 1989 Letter from applicant transmitting Licensee Event Report 89-006-00 on mispositioned unborated water source locked valves.
April 28, 1989 Letter from applicant transmitting offsite hazardous chemical analysis update.
April 28, 1989 Letter from applicant certifying the Technical Specifica-tions for Seabrook Station, Unit 1.
April 28, 1989 Letter from applicant transmitting the Annual Environmental Operating Report for the period January 1 through December 31, 1988.
May 1, 1989 Letter from applicant concerning residual heat removal /
containment building spray system interface.
1 May 2, 1989 letter to applicant concerning supplementary decommission funding assurance submitted by PSNH on April 27, 1989.
l May 2, 1989 Letter from applicant certifying readiness for issuance of a low power operating license for Seabrook Station, Unit 1.
May 3, 1989 Letter from applicant concerning Security Event Report 89-501-00, " Identification of a Suspected Controlled Substanc. in the Protected Area."
l May 3, 1989 Letter to applicant concerning review of the decommissioning funding assurance plan for Seabrook Station.
May 8, 1989 Letter from applicant responding to the NRC staff's request for additional information regarding NUREG-0737, Item II.D.1.
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Seabrook SSER 8 11 Appendix A
APPENDIX B REFERENCES American National Standards Institute, ANSI B16.5, " Steel Pipe Flanges and
- Flanged Fittings."
-- , B16.11, " Forged Steel Fittings, Socket-Welding and Threaded."
-- ,- Standard N18.2A-1975, " Revision and Addenda to Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."
American National Standards Institute /American Society of Mechanical Engineers, ANSI /ASME B31.1, " Power Piping."
-- , ANSI /ASME OM-6 Standard, " Inservice Testing of Pumps," Draft 11, July 15, 1987.
American Society for Testing and Materials, SA-370, " Specification for Mechani-cal Testing of Steel Products."
American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code),Section III, Subarticle NC-3641, " Pressure Design of Piping Products."
-- ,Section III, Subarticle NCA-3800, " Metallic Materials Manufacturer's and Material Supplier's Quality System Program."
-- ,Section III, Subarticle NC/ND-3658, " Analysis of Flanged Joints."
-- ,Section XI, 1983 Edition through Summer 1983 Addenda.
U.S. Department of Defense, MIL-STD-1050, " Military Standard Sampling Procedures and Tables for Inspection by Attributes."
Westinghouse Electric Corporation, Topical Report WCAP-12151 (proprietary) and WCAP-12152 (nonproprietary), " Assessment of Thermal Stratification for the Seabrook Unit 1 Pressurizer Surge Line," February 1989.
-- , Supplement 1 to WCAP-12151 (proprietary) and WCAP-12152 (nonproprietary),
" Additional Information in Support of the Assessment of Thermal Stratification for the Seabrook Unit 1 Pressurizer Surge Line," April 1989.
Seabrook SSER 8 1 Appendix B c .- __ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _
i APPENDIX D ACRONYMS AND INITIALISMS ANSI American National Standards Institute ASME American Society of Mechanical Engineers ATSM American Society for Testing and Materials ATWS anticipated transient (s) without scram BHN Brinell hardness number CBA control building air CBS containment building spray CFR. Code of Federal Regulations CL Chews Landing Metal Manufacturers Incorporated CMTR certified materials test report E0C emergency operations center E0F emergency operations facility EPIA emergency preparedness implementation appraisal EPZ emergency planning zone FEMA Federal Emergency Management Agency FR Fe Jeral Register F5AR final safety analysis report FWLB feedwater line break GA General Atomics GDC general design criterion (a)
HVAC heating, ventilation, and air conditioning
( IE Office of Inspection and Enforcement INP0 Institute of Nuclear Power Operations ISEG independent safety engineering group ISI inservice inspection IST inservice testing LER licensee event report LOA letter of agreement MSIV main steam isolation valve NHY New Hampshire Yankee NHY-0R0 New Hampshire Yankee's Offsite Response Organization NRC U.S. Nuclear Regulatory Commission NSARC nuclear safety audit and review committee NSSS nuclear steam supply stem NUMARC Nuclear Management and Resources Council Seabrook SSER 8 1 Appendix D
L l l l'
l l ODCM Offsite Dose Calculation Manual j PSI Piping Supplies Incorporated PSNH Public Service Company of New Hampshire PTMS piping thermal monitoring system PWR pressurized-water reactor RCCA rod cluster control assembly RDMS radiation data management system RHR residual heat removal SAIC Science Applications International Corporation SALP systematic assessment of licensee performance SER safety evaluation report SPDS safety parameter display system SPMC Seabrook Plan for Massachusetts Communities SRP Standard Review Plan SSER supplemental safety evaluation report SW service water TER technical evaluation report TMI Three Mile Island UTS- ultimate tensile strength MJM West Jersey Manufacturing Company.
WOG Westinghouse Owners Group YNSD Yankee Nuclear Services Division Seabrook SSER 8 2 Appendix D
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1 1
i APPENDIX F l l
NRC STAFF CONTRIBUTORS AND CONSULTANTS ,
i Name Title Review branch
- Frederick Allenspach Operations Engineer Performance Evaluation Branch o George H. Bidinger Section Leader Fuel Cycle Safety Branch Donald Brinkman Senior Project Manager Project Directorate I-3 Margaret Chatterton Nuclear Engineer Reactor Systems Branch Herbert F. Conrad Senior Materials Materials Engineering Branch Engineer Jane Gibson Physical Security Safeguards Branch Specialist Charles G. Hammer Mechanical Engineer Mechanical Engineering Branch Shou-Nien Hou Senior Mechanical Engineer Mechanical . Engineering Branch Richard Lasky Electrical Engineer Instrumentation & Control Systems Branch David A. McCaughey Project Manager Fuel Cycle Safety Branch Thomac K. McLellan Mechanical Engineer Mechanical Engineering Branch Dolores Morisseau Training Assessment Human Factors Assessment Specialist Branch Victor Nerses Senior Project Manager Project Directorate I-3 Charles Nichols ' Senior Reactor Plant Systems Branch Systems Engineer Donald Perrotti Emergency Preparedness Emergency Preparedness Specialist Branch Sang C. Rhow Electrical Engineer Instrumentation & Control Systems Branch Madelyn Rushbrook Licensing Assistant Project Directorate I-3 James C. Stewart Electrical Engineer Instrumentation & Control Systems Branch Vincent D. Thomas Electrical Engineer Instrumentation & Control Systems Branch James Wing Chemical Engineer Chemical Engineering Branch The following contractor personnel contributed to this report:
Clair B. Ransom, EGG &G Idaho, Inc.
Herbert C. Rockhold, EGG &G Idaho, Inc.
Alan C. Udy, EG&G Idaho, Inc.
- 0ffice of Nuclear Reactor Regulation Seabrook SSER 8 1 Appendix F
l APPENDIX H ERRATA TO THE SEABROOK STATION SAFETY EVALUATION REPORT AND ITS SUPPLEMENTS SER Supplement No. 4 Section 8.3.1.2.8, 4th paragraph of section Delete last sentence of paragraph.
SER Supplement No. 5 Section 9.3.4.2, page 9-2, Criterion (5)
Change "0.15 part per million chloride." to "10 parts per million chloride."
Section 9.3.4.2, page 9-3, Criterion (5), 1st paragraph Delete last sentence of paragraph.
Section 13.1.1.1, page 13-2, 5th paragraph of section Insert " Regulatory Service Manager, Director of Emergency Preparedness" after Startup Manager in line 3 of paragraph.
Section 13.6.4, page 13-6, 2nd paragraph of section, 2nd sentence Insert a comma after the word " operational."
Section 13.6.6, page 13-7, last sentence Change last sentence to read, "All nonportable communication links are provided with an uninterruptible emergency power source."
Figure 13.1, page 13-9 Add " Regulatory Service Manager" and " Director of Emergency Preparedness" to those reporting to the Vice President-Nuclear Production.
SER Supplement No. 6 Section 8. 3.1. 2. 6, 1st senter.ce Change first sentence to read, "The cooling tower pump load (800 hp) is normally connected at the 52-second interval...."
Seabrook SSER 8 1 Appendix H
P Section 18.2,-page 18-5, Item (7) under " Conclusions" Delete- the words "not yet," and after the word " approved," add the words
-"(refer to Appendix 18A)".
1
'SER Supplement No. 7 Section 6.2.2, page 6-1, 2nd paragraph of section j Replace first sentence with "The RHR system design pressure is 600 psig." !
-Section 6.2.2. page 6-1, 3rd paragraph of section, 1st sentence Replace " check valves V-25 and/or V-26 into" with " check valves V-25, V-26, V-55, and/or V-56 into". i
'Section 6.2.2.1, page 6-1, 1st paragraph of section, 1st sentence Add "(i.e. , valves V-25, V-26, V-55, and V-56)" after " check valves".
Section 6.2.2.1, page 6-2, 5th paragraph of section, 1st sentence
. Change " applicant will provide a piping".to " applicant has provided a piping".
Section 6.2.2.1, pages 6-2 and 6-3, 6th paragraph of section Replace existing paragraph with "A high pressure alarm also is provided at the CBS pump discharge. The alarm setpoint will be adjusted to less than or equal to 350 psig. Actuation of this alarm will indicate excess leakage through the check valves. After a high pressure alarm or an alarm from the PTMS, the operator will act in accordance with the operating pro-cedure that has been developed by the applicant. The operating procedure is summarized as follows:"
.Section 6.2.2.1, page 6-3, subsection titled "Long-Term Actions," paragraph numbered (2)
Replace first sentence with " Upgrade the CBS system to RHR system suction pressure conditions." -l L Section 6.2.2.2, page 6-4, paragraph numbered (1), 1st sentence i
Change "under design, upset, and faulted plant conditions." to "under design, upset, faulted plant, and emergency conditions."
Appendix C, page 1, section titled "A-3 Westinghouse Steam Generator Tube l Integrity," 2nd sentence Delete " full flow demineralizert and".
Appendix V, page 5, Section 2.2.1, 1st paragraph of section, 1st sentence Change "CBA-F-8" to "CBA-F-38". I i
Seabrook SSER 8 2 Appendix H
i d
ijj - Appendix V, page 5, Section 2.2.1, 1st paragraph of section, 2nd sentence k Add " control building," after "in the".
b J Appendix V, page 5, Section 2.2.1, 3rd paragraph of section, first and last
( sentences i
Change "CBA-F-8" to "CBA-F-38".
Appendix V, page 5, Section 2.2.1, 4th paragraph of section Replace last sentence of paragraph with " Filter units EAH-F-9, EAH-F-69, FAH-F-41, FAH-F-74, CAP-F-40, and PAH-F-16 are charcoal-bed-type filters 4 inches deep. Filter units CBA-F-38 and CAH-F-8 are the tray type with 2-inch-deep charcoal beds."
Appendix V, page 6, Section 2.2.1, 6th paragraph of section Replace 2nd line of paragraph with "by doors located on one side of the unit with access to either side of each charcoal bank."
Appendix V, page 11, Section 3, 2nd paragraph of section, line 5 Change "CBA-F-8" to "CBA-F-38".
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Seabrook SSER 8 3 Appendix H
f l
APPENDIX W j CONFORMANCE TO GENERIC LETTER 83-28 ITEM 2.2.1 - EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS Seabrook SSER 8 Appendix W
EGG-NTA-7408 TECHNICAL EVALUATION REPORT CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1--
EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:
SEABROOK-1 AND -2 ,
1 Docket Nes. 50-443/50-444 i
Alan C. Udy 1
Published June 1987 Idaho National Engineering Lcboratory EG&G Idaho, Inc.
Idaho Fcils, Idaho S3415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN Nos. D6001 & D6002 Seabrook SSER 8 Appendix W
ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittals for Unit Nos. I and 2 of the Seabrook Station for conformance to Generic Letter 83-28, Item 2.2.1.
I i
l Docket Nos. 5b-443/50-444 TAC No 76416 Seabrook SSER 8 ii Appendix W
_ = _ - _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ -
FOREWORD This report is supplied as part of the program for evaluating licensee /apolicant conformance to Generic Letter 83-28 "Recuired Actions I
Based on Generic Implications of Salem ATWS Events." This work is being conducted for the U.S. Nuclear Regulatory Ccmmission, Office of Nuclear ,
Reactor Regulation, Divisien of Engineering and System Technology, by EG&G Idaho, Inc., NRR and I&E Support Branch.
The U.S. Nuclear Regulatory Ccmmission funded this work under the authorization B&R Nos. 20-19-10-11-3 and 20-19-40-41-3, FIN Nos. D6001 and C6002.
Docket No. 50-443/50-444 TAC No. 76416 Seabrook SSER 8 iii Appendix W
CONTENTS ABSTRACT ... ........................................ ...... .... ..... 11 FOREWORD .............................................................. iii
~1. INTRODUCTION ..................................................... I
- 2. REVIEW CONTENT AND FORMAT ........................................ 2
- 3. ITEM 2.2.1 - PROGRAM .............................................. 3 3.1 Guideline ...... ... ....................................... 3 3.2 Evaluation ................................................. 3 !
3.3 Conclusion ................................................. 3 4
- 4. ITEM 2.2.1.1 - IDENTI FICATION CRITERI A . . . . . . . . . . . . . . . . . . . . . . . . . . .
A 4.1 Guideline ........... ........... ...........................
4 4.2 Evaluation ....... .........................................
4 4.3 Conclusion ................ .... .......... ........ .......
ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM ....................... 5 5.
5 5.1 Guideline ................... ..............................
5.2 Evaluation ................. ... ........................... 5 5.3 Conclusion ................................................. 5
- f. ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING . . . . . . . . . . 6 6.1 Guideline . ........ ..... . . ..... . ... ... ... 6 6
6.2 Evaluation ...... ......... .... .................. ..... . .
6.3 Conclusion ............. . .. ............ . ............... 6 7
- 7. ITEM 2.2.1.4 - MANAGEMENT CONTROLS ............. .. ............
7.1 Guideline ............ ... ... . ....... . ............ 7 Evaluation . ......... 7 7.2 . ..... . ........ . ..... ... ..
Conclusion ...... 7 7.3 ... . . .. .. ...... . ........ ...
S
- 8. ITEM 2.2.1.5 - DESIGN VERIFICATION AND PROCUREMENT .............
8.1 Guideline . 8
{
8.2 Evaluation . . . .... . . . . . .. ... . . . 8 8.3 Conclusion ... ...... . ..... .. . . ...... . .... .. 8 9
- 9. ITEM 2.2.1.6 "IMPORTANT TO SAFETY" COMPONENTS . . ... . ..
9.1 Gaideline .... .. ..... . ..... ... ... .... 9 1C. CONCLUSION .. . .. . .. . .. . . .... . .. .. ..... . 10
- 11. REFERENCES .. . ... .. .. . . .. . ... . . ..... . . . 11 Seabrook SSER 8 iv Appendix W
CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1--
EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:
SEABROOK-1 AND -2
- 1. INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit I of the Salem Nuclear Power Plant f ailed to open upon an automatic reactor trip signal f rom the reactor protection system. This incident was terminated manually by the operator about 30 seconds after the initiation of the autcmatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior l
to tnis incident, on February 22. 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated cased on steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.
Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the NRC staff to investigate and report on the generic implications of these occurrences at Unit I cf the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implic:tions of the Salem unit incidents are reported in NUREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Commission (NRC)
I requested (by Generic Letter S3-28 dated July 8,1983 ) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to the generic issues raised by the analyses of these two ATWS events.
This report is an evaluation of the responses submitted by the Public Service Company of New Hampsnire, the licensee for Seabrook-1 and the applicant for Seaorock-2, for Item 2.2.1 of Generic Letter 83-28. The documents reviewed as a part of this evaluation are listed in the references at the end of this report.
Seabrook SSER 8 1 Appendix W
4.
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- 2. REVIEW CONTENT AND FORMAT Item 2.2.1 of Generic Letter 63-28 requests the licensee or applicant to submit, for the staff review, a description of their programs for safety-related equipment classification including supporting information, in considerable detail, as indicated in the guideline section for each sub-item within this report.
As previously stated, each of the six sub-items of Item 2.2.1 is
( evaluated in a separate section in which the guiceline is presented; an evaluation of the licensee's/ applicant's response is made; and conclusions l about its acceptability are drawn.
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l Seabrook SSER 8 2 Appendix W
- 3. ITEM 2.2.1 - PROGRAM 3.1 Guideline Licensee and applicants should confirm tnat an equipment classification program is in place which will provide assurance that all safety-related components are designatec as safety-related on plant documentation such as
. procedures, system descriptions, test and maintenance ' instructions and in information handling systems so that personnel performing activities that affect such safety-rela *.ed ccmponents are aware that they are working on safety-related components and are guided by safety-related procedures and constraints. Licensee and applicant responses which adoress the features-of this program are evaluated in the remainder of this report.
3.2 Evaluation The Public Service Conipany of New Hampshire responded to these requirements with submittals dated November 4, 19832and May 4, 1987.3 (bese submittals include information that describes their safety-related
~
equipment classification program. In the review of the utility's response to this item, it was assumed that the information and documentation supporting this program is available for audit upon request. We have reviewed this information anc note that the utility states that plant documents, operational procedures, system descriptions and the computerized information handling system designate the safety-related status of structures, systems ccmponents and parts.
3.3 Conclusion We have reviewed the utility's information and, in general, find that their response is adequate. I i
Seabrook SSER 8 3 Appendix W l
- 4. ITEM 2.2.1.1 - IDENTIFICATION CRITERIA 4.1 Guideline The applicant or licensee should confirm that their program used for equipment classification includes criteria used for identifying components as safety-related.
4.2 Evaluation The utility states that Section 3.2.2 of the Seabrook Station FSAR provides details on the criteria used to identify a component as .
safety-related. They further state that the criteria is consistent with the staff position footnoted in Section 2.2 of tne generic letter, witn Regulatory Guide 1.26 and with ANSI N18.2A-1975.
4.3 Conclusion We find that the utility has confirmed that they have identified the criteria used in the identification of safety-related components, thus neeting the requirements of Item 2.2.1.1.
I i
Seabrook SSER 8 4 Appendix W 1
- 5. ITEM 2.2.I.2 - INFORMATION HANDLING SYSTEM 5.1 Guideline The licensee or applicant should confirm that the program for equipment classification includes an information handling system that is used to identify safety-u 'ated components. The response should confirm that this information handling system includes a list of safety-related equipment and that procedures exist which govern its development and validation.
5.2 Evaluation The utility states that the architect / engineer (United Engineers and !
Constructors) developed the following lists:
i
- 1. Equipment list
- 2. Line list
- 3. Class-1E list ,
- 4. Standard instrument senedule
- 5. Cable schedule These lists and schedules were developed and used during the design, construction and startup of the Seabroek Station. The procedures used in preparing these lists and schedules, which are incorporated into the licensee's/ applicant's equipment classification information handlir.g system, are stated to have included specific instructions for documentation and verification. The lists are routinely re-verified and incorporate change documents upon completien of the work involved in the change.
5.3 Conclusion The utility's response for this item is considered to be complete and is acceptable.
Scabrook SSER 8 5 Appendix W
- 6. ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING 6.1 Guideline The licensee's or applicant's description should confirm that their program for equipment classification includes criteria and procedures governing the use of the equipment classification information handling system to determine that an activity is safety-related and what procedures for maintenance, surveillance, parts replacement and other activities defined in the introduction to 10 CFR 50, Appendix B, apply to safety-related components.
6.2 Evaluation The utility states that administrative and quality assurance programs are in place, providing the guidance and instructions necessary for maintenance and spare parts programs. Further, they state that their Design Control Program provides instructions and procedural controls to insure the appropriate technical, administrative and quality reviews. The licensee /aoplicant has described the use of the information handling system in determining whet,c e an activity is safety-relatec and in determining wnat crocecures are used for purchasing, operation, surveillance testing and maintenance activities.
5.3 Conclusion We fina that the utility's description of plant administrative controls anc procecures meets the recuiremerts of this item arc is, tne-efore, acceptable.
Seabrook SSER 8 6 Appendix W
- 7. ITEM 2.2.1.4 - MAf4AGEMENT CONTROLS 7.1 Guideline The applicant or licensee should confirm that the management controls used to verify that the procedures for preparation, validation and routine utilization of the information handling system have been followed.
7.2 Evaluation The utility states that administrative and quality assurance programs, as well as Station Manual procedures, ensure that station requirements are followed. They state that there are management controls governing the input to, the outputs of and process activities related to the info'rmation handling systen.. The controls include the Station Operational Review Committee for 10 CFR 50.59 reviews, and periodic Quality Assurance audits and surveillance of the information handling system. Comparisons to manual data bases are also utilized as a validation tool. Routine use of the computer data base is called for in procedures and in design change instructions.
7.3 Conclusion We find that the msnagement controls used by the utility assure that the information handling system is maintaineo, is current and is used as intended. Therefore, the licensee's/ applicant's response for this item is acceptable.
Seabrook SSER 8 7 Appendix W
l
- 8. ITEM 2.2.1.5 - DESIGN VERIFICATION AND PROCUREMENT
- l. 8.1 Guideline l
The applicant's or licensee's submittal should document that past l usage demonstrates that appropriate design verification and qualification testing is specified for the procurement of safety-related components and
! parts. The specifications should include qualification testing for expected safety service conditions and provide support for the applicant's/ licensee's receipt of testing documentation to support the limits of life recommended by the supplier. If sucn documentation is not available, confirmation that the present program meets these requirements should be provided.
8.2 Evaluation The utility states that the approved station criteria for Material Purchase Request Preparation and Review ensures that design verification and qualification testing is specified for the safety-related ecuipment and components procured. They also state that cesign and service conditions are identified as a result.
S.3 Conciusion Altnougn the utility dic not specify the design criteria acclied to tnis item, we conclude that they have accressec the concerns of tnis item anc, inerefore, we finc the resoonse for inis item acceptable.
Seabrook SSER 8 8 Appendix W
- 9. ITEM 2.2.1.6. "IMPORTANT TO SAFETY" COMPONENTS 9.1 Guideline Generic Letter 83-28 states that the utility's equipment classification prugram should include (in addition to the safety-related components) a broader class of components designated as "Important to Safety." However, since the generic letter does not require the utility to furnish this information as part of their response, review of this item will not be performed, l
Seabrook SSER 8 9 Appendix W
- 10. CONCLUSION Based on our review of the utility's response to the specific requirements of Item 2.2.1, we find that the information provided by the utility to resolve the concerns of Items 2.2.1.1, 2.2.1.2, 2.2.1.3, 2.2.1.4 and 2.2.1.5 meet the requirements of Generic Letter 83-28 and is acceptable. Item 2.2.1.6 was not reviewed as noted in Section 9.1. :
)
i i
Seabrook SSER 8 10 Appendix W
- 11. REFERENCES
- 1. NRC Letter, D. G. Eisenhut to all Licensees of Operating Reactors, Applicants for Operating License, and Holders of Constructf on Permits,
" Required Actions Based on Generic Implications of Salem ATWS-Events (Generic Letter 83-28)," July 8,1983.
- 2. Letter, Public Service Company of New Hampshire (J. DeVincentis) to NRC (G. W. Knighton), " Response to Generic Letter 83-28,"
November 4,1983, SBN-576, T. F. B4.2.99.
- 3. Letter, Public Service Company of New Hampshire (G. S. Thomas) to NRC,
" Additional Information for Items 2.1 (Part 2) and 2.2 (Part 1) of Generic Letter 83-28," May 4, 1987, NYN-87061.
1 Seabrook SSER 8 11 Appendix W
J APPENDIX X TECHNICAL EVALUATION OF CHANGES TO THE PUMP AND VALVE INSERVICE TESTING PROGRAM, SEABROOK STATION, UNIT 1 Seabrook SSER 8 Appendix X
TECHNICAL EVALUATION OF CHANGES TO THE PUMP AND VALVE INSERVICE TESTING PROGRAM, i SEABROOK STATION, UNIT 1 This report documents EG&G Idaho's review of Seabrook Station, Unit I, IST program revisions, as described and forwarded to the NRC by letters dated j
June 17, 1987, January 25, 1988, and September 26, 1988.
l Changes to the licensee's IST program were reviewed utilizing the acceptance criteria and guidance contained in the following documents: the i
ASME Code Section XI, 1983 Edition with addenda through Summer 1983 and Code interpretations when applicable, the Code of Federal Regulations 10CFR50, the Standard Review Plan, Section 3.9.6, and the Draft Regulatory Guide and Value/ Impact Statement titled, " Identification of Valves for Inclusion in Inservice Testing Programs".
The licensee's new Cold Shutdown Justification No. V-48, revised Relief Request P-2 and additional Relief Request No. P-4 were evaluated to determine if testing the affected components in accordance with the Code requirements would be impractical or would place an unreasonable burden on the licensee, and whether the licensee's proposed testing would provide a reasonable alternative to the Code requirements.
The licensee also made other minor changes to their IST program and these changes have been evaluated and determined to be in accordance with the Code requirements and/or the NRC staff positions.
A conference call was held on December 18, 1987, with Public Service of New Hampshire, NRC, and EG&G Inc., representatives to discuss the changes to the IST program which were considered to be unacceptable or where further information was necessary. The licensee's letter dated January 25, 1988, made changes to the IST program in response to the NRC positions and concerns expressed during this conference call.
Seabrook SSER 8 1 Appendix X
- 1. PUMP TESTING PROGRAM The Public Service of New Hampshire bases for requesting relief from the p=p testing requirements and the reviewers evaluation of these requests are summarized below.
J.1 All Pumos in the IST Procram 1.1.1 Measure Pumo Bearino Temperatures. Pumo Relief Reouest P-2 1.1.1.1 Relief Reouest. The licensee has requested relief from the requirements of Section XI, Paragraph IWP-3300, for the annual measurement of pump bearing temperature for all pumps in the IST program and proposed to expand the quarterly measurement of pump vibration amplitude from one to multiple readings in two orthogonal directions.
1.1.1.1.1 Licensee's Basis for Reouestino Relief--The referenced cdition of the Code requires bearing temperature to be recorded annually. It has been shown by experience that bearing temperature rise occurs only minutes prior to bearing failure. Therefore, the detection of possible bearing failure by a yearly temperature measurement is extremely unlikely.
It requires at least an hour of pump operation to achieve stable bearing temperatures. The small probability of detecting bearing failure by temperature measurement does not justify the additional pump operating time required to obtain the measurements. As an alternative, the pump vibration testing will be expanded form one to multiple readings in two orthogonal directions.
1.1.1.1.2 Evaluation--The licensee has indicated that a yearly )
measurement of pump bearing temperature for these pumps is not a meaningful test for detecting pump bearing degradation. There are several factors such )
as the temperature of the working fluid, the ambient temperature, and the lubricant temperature that would affect the measured bearing temperature and mask any bearing condition change short of a catastrophic bearing failure.
The Code required quarterly pump vibration measurements give a much more accurate indication of pump bearing condition than the temperature Seabrook SSER 8 2 Appendix X
measurement, and the vibration measurement is not substantially affected by any system parameter or other factor that could mask problems or result in erroneous indications of bearing degradation. A yearly bearing temperature measurc:::ent is impractical for these pumps because it would require operating
-the pumps for over an hour and would not provide any information-about pump bearing condition above that provided by the proposed multiple vibration measurements in two orthogonal directions. Yearly pump bearing temperature measurements are not justified by the limited information that would be provided about' pump mechanical condition.
Based on the impracticality of complying with the Code and the burden on
. the licensee if the Code requirements were imposed and considering the quarterly pump vibration measurements that will be taken to determine pump mechanical condition and to detect pump bearing degradation, relief may be granted from the Section XI requirement of annually measuring bearing temperature for these pumps.
1.2 Service Water Pumot 1.2.1 Measurement of Pumo Inlet Pressure. Pumo Relief Recuest P-4 1.2.1.1 Relief Recuest. The licensee has requested relief from the pump inlet pressure measurement and instrument accuracy requirements of Section XI, Paragraphs IWP-3100 and 4110, for the. service water pumps and proposed to calculate pump inlet pressure based on the water level above the pump inlet using the installed level instruments.
1.2.1.1.1 Licensee's Basis for Recuestina Relief--The service water pumps are vertical turbine pumps with no direct means to obtain the inlet pressure measurements as required by IWP-4200.
The inlet pressure will be calculated based on the water level above the pump inlet using existing plant instrumentation to measure pump suction pressure. Plant installed level instrumentation is accurate to i 0.5% which is within the requirements of Table IWP-4110-1, but total loop accuracy is i 2.5% which exceeds the requirements of Table IWP-4110-1.
Seabrook SSER 8 3 Appendix X
1.2.1.1.2 Evaluation--The service water pumps are vertical shaft deep draft pumps that are submerged in the sea water. The inlet pressure is due to the head of the sea water above the level of the pump inlet. The inlet pressure when the pump is operating cannot be determined because there are no installed inlet pressure instruments. However, any significant blockage in the pump suction would be indicated by a reduction in the pump flow rate. Calculation of the rump inlet pressure by measuring the water level above the pump suction will allow the licensee to determine the pump differential pressure. Using the calculated pump differential pressure in conjunction with the pump flowrate should provide adequate information to ascertain the hydraulic condition of the pump and to detect any pump hydraulic degradation.
A system modification would be necessary to allow direct measurement of ;
pump inlet pressure and the additional information provided would have a minimal impact on the licensee's ability to detect pump hydraulic '
degradation. Also, a system modification would be required to install instrumentation that meets the accuracy requirements of the Code, however, it would be burdensome to require the licensee to make these modifications considering the small improvement they would make in the licensee's ability to detect pump hydraulic degradation.
Based on the impracticality of these measurements, the burden on the licensee if these Code requirements were imposed, and the licensee's proposed alternate testing of measuring sea water level using existing instrumentation and calculating pump inlet pressure relief may be granted from the s
Section XI requirements as requested.
Seabrook SSER 8 4 Appendix X
- 2. VALVES I'ESTED DURING COLD SHUTDOWNS 2.1 Auxiliary Feedwater System 2.1.1 Cateaory C Valves. Cold Shutdown Justification V-48 Valves FW-V99, FW-V216, and FW-V357, the startup feed pump discharge check valves, and CO-V340, the startup feed pump suction check valve, cannot be full-stroke exercised quarterly during power operations because the only flow path through valves FW-V99, FW-V216, and FW-V357,and the only full flow path through valve CO-V340 is into the steam generators. Establishing full auxiliary feedwater flow into the steam generators during power operations could thermal shock the feed nozzles and result in their premature failure.
Valve CO-V340 will be partial-stroke exercised quarterly during power operations and all four valves will be full-stroke exercised during cold shutdowns and refueling outages.
Seabrook SSER 8 5 Appendix X
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4 HECIPIE NT S ACCESSION NUMBE6 3 TITLE AND SuST TLE Safety Evaluation Report related to the operation of 6 oA ama' wMP'e'to Seabrook Station, Unit 1 MAY 1989 e AuTHOHtSI 7 DATE REPORT ISSUED MONT.4 VEAR !
MAY 1989 f 9 PROJE CT IT ASK' WORK UNs1 NUMBEM S PE HF OHMING ORGAN 61 AIION NAME AND MAILING ADDHESS finclue Esp CpWJ Division of Reactor Projects I/II Office of Nuclear Reactor Regulation io > >N NuMn a U. S. Nuclear Regulatory Commission Washington, D. C. 20555 .
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j ?e TYPE OF REPOHT l 11 SPONSORING ORGANIZATION NAME AND MAIL 4NG ADDHESS (inctus g,p Cog)
Same as 8. above Technical I70 PEHsOD COVERED (Jaciusswe astest October 1987 - May 1989 13 SUPPL.EMtNT ARY NOTE S Docket No. 50-443 14 AdSTH ACT (2fW words or Mai Supplement No. 8 to the Safety Evaluation Report related to operation of the Seabrook Station, Unit 1 addresses items relating to the issuance of a 5% low pcwer license.
The report relates to the application filed by the Public Service Company of New Hampshire for a license to operate the Seabrook Station, Unit 1 located in Rockingham County, New Hampshire.
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