ML20210B004

Safety Evaluation Report Related to the FULL-TERM Operating License for Millstone Nuclear Power Station,Unit No. 1. Docket No. 50-245.(Northeast Nuclear Energy Company)

ML20210B004
Person / Time
Site:	Millstone
Issue date:	08/31/1986
From:	James Shea Office of Nuclear Reactor Regulation
To:
References
NUREG-1143, NUREG-1143-S01, NUREG-1143-S1, NUDOCS 8609170421
Download: ML20210B004 (39)

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NUREG-1143 Supplement No.1 l

l Safety Evaluation Report related to the full-term operating license for Millstone Nuclear Power Station, l

Unit No.1 Docket No. 50-245 Northeast Nuclear Energy Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1986 p re "%,

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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room,1717 H Street, N.W.

Washington, DC 20555

2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7082

3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federa: Regulations, and Nuclear Regulatory Commission issuances. ,

Documents available from the National Technical Information Service include NUREG series

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Documents 'available from public and special technical libraries include all open literature items, such as books, joumal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.

. Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.

l NUREG-1143 Supplement No.1 Safety Evaluation Report related to the full-term operating license for Millstone Nuclear Power Station, Unit No.1 Docket No. 50-245 Northeast Nuclear Energy Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1986 p= =m ,

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ABSTRACT This report, prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission, supplements the Safety Evaluation Report (NUREG-1143, October 1985). It fulfills a commitment to provide the Advisory Committee on Reactor Safeguards report, identifies the changes that have oc-curred since the Safety Evaluation Report was issued, and specifies the effec-tive lifetime for the Full-Term Operating License.

-Millstone Unit 1 SSER 1 iii

TABLE OF CONTENTS Page ABSTRACT.............................................................. iii 1

INTRODUCTION..................................................... 1-1

1. 3 Operating Experience........................................ 1-3 1.3.1 Design Confirmation.................................. 1-3 1.5 License Changes............................................. 1-4 1.5.1 Special Nuclear Materia 1............................. 1-5 1.5.2 Byproduct, Source and Special Nuclear Material As Sealed 1.5.3 Reports and Sources....................................... 1-5 Records..................................

1.5.4 Fire Protection......................................

1-5 1-5 3 DESIGN CRITERIA - STRUCTURES, SYSTEMS AND COMPONENTS............. 3-1 3.5 Missile Protection.......................................... 3-1 3.5.3 Turbine Missiles. ................................... 3-1 3.7 Environmental Qualification of Electric Equipment Important to Safety-Dr Profile...............ywell......................................

Accident Temperature 3-1 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS..................... 5-1 5.3 Integrity of Reactor Coolant Pressure Boundary. . . . . . . . . . . . . . 5-1 5.3.1 Inservice Testing.................................... 5-1 5.3.2 Reinspection, Analysis and Repairs of the Reactor Coolant System Piping........................ 5-1 5.4 Reactor Vesse1 RTNDT''''''''''''''''''''''''''''..... 5-2 6 ENGINEERED SAFETY. FEATURES (ESFs)................................ 6-1 6.2 Containment Systems......................................... 6-1 6.2.3 Combustible Gas Control - Hydrogen Monitor........... 6-1 8 ELECTRIC POWER SYSTEMS........................................... 8-1 8.1 Potential Equipment Failures Associated with Degraded Grid Voltage................................................ 8-1 Millstone Unit 1 SSER'l v

b Table of Contents (continued) 9 AUXILIARY AND EMERGENCY SYSTEMS.................................. 9-1 9.4 Fire Protection............................................. 9-1 9.4.4 Title 10 Code of Federal Regulations, Part 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979...................................... 9-1 18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS. ... ... .. . . IS-1 l 22 CONCLUSION....................................................... 22-1 APPENDICES APPENDIX A REFERENCES APPENDIX C UNRESOLVED SAFETY ISSUE APPENDIX E REPORT 0F THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ON FULL-TERM OPERATING LICENSE FOR MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1

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1 INTRGOUCTION In October 1985, the Nuclear Regulatory Commission staff (NRC or staff) issued a Safety Evaluation Report (SER) on the Millstene Nuclear Power Station, Unit No.1 plant (MP-1), NUREG-1143,1 related to the Full-Term Operating License (FTOL) application of Northeast Nuclear Energy Company (NNECO or the licensee) for ccnversion from a Provisional Operating License (POL) to a standard Facil-ity Operating License. By letter dated February 14, 1986,33 the licensee with-drew an earlier request for a 40 year FTOL dating from the issuance of the POL and reiterated the original request (NUREG-1143 reference 7) for a 40 year FTOL beginning with the authorization for construction, i.e., an expiration date of May 19, 2006.

This document is the first supplement to that SER (SSER 1). This supplement provides updated information available through August 1986, and includes the report of the Advisory Committee on Reactor Safeguards (ACRS) (see Appendix E).

There are a number of ongoing actions for MP-1 that are currently under_ staff review as noted in NUREG-1143. The staff has determined that these items do not require resolution before the issuance of an FTOL and should not delay the POL-FTOL conversion process. These items will be addressed as routine operating reactor licensing actions after the FTOL is issued.

Amendment 107 to the Provisional Operating License,2 dated December 6, 1985, is responsive to the licensee application dated August 26, 1985, to refuel the core with 200 fuel assemblies identical to those used in_the April 1984 refuel-ing outage. The refueling outage for Reload No. 10 spanned the period from October 25 to December 21, 1985. (Other amendments to the license are listed in Appendix A references 3, 4, and 5.)

Each section of this supplement is designated the same as the related portion of the SER, NUREG-1143.

In a related activity, the licensee proposed in letters dated June 13, 1983, September 14, 1983, December 28, 1983, and May 17, 1985 to conduct an expanded integrated assessment for MP-1 that would address the remaining SEP issues (Category 3), all pending licensing actions, and significant licensee-sponsored plant improvements. This effort is referred to as the Integrated Safety Assessment Program (ISAP).

On November 15, 1984, the Commission published a policy statement in the Federal Register (49 FR 45112) which describes the elements and objectives of ISAP, as a regulatory vehicle to develop plant-specific, integrated implementation schedules for plant modifications. MP-1 has been selected as one of two plants to participate in an ISAP pilot program. Consequently, the results of the li-censee's additional analyses for several of the issues discussed in this report refer to ISAP, where alternative corrective actions will be considered and a prioritized implementation schedule for any plant modifications will be de-veloped. The general scope of the ISAP review for Millstone 1 is presented in a July 31, 1985, letter, to the licensee. Also, in that letter the staff Millstone Unit 1 SSER 1 1-1 b

requested that a safety analysis summary and a Probabilistic Safety Study (PSS)  :

summary, as appropriate, for each topic be submitted for review. By letters dated August 13 through November 25, 1985, the requested safety analyses and PSS summaries were submitted by the licensee. In addition, supplemental infor-mation for several ISAP topics was submitted through February 4, 1986.

The staff' reviewed each of the topic submittals to determine if the content of the detailed topic scope was appropriate and to determine if any of the topics, '

based upon the information submitted, could be resolved in whole or in part at this time. The staff also reviewed the MP-1 PSS and the plant operating history to identify any possible additional topics that would warrant consideration in the integrated assessment. The conclusions reached by the staff in their review of the topic scope and summaries, the PSS and the plant operating experience can be categorized as one of the following:

(1) The staff concurs that the topic scope as presented is appropriate for consideration in the integrated assessment. ,

(2) The staff has redefined the scope of the topic to reflect related topic ,

issues; ,

(3) The staff has developed a new topic for int usion in the integrated assess- l ment; or (4) The scope of the topic has been resolved and, therefore, does not need to be evaluated in the integrated assessment.

By letter dated March 3, 1986,8 the staff issued its review of the detailed scope of topics for the integrated assessment. The staff plans to meet with the li-censee to discuss the licensee's proposed corrective actions for the issues identified to support the preparation of a draft integrated assessment report.

Following a peer review and resolution of the corrective actions required, the licensee will be directed to submit its proposed implementation schedules so that the final integrated assessment review report and license amendment can be issued. .

In a letter dated July 31, 1986,7 NNECO submitted the integrated assessment proposal for MP-1. The objective of that proposal was to make a decision, based upon topic safety analyses, PSS summaries and a prioritization methodology, as to which topics warrant further action and which do not. The staff currently has this submittal under review a.1d is scheduled to issue its draft Integrated Safety Assessment Report in September 1986.

The staff has reviewed the issues being addressed under ISAP and concludes that none of them constitute an immediate undue hazard to the health and safety of the public. Therefore, the staff further concludes that these issues need not i be resolved prior to the issuance of a Full-Term Operating License. NNECO and ,

NRC representatives met in the NRC Offices in Bethesda, Maryland, on January 15, 1986, to review the timeliness, with respect to the expiration date for the FTOL, of the original application to convert the Provisional Operating License for MP-1 to a Full-Term Operating License, dated September 1, 1972 (Reference 7 -

NUREG-1143, Appendix A), as noticed in the Federal Register 37 FR 25187. The 1972 application requested an FTOL at its licensed power level of up to 2011 Millstone Unit 1 SSER 1 1-2

1 MWth for a period of 40 years from May 19, 1966, the issuance date of the con- l struction permit. On this basis the requested FTOL would expire on May 19,  !

2006. The licensee later requested, by letter dated January 15, 1982, an FTOL '

expiration date 40 years after the issuance of the initial operating license (i.e., October 2010), but later withdrew this request by letter dated Febru-ary 14, 1986. The staff concludes, therefore, that the FTOL expiration date should be May 19, 2006 as originally requested and noticed in the Federal Register.

The NRC Project Manager assigned to the FTOL review for MP-1 is Mr. James J.

Shea. Mr. Shea may be contacted by calling (301) 492-7639 or writing:

Mr. James J. Shea U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 1.3 Operating Experience On September 27, 1985, Hurricane Gloria moved through Connecticut with sustained winds up to 58 mph and gusts up to 75 mph at the Millstone site. By letter l dated December 31, 1985,8 the licensee submitted information to provide insights useful in the staff's ongoing efforts to resolve Unresolved Safety Issue (USI)

A-44, " Station Blackout." (See Appendix C, Update Station Blackout - Task A-44.)

1.3.1 Design Confirmation By letter dated December 23, 1985,8 the licensee reported the discovery of a fully severed "Hilti Kwik Bolt" on one of the main steam line support base plates during a plant housekeeping walkdown, prior to returning to power opera-tion after reload outage No. 10. As a result of this damage, a comprehensive inspection was carried out. This inspection showed that seven of twenty-four bolts, on a total of four base plates for two supports, had failed. An inspec-tion of the failure surface indicated overloading as the cause of bolt failure.

The bolts which failed were tested and found to demonstrate the anticipated properties of the AISI 1144 material. The licensee conc!uded that the bolts did not have any material deficiencies that contributed to this failure.

The design calculation for the damaged supports was reviewed by the licensee and found to fully comply with the requirements of AISC and I&E Bulletin 79-02.

A review of the piping and support analysis indicates that normal operating conditions could not have generated sufficient load to cause the damage observed.

The licensee believes that the two main steam lines have been subjected to tran-sient loadings which damaged the supports and caused excessive pipe movement.

This movement was evidenced in damaged pipe insulation that impacted structures

• adjacent to the main steam lines.

In order to assess the impact of this event on system integrity, several steps i were taken. The initial step was a visual inspection of all the other pipe supports on the portion'of the main steam system where the damage was found.

The results of inspections conducted revealed no damage that would affect the function or operability of the supports.

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The second step involved the analysis of the main steam lines utilizing observoa displacements as input. The peak stress locations predicted by this analysis were selected for magnetic particle inspection. This magnetic particle inspec-tion of the pipe outside diameter where the maximum stress occurs addresses pressure boundary integrity. This inspection revealed no service-related indications.

l A third step involved the magnetic particle inspection of the welded attach-ment to the pressure boundary on the supports which sustained damage. This magnetic particle inspection revealed no indications of damage.

The damaged supports which serve as seismic restraints have been redesigned.

The redesign utilizes a postulated load derived from the observed failure. The resulting design satisfies all criteria of AISC, B31.1 and I&E Bulletin 79-02.

The design is such that at ultimate capacity the support will fail prior to the attachment to the pressure boundary, thus assuring its continued integrity through potential transients.

The redesigned supports have been instrumented, along with the corresponding supports on the adjacent main steam lines. The instrumentation consists of strain gages located on structural members in positions allowing determination of maximum applied loadings and directions of loading. This instrumentation will allow verification of normal operating loads and identify any operating transient load cases.

The staff concluded the corrective actions are adequate to assure continued safe operation of the plant.

However, since there is no conclusive determination of the cause of these failures and the loading condition experienced appears to have far exceeded the original design basis, the licensee was requested by letter dated January 2,1986 to provide:

(1) The results of a review of the plant operating experience with regard to transients, or other events that could have possibly caused the observed damage.

(2) A figure of the main steam line support configuration, specifically identifying the location of the failed anchor bolts and observed pipe i

displacement.

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! (3) The calculation of the loading required to cause the observed failures.

(4) The data from the special monitoring instrumentation during the plant startup from the recent refueling outage and a description of plans for periodically reviewing this data in the future.

The licensee responded to the NRC request by letter dated February 6, 1986,"

1.5 License Changes The Facility Operating License (f0L) for Millstone Nuclear Power Station, Unit No. 1, should be in accordance with the current Commission policy and, therefore, similar to the Facility Operating Licenses for the most recently i

Millstone Unit 1 SSER 1 1-4 l

licensed nuclear power plants. Since the plant has operated successfully for more than 15 years, the staff has concluded that item IC of the Provisional Operating License (POL) that requires actual operating experience before issu-ance of a full term operating license has been satisfied, The staff has also concluded that the specific refennces to initial fuel loading, installation and testing of the diesel generator, and reactor criti-cality in POL item 10 are no longer relevant since during the past 15 years the reactor has been refueled 10 times (most recently October-December 1985) and the emergency diesel generator has been operable over the same period in compliance with the technical specifications.

1.5.1 Special Nuclear Material The proposed FOL item 28(2) restricts the amount of special nuclear material (i.e. reactor fuel) to the storage limits provided at the facility and the amounts required for reacter operation as described in the Final Safety Analy-sis Report, as supplemented and amended. POL item 28(2)(b) imposes a limit of 4600 kilograms of contained uranium-235 as reactor fuel, 5.5 grams of ura-nium-235 contained in monitoring systems and 0.5 microcurie of plutonium-239 as sealed discs. The staff has reviewed and approved all changes to the nuclear fuel storage (new and spent fuel assemblies) facilities and fuel assemblies since initial operation and has concluded that the quantities of fissionable material specified in the POL are somewhat arbitrary and unjustifiably restric-tive. The staff has determined that the original intent, i.e. to limit the amount of new fuel stored at the facility for refueling and the amount used in the core, will be satisfied by the proposea' FOL item 2B(2).

1.5.2 Byprcdoct, Source and Special Nuclear Material As Sealed Sources The POL identified arbitrary quantitative limits, in item 2B(2)(c), for the amounts of byproduct material 'such as cobalt, tritium, americium, antimony, strontiun and cesium. These absolute liraits should be replaced (FOL. item 283) by a restriction applicable to source and special nuclear material, as well as byproduct material. The new license requirement should eliminate quantitative Ifmits and be similar to the restrictions that have been included in new plant FOLs for the past several years. The staff has concluded that Parts 30, 40, and 70 of the Code of Federal Regulations, Title 10, Chapter I, adequately describe the license conditions for the receipt, possession and use of these materials at the facility and this change, therefore, is acceptable.

1.5.3 Reports and Records The POL items relating to reports and operating records, POL items 2C(d) and (5), are adequately addressed in the currently approved technical specifica-tions for Mi11stene Unit 1 operation. Accordingly the staff has concluded that these items are redundant and should not be included in the FOL.

1.5.4 Fire Protection POL item 2.C(6) " Fire Protection" should be changed, the staff has concluded, to assure that all fire protection features described by the licensee and ap-proved by the NRC are maintained in effect, that no changes are made that could ,

adversely affect the ability to achieve and maintain safe shutdown in the event Millstone Unit 1 SSER 1 1-5 I s

of a fire, without prior staff approval, that records are maintained in auditable form for any changes that are made, and that all changes are reported annually to the NRC. The staff also concluded that the paragraph relating to administra-tive controls should be omitted in the new FOL because it merely refers to a section of the technical specifications and the specified completion date has been satisfied.

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3 DESIGN CRITERIA - STRUCTURES, SYSTEMS AND COMPONENTS 3.5 Missile Protection 3.5.3 Turbine Missiles (SEP Topic III-4.B)

By letter dated December 13, 1985,12 the licensee informed the staff that during the Reload 10 outage (October-December 1985) inspection of the "0" low prescure turbine, stress corrosion cracking was identified above the keyway of the num-ber 4 generator - end. wheel. The 1.4-inch measured crack indication was sp-proximately 1.2 inches greater than the same indication measure during the 1980 refueling outage. By letter dated December 20, 1985,13 the licensee, in re^

sponse to a staff request, provided a representative cross-section view of the "B" low pressure turbine indication that also showed the flaw size assumed by General Electric (G.E.) to calculate the safe operating interval before the next repeat turbine inspection. The results of the most recent calculations made by G.E. , in accordance with approved NRC methods, have reduced the earlier inspection interval of 2.5 years calculated by deterministic methods (2.1 years calculated by probabilistic methods) to 1.2 years. Since 1.2 years is less than the normal 1.5 year (18 months) operating cycle for MP-1, the licensee intends to further evaluate the possibility of extending the 1,2 year inspec-tion interval to coincide with the normal refueling cutage scheduled for August 1987. Based on this information, the staff allowed return to normal power operation late in December 1985. An inspection of the "B" low pressure turbine was conducted during a May 1986 outage. No increase in flaw size was noted.

3. 7 Environmental Qualification of Electrical Equipment Important to Safety -

Erywell Accident Temperature Profile ,

The licensee, in response to a comnitment (NUREG-1143, Ref. 23) submitted by letter dated December 31, 1985," the results of containment drywell accident temperature calculations. A plant-specific analysis was performect by G.E. and consisted of the generation of mass and energy releases and the subsequent development of a plant-specific containment model. Composite temperature pro-files were determined for the drywell airspace and the drywell wall based on steamline breaks inside the drywell of 0.012 ft , 0.1 ft2 , and 0.5 ft2. Steam-line breaks were considered because they result in higher drywell temperatures than liquid breaks. The maximum drywell air / nitrogen space temperature pre-dicted is 330'F for approximately 3 minutes.

The use of drywell spray to cool the air /N2 space was not considered in any of the calculations. By separate letter to Commission Chairman N. J. Palladino dated January 6, 1986,15 the licensee stated that it is extremely unlikely that drywell sprays would ever be activated at MP-1 even for accidents beyond the design basis. Thisconclusionisreachedbecause,atlowdrpellpressure (10 psig), the drywell air temperature must be less than 180 F and at the same time the torus air /N2 temperature must be less than 90 F in order to be able to activate the drywell sprays. At a higher drywell pressure (50 psig), the dry-well air temperature must be less than 120 F according to the emergency Millstone Unit 1 SSER 1 3-1

operating procedures. Drywell spray use is strictly controlled as specified in the emergency operating procedures due to containment structural integrity Concerns.

All scenarios involve conditions where drywell pressure is elevated (greater than 10 psig) and drywell temperature is greiter than 250*F. In such a regime, drywell sprays cannot be operated because of concerns related to structural integrity (e.g., drywell collapse due to rapid steam conder.sation and the re-sultant pressure gradiant between the reactor building and drywell compartments).

1ha December 31, 1985 licensee letter notes that qualification documentation for all electrical equipment located in the drywell and governed by 10 CFR 50.49 was reviewed against the new composite drywell temperature profiles and all such equipment has been demonstrated to be qualified for the new composite pro-files. In addition, all documentation files have been revised to incorporate the new composite profiles.

The NRC staff did not identify any deficiencies related to the qualification of electrical equipment inside the drywell due to the new composite temperature profiles during the November 22, 1985 audit inspection, but the staff evaluation of the adequacy of the limiting accident temperature profiles is not complete at this time. It is expected that the staff safety evaluation of the licensee submittal will be completed and a report issued in the near future.

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5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.3 Integrity of Reactor Coolant Pressure Boundary 5.3.1 Inservice Testing By letter dated September 26, 1985,18 the licensee responded to the NRC Draft In-Service Testing (IST) SER (NRC letter dated May 22,1985). Revised exerp-tion requests are included within Enclosure 1 (Attachment 2 of the letter), a revised 10 year (1980-1990) In-Service Pump and Valve Test Program (Revision 2) that incorporates the comments, clarifications, and corrections resulting from review of the NRC Draft SER. Also included, responsive to staff requirements,-

is a revised list of valves to be tested in the IST program. The revised list includes the valves in:

Hydrogen and Oxygen Analyzer System

- Post Accident Sampling System Traversing In-Core Probe System

- Drywell Nitrogen Supply System The report notes.that the test program has been revised to meet, as nearly as possible, the provisions of the ASME Boiler and Pressure Vessel Code,Section XI, 1980 Edition, including the 1980 Winter Addenda, and to supersede the program contained in the January 16, 1985 NNECO submittal, and that implementing proce--

dures are being revised to reflect this latest revision to the test program.

NNEC0 believes that the current classification and testing program for the sys- .

tems addressed in the test prtgram is more than adequate to ensure a continued high reliability of these systems, while also ensuring that operation of the unit will not endanger life, property, or the common defense and security of the public.

The staff evaluation of this program is scheduled for completion by March 1987.

5.3.2 Reinspection, Analysis and Repairs of the Reactor Coolant System Piping By letter dated July 1, 1985,17 the licensee presented the inspection and inter-granular stress corrosion cracking (IGSCC) mitigation history and other infor-mation that included the basis for selection of welds for inspection during the October-December 1985 (Reload 10) refueling outage. The staff responded by letter dated August 14, 1985,1s noting that 111 piping welds in the reactor coolant stainless stcol piping were to be ultrasonically inspected. Staff members met with NNEC0 representatives in Bethesda, Maryland, on December 3, 1985, for a preliminary review of the October-December 1985 (Reload 10) outage IGSCC inspection results. The inspection program included a thorough 1.nspec-tion of 115 susceptible welds, weld overlays (5 in the jet pump instrument B

nozzle assemblies, letters dated December 1 in 12,the isolation 1985,18 condenser),

January andand 13, 1986,20 supporting Februaryanalyses.12,1986,y1 the licensee documented the information reviewed at the meeting; i.e., the list of welds inspected for IGSCC at MP-1 during the Reload 10 outage, dispositions Millstone Unit 1 SSER 1 5-1 L - . _. - - . -

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of the flawed welds, dimensional data for the six welds that were repaired by weld overlay, and responses to staff requests. The licensee concluded that:

- The 1985 IGSCC program exceeds the provisions of the ASME Boiler and Pressure Vessel Code and the guidelines of NUREG-1061 because the percentages of welds, per pipe sire, inspected meet or exceed NUREG-1061 guidelines.

The thorough inspection program using recently requalified personnel and equipment (including ultrasonic imaging and " master / slave" scan-ning) give a high degree of confidence that the welds inspected do not contain IGSCC.

Where IGSCC has been mitigated by weld overlays, the design implemen-tation and inspection procedures result in a safe weld joint.

No unreviewed safety questions currently exist with respect to IGSCC at Millstone Unit No. 1.

• The flawed joint accepted by analysis (RCAJ-1).and treated by induc-tion heating stress improvement (IHSI) in 1984 showed no evidence of crack growth. This result serves to assure the validity of the con-clusions in the original analysis, thereby providing justification for operation for another fuel cycle.

On the basis of the information presented by the licensee, the staff allowed (Amendment 107 dated December 6 1985) MP-1 return to full power for the next 18monthfuelcycle, cycle 11(l.e.,scheduleduntilAugust1987).

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5.4 Reactor Vessel Reference Temperature Nil Ductility Transition - RTNDT The April 12, 1985 NNECO letter referenced in NUREG-1143 presented the GE evalu-ation of reactor vessel test specimens removed from the MP-1 reactor vessel in May 1984 at the end of fuel cycle 9 operation. By letter dated December 17, 1985, the licensee reported that these test results indicate that the pressure /

temperature curves of the operating technical specifications, Figuras 3.6.1, 3.6.2, and 3.6.3 are valid to 8.6x108 MWDth or 11.7 Equivalent Full Power Years (EFPY) instead of 9x108 MW0th or 12.26 EFPY as previously projected. The existing technical specifications are valid for ah aoditional 1.6 EFPY. However, the licensee has committed to evaluate the limiting reactor vessel pressure /

temperature data provided by G.E. and propose changes to the technical scecifi-cations by September 15, 1986, approximately one year before new limits,' based on the surveillance specimen test results and reactor oper& ting history to date, must be imposed. The staff considers this schedule acceptable.

Millstone Unit 1 SSER 1 5-2

6 ENGINEERED SAFETY FEATURES (ESFs) 6.2 Containment Systems 6.2.3 Combustible Gas Control - Hydrogen Monitor By letters dated November 25, 1985, and July 31, 1986,23 the licensee submitted additional information relating to Beyond DBA hydrogen monitoring capability Hydrogen Monitoring System (HMS) Reliability and Operability requirements Emergancy Procedure Guidelines relating to the HMS, and

+ Hydrogen Generation risk considering deinerting and reinerting periods.

Each of the above items was addressed by the lice?see in the following manner:

(1) Beyond DBA Hydrogen Monitoring Capability As noted in NUREG-1143, the staff has concluded that an inerted containment at MP-1 is sufficient by itself to preclude the formation of a flammable gas mixture for an indefinite period of time following a postulated DBA LOCA. As long as the oxygen concentration is below about 5 percent, there is no possibility of combustion or c6nficaration. The oxygen concentration in the MP-1 containment is maintained, in accordance with the Technical Specifications, below 4% by volume dur'ing plant operation. To meet this requirement, oxygen ccncentration is normally inaintained well below this value. After a pcstulated destgri-basis acetdent, the oxygen concentration in containinent would be diluted by the production of hydrogen generated by the metal-water reaction. The licensee's analysis was limited to DBA's

only.

(2) Hydrogen Monitoring System Reliability and Operability Requirements The nydrogen monitor was declared operational on December 31, 1984. Since '

then, typical problems associated with the operation of a new instrument have occurr4d, The overall reliability of the instrument is still being evaluated, and 6dditional information on this subjec*, will be provided by the licensee.

Tite fronitor is used during corma) operation to determine the hydrogen and oxygen concentration in the drywell and is operated ctntinnopsly. Techni-cal Specification 4.7. A.6 addresst:s maximoin allowable oxygen concentrbtien and minimua permissibic oxygen samp!ing frequency. Tha operability require-

{ cent for hydrogen monitoring parallels that specified for oxygen monitoring.

Existing plant surveillance practices provide greater than 95% availability during plant conditions when the monitor may be necessary, based on typ fcal time reqqired to p?rform system 11cearity checks. The monitor could be returned to service, if required, during performance of these linearity Millston.e Unit 155ER 1 6-1 P

- - . _w - - - - -

checks within 30 - 60 minutes of establishing the need to return the moni-tor to service. The 30.60 minutes reflects the time period from realign-ment of the monitor to the " SAMPLE" mode until monitor output has stabi-lized. Since this monitor is continuously operating, operability of this monitor is essentially assured following an accident.

(3) Emergency Procedure Guidelines Relating to the Hydrogen Monitoring System Revision 4 to the Boiling Water Reactor Owner Group Emergency Procedure Guidelines (EPGs) has recently been developed and pertains, in part, to the hydrogen and oxygen concentrations in the drywell and suppression chamber. This revision has not yet been implemented and is currently being evaluated to determine the extent of applicability for the design of MP-1. Furthee information regarding the applicability of those por-tions of Revision 4 to the EPGs related to hydrogen and oxygen control will be provided by the licensee.

(4) Hydrogen Generation Risk Considering Deinerting and Reinerting Periods The probability of a large-break loss of coolant accident for Millstone Unit No. 1 is roughly 10 4 per year. Although there is no maximum cumu-lative time limit for deinerting periods, it is an infrequent occurrence conservatively estimated at 96 deinerted hours per year, while conforming to limiting-Technical Specification action statements. The probability of these two events occurring simultaneously is, therefore, estimated at approximately 10 8 per year, and the probability of these events occurr-ing with a simultaneous failure of the hydrogen monitor is even lower.

NNECO concludes that the expense of installing a redundant hydrogen moni-tor solely for the deinerted periods cannot be justified.

The staff will continue the evaluation of " Post Accident Hydrogen Monitors" for MP-1 in item ISAP Topic 1.11.

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Millstone Unit 1 SSER 1 6-2 I

l 8 ELECTRIC POWER SYSTEMS 8.1 Potential Equipment Failures Associated With Degraded Grid Voltage The staff evaluation enclosed with license Amendment 98 dated June 14, 1984 (NUREG-1143), noted the licensee commitment to complete the installation of the degraded grid protection system not later than the October-December 1985 re-fueling outage. The degraded grid modifications are required to protect the Class 1E equipment from sustained degraded voltages of the offsite power sys-tem. By letter dated November 14, 1985,24 the licensee proposed a further delay in implementation of the degraded grid voltage modifications. An inte-grated safety evaluation by the licensee showed that the chance of a station blackout following loss of the switchyard would be approximately 2.4 times greater with the planned design change than with the existing undervoltage (70% and 90%) protection. The increase in probability is referenced to the base' conditions as analyzed in the Probabilistic Safety Study (PSS) for MP-1 (dated July 1985). The application of Probabilistic Risk Assessment (PRA) methods in the integrated safety evaluation of the Degraded Grid Protection planned modifications was one of the first such applications of the MP-1 plant-specific PSS by the licensee. Since the circuit design change, based on nor-mally accepted methods and procedures, was found to have an unintentional adverse impact that outweighs the intended improvement, the final implementa-tion was not completed during the 1985 outage as planned. However, it is expected that the redesign and prccurement effort can be completed in time to support the final implementation during the August 1987 refueling outage. The panels installed during the 1984 outage .sith tie-ins to alarms will be main-tained in their current configuration until final tie-ins are made in 1987.

Interim procedures will be used to respond to these alarms.

By letter dated December 12, 1985,2s the staff agreed that the possible increase in the probability of a station blackout is sufficient to justify a delay in completing the modifications to the next refueling outage. The delay allows sufficient time for redesign and procurement of materials. However, the licensee was requested to provide additional information to confirm that the adverse impact of completing the originally planned degraded grid protection modifications outweighs the intended improvements.

By letter dated January 13, 1986,2s the licensee provided a description of the probabilistic assessment that resulted in an increased probability of blackout and an analysis of the degraded grid protection modifications that would have contributed to this increase. An attachment to the letter addressed background, the original proposed change, analysis of the change, and conclusions as pre-sented below:

Background

The existing undervoltage protection logic is designed to monitor station AC voltage levels on the "high" side of the reserve station service transformer (RSST). In the event that voltage levels drop to 71% of the normal value (LNP),

undervoltage relays signal the combined S1 and S2 power train logic schemes to Millstone Unit 1 SSER 1 8-1 t _____________________

simultaneously isolate and load-shed all Class 1E buses, as well as to start the two emergency onsite generators. After the diesel and gas turbine powered gen-erators achieve their rated speeds, they are loaded onto each of their associated 4160V buses in order to restore AC power to emergency equipment loads. The undervoltage protection logic has the identical response for conditions of j degraded voltage (345 kV on RSST high side) combined with a loss-of-coolant accident.

Proposed Design Change j The proposed design change to the existing undervoltage protection scheme was

'as follows:

(1) Voltage detection would be relocated from the RSST "high" side to individual 4160V buses in the S1 and S2 power trains.

(2) The S1 and S2 power train logic schemes associated with the gas turbine and diesel generators respectively, would be separated so they can function independently.

(3) Automatic reinstatement of the load shed feature would be provided in order to alleviate manual load shedding, which is presently required prior to emergency generator restart after trip.

Actual modifications would have involved the installation of new undervoltage and degraded voltage relays along with changes to the existing LNP logic which would have been split into separate S1 and S2 divisions. The separation would have allowed a single S1 or S2 power train to respond to undervoltage conditions without affecting the other train. For example, the S1 train could experience an LNP or " degraded voltage with accident present" event and independently perform the necessary steps that are required to repower vital loads from the gas turbine generator. Unless the S2 train received indications of a similar condition, it would continue to remain connected to the offsite power supply without its protective logic being challenged.

Analysis Although the change would have removed logic interdependencies between the S1 and S2 power trains, it had the potential to degrade the plant response to a design-basis accident. This degradation is the direct result of the proposed modifications that would have been implemented to separate retained portions of the existing LNP logic. In separating the S1 and S2 power train logic, the level of redundancy in the area of propagating signals to actuate the load shed and emergency generator start functions would have been reduced. The net result would have been $1 and S2 power train unavailabilities increasing by factors of approximately 130% and 220%, respectively. The overall combined unavailability of the S1 and S2 trains would have increased by a factor of 240%.

As an example of how power train unavailability would have been increased by the proposed change, the following explanation for the S2 train is offered.

The existing undervoltage protection circuits incorporate combined S1 and S2 .

logic to control S2 equipment. LNP relays in either logic would produce a start signal for the diesel generator and cause the S2 bus 14F to be isolated from the offsite power supply. With the proposed logic scheme, there would Millstone Unit 1 SSER 1 8-2

]

have been a significant loss in control circuit redundancy for the S2 train as compared to the existing design. Single relays, instead of pairs, would have been used to cause isolation of bus 14F and transmit a start signal to the emer-gency diesel generator that results in a higher unavailability for the S2 power train. Coincident with this increase is a reordering of the dominant contribu-tors to S2 unavailability as new component failures surface. As a result, the S2 power train unavailability would be is increased by a factor of 220%.

Conclusions

-At Millstone Unit 1, a high margin of safety is maintained via a combination of:

redundant diverse emergency electrical trains high reliability features built into the design of the individual trains a surveillance program on the critical electrical system components to detect inoperable equipment technical specification LCOs that limit plant operation when portions of the electrical system are inoperable.

The originally proposed design change increases the probability'that both (S1 and S2) power trains will be unavailable by a' factor of 240%. This causes the frequency of a station blackout to increase by the same amount, since station blackout frequency is computed by multiplying the frequency for loss of offsite power by the probability that both power trains are unavailable.

As a result, the proposed design change would have a net effect of reducing the high reliability features built into the original design of the individual electrical trains, thus serving to reduce the margin of safety at Millstone Unit No. 1.

As part of the PRA assessment, the proposed changes for the S2 power train logic were further modified in order to perform sensitivity studies based on engineering insights:

(1) " doubling up" of LNP pulse circuit and load shed relays to increase circuit redundancy in the S2 logic.

(2) incorporation of a second parallel LNP relay in the diesel generator start circuitry which receives a signal from the $1 train logic.

The results of the PRA analyses of the above changes show that S2 unavailability is reduced to a value that is slightly less (a 1%) than that for the existing design. To complete the sensitivity study, a second variation on the proposed design was made by including all of the modifications from the first variation, plus additional redundancy in the LNP lockout circuitry. Quantification results for this design show a marginal improvement over the first variation. This is because further reductions in S2 unavailability through logic modifications are limited by failure of the emergency diesel generator which accounts for over 98% of the total S2 unavailability.

Millstone Unit 1 SSER 1 8-3 t____. _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ .__

The staff will continue the evaluation of this additional information in ISAP t Topic No. 1.25 included in the Final. Report for Millstone Unit No. 1.27 l

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1 Millstone Unit 1 SSER 1 8-4

-. -. a

9 AUXILIARY AND EMERGENCY SYSTEMS 9.4 Fire Protection 9.4.4 Title 10 Code of Federal Regulations, Part 50, Appendix _R, Fire Protec-tion Program for Nuclear Power Facilities Operating Prior to January 1, 1979 By letter dated November 6, 1985,28 the Commission granted five exemptions to the specific provisions of Section III.G of Appendix R to 10 CFR Part 50.

Subsection III.G specifies the separation, fire barrier, and suppression requirements where both trains for redundant safe shutdown components are located within the same fire area.

'The fire areas related to the five exemptions are:

(1) Main Control Room (Fire Area T-21)

(2) Turbine Building Reactor Feed Pump Area, Elevation 14'-6" (Fire Area T-5 B and C)

(3) Turbine Building Switchgear Area (T-19A), Elevation 34'-6" (4) Turbine Building Switchgear Area (T-19 C, D and E), Elevation 34'-6" (5) Reactor Building - Northeast, Elevation 42'-6" (Area R-19)

The staff in a draft safety evaluation, dated January 6, 1983, indicated its intention to grant exemptions in the following two areas:

(1) Turbine Building Cable Vault (Fire Area T-16)

(2) Switchgear Area (Fire-Area T-198)

The details are included in NUREG-1143 and the staff's evaluation accompanying the November 6, 19852s exemption from the requirements of Appendix R to 10 CFR 50, Section III.G.2.

As a result of a reevaluation of the MP-1 fire protection provisions, by letter dated November 21, 1985,29 the licensee submitted a request for eight additional exemptions (subsequently revised to six) from the requirements of Section III.G of Appendix R to 10 CFR 50 for the barriers and penetrations listed below:

(1) Lack of rated fire door and frame at (T-9/R-1).*

( 1 )

(2) Lack of a full enclosure around the Shutdown Cooling Pump Room.

• The code such as (T-9/R-1) indicates fire door 1 between Fire Areas T-9 and R-1. ( 1 )

Millstone Unit 1 SSER 1 9-1 i _ _ _ _ _ _ -

(3) Lack of fire damper in ventilation penetration at (T-1/T-7) through (T-1/T-7), and (T-1/T-9) through (T-1/T-9).

( A ) ( D ) ( A ) ( D )

Lack of rated fire door.and frame at (T-1/T-9) through

( 1 )

(T-1/T-9),

( 3 )

Lack of rated steel plate barrier at (T-1/T-9).

( A )

(4) Use of water curtain in the Switchgear Room.

(5) Lack of a 3-hour rated fire door at (T-2/T-9).

( 1 )

(6) Lack of fire detection and suppression in the area of the Unit 1/ Unit 2 -

power interconnect cable.

The licensee also determined that certain areas and zones should be incorporated into larger fire areas to; (1) Eliminate from consideration those areas and zones which.do not contain safe shutdown components, instruments or cables; and (2)' Minimize the nutter of modifications required to bring barriers into com-pliance with Appendix R; and (3) Reduce surveillance requirements by minimizing the number of Appendix R barriers and penetrations.

New fire areas were established and the corresponding fire boundaries were identified. The new fire areas are based on the safe shutdown equipment they contain or the potential hazard they present to areas containing safe shutdown equipment. This division results.in the optimal number of areas in term <. of maintaining the integrity of safe shutdown systems.

The licen',ee has initiated detailed design and construction of the proposed modifications in a good faith effort to comply with Appendix R. These modifi- }

cations will be implemented on a schedule in accordance with 10 CFR 50.48 and constr. tent with-the schedule for the fire protection modifications approved in the previous exemption ~s to Appendix R.

The licensee asserts that the proposed modifications and exemption requests identified obviate the previous modifications and exemptions for the Switchgear Areas (Fire Areas T-19A, T-19B, T-19C, D and E), the Turbine Hall (Fire Areas l T-5B, T-5C) and for the Reactor Building (Fire Areas R-2A, R-28, R-2C and R-20).

NNECO notes that approval of the previous exemptions does not affect the Novem-ber 21, 1985 NNECO submittal insofar as the previous modifications and exemptions f for the fire areas identified above are no longer required. They have been replaced with the proposed modifications and exemption requests in this submittal.

Millstone Unit 1 SSER 1 9-2

The staff discussed the November 21, 1985 submittal with NNEC0 representatives on May 14, 1986.31 Thefinalevaluationofthegendingexemptionrequests including later NNECO responses to NRC concerns 3 is currently seneduled for completion in October 1986. The schedule for the plant modifications required in the Commission's November 6, 1985 letter will be in accordance with 10 CFR 50.48.

f

-Millstone Unit 1 SSER 1 9-3

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18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS)

During its 308th meeting, December 5-7, 1985, the ACRS completed its review of the application by the NNECO (licensee) for conversion of the Provisional Oper-ating License for its Millstone Nuclear Power Station, Unit 1, to a full-term operating license. This application was considered also by the ACRS Subcommit-tee on Millstone Point, Units 1,'2 and 3, during a meeting at the plant site in Waterford, Connecticut on November 18-19, 1985. Transcripts of the full committee and subcommittee meetings are available from Ace Federal Reporters, Inc., 444 North Capitol Street, Washington, DC 20001 (Telephone [202] 347-3700).

Copies of the transcripts are also available for review at the NRC Public Docu-ment Room, 1717 H Street, NW, Washington, DC, and at the Waterford Connecticut' Public Library, 49 Rope Ferry Road, Waterford, Connecticut 06385.

~

A copy of the ACRS report on the FTOL for MP-1 is included as Appendix E to this SER supplement. The report states that the ACRS believes there is reasonable assurance that the Millstone Nuclear Power Station, Unit 1, can continue to be operated at power levels up to 2011 MWth under a full-term operating license without undue risk to the health and safety of the public.

Millstone Unit 1 SSER 1 18-1

22 CONCLUSION

.In accordance with 10 CFR 50.51, the staff concludes that the licensee's requestsa for an FTOL at the presently authorized power level of up to 2011 MWth with an expiration date of May 19, 2006 should be granted.

Millstone Unit 1 SSER 1 22-1

Appendix A CONTINUATION OF REFERENCES

1. NUREG-1143, " Safety Evaluation Report Related to the Full-Term Operating License for Millstone Nuclear Power Station, Unit No. 1," issued October 1985.

2. Amendment No. 107, " Reload 10," issued December 6, 1985.

3. Amendment No.108, " Administrative Procedures Relating to Working Hours,"

issued December 10, 1985.

4. Amendment No. 109, " Guard Training and Qualification Plan," issued Decem-ber 19, 1985.

5. Amendment No. 110, " Disarmed Control Rods," issued April 4, 1986.

6. NRC Letter to J. F. Opeka, dated March 3, 1986, " Safety Evaluation of the Detailed Scope of Issues for the Integrated Assessment Re: Millstone Nuclear Power Station, Unit No. 1."

7. NNECO Letter, Opeka to Grimes, dated July 31, 1986, " Integrated Safety Assessment Program Final Report for Millstone Unit No. 1."

8. NNECO Letter, Opeka to Denton, dated December 31, 1985, "Haddam Neck Plant-Millstone Nuclear Power Station, Units Nos.1, 2, and 3, Effects of Hur-ricane GLORIA."

9. NNECO Letter, Opeka to Grimes, dated December 23,-1985, " Millstone Nuclear Power Station, Unit No. 1 - Main Steam Line' Support Failures."

10. NRC Letter to J. F. Opeka, dated January 21,1986, " Main Steam Line Sup-port Failures - Millstone Unit 1."

11. NNEC0 Letter, Opeka to Grimes, dated February 6,1986, " Main Steam Line Support Failures."

l

12. NNECO Letter, Opeka to Grimes, dated December 13, 1985, " Millstone Nuclear Power Station, Unit No. 1 - Main Turbine Inspection Program."

13. NNECO Letter, Opeka to Grimes, dated December 20, 1985, " Millstone Nuclear j

Power Station, Unit No. 1 - Main Turbine Inspection Program."

14. NNECO Letter, Opeka to Grime!., dated December 31, 1985, " Millstone Nuc1 car Power Station, Unit No. 1 Environment,s1 Qualification of Electrical Equipment."

Millstone Unit 1 SSER 1 1 Appendix A

15. NNECO Letter, Opeka to Chairman Palladino, dated January 6, 1986, " Mill-stone Nuclear Power Station, Unit No. 1 Environmental Qualification of Electrical Equipment."

16. NNECO Letter, Opeka to Zwolinski, dated September 26, 1985, " Millstone Nuclear Power Station, Unit No. 1 Review of Oraft Safety Evaluation Report on Inservice Pump and Valve Test Program."

17. NNECO Letter, Opeka to Zwolinski, dated July 1,1985, " Millstone Nuclear Power Station, Unit No. 1 Recirculation Pipirg Inspection During Reload 10 Outage."

18. NRC Letter to J. F. Opeka, dated August 4, 1985, " Piping Inspection Plan for Millstone Nuclear Power Station, Unit No. 1."

19. NNECO Letter, Opeka to Grimes, dated December 12, 1985, " Millstone Nuclear Power Station, Unit No. 1 IGSCC Program - 1985 Inspections."

20. NNECO Letter, Opeka to Grimes, dated January 13, 1986, " Millstone Nuclear Power Station, Unit No. 1 IGSCC Program - 1985 Inspections."

l

21. NNECO Letter, Opeka to Grimes, dated February 12, 1986, " Millstone Nuclear Power Ststion, Unit No.1 IGSCC Program - 1985 Inspections."

22. NNECO Letter, Opeka to Grimes, dated November 25, 1985, " Millstone Nuclear Power Station, Unit No. 1 Integrated Safety Assessment Program." =

23. NNECO Letter, Opeka to Grimes, dated July 31, 1986, " Millstone Nuclear Power Station, Unit No.1: ISAP Topic 1.11 - Post-Accident Hydrogen Monitor." '

24. NNECO Letter, Opeka to Grimes, dated November 14, 1985, " Millstone Nuclear Power Station, Unit No. 1 Degraded Grid Protection for Class 1E Power Systems."

25. NRC Letter to J. F. Opeka, dated December 12, 1985, " Implementation of Degraded Grid Protection for Class 1E Power Systems - MP-1."

26. NNECO Letter, Opeka to Grimes, dated January 13, 1986, " Millstone Nuclear Power Station - Unit No. 1; Integrated Safety Assessment Program Supple-ment to ISAP Topic No. 1.25."

27. NNECO Letter, Opeka to Grimes, dated July 31,1986, " Millstone Unit No.1 -

Integrated Safety Assessment Program - Final Report for Millstone Unit -

No. 1." '

28. NRC' Letter to J. F. Opeka, dated November 6, 1985, " Exemption From Require-ments of Appendix R to 10 CFR Part 50, Section III G.2 - Re Millstone Nuclear Power Station, Unit No. 1."

l

29. NNECO Letter, Opeka to Grimes, dated November 21, 1985, " Millstone Nuclear Power Station, Unit No. 1 Fire Protection."

Millstone Unit 1 SSER 1 2 Appendix A e

I i

30. Summary NRC/NNEC0, January 27, 1986, Telephone Conference Call dated Feb-ruary 12,1986.

31. Summary of May 14, 1986 Meeting to Discuss the Status of Outstanding Fire Protection Review Issues, dated May 28, 1986.

32. NNEC0 Letter, Opeka to Grimes, dated May 19, 1986, " Millstone Nuclear Power Station, Unit No.1 - Additional Information Regarding Appendix R Exemp-tion Request."

33. NECC0 Letter, Opeka to Grimes, dated February 4, 1986, " Millstone Nuclear Power Station, Unit No. 1 Provisional Operating License Conversion."

Millstone Unit 1 SSER 1 3 Appendix A

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APPENDIX C NUCLEAR REGULATORY COMMISSION (NRC)

C. Unresolved Safety Issues C.3 Discussions of USIs as They Relate to Millstone Unit 1 Task A-44 Station Blackout By letter dated December 31, 1985, the licensee presented an assessment of the events at the Millstone Nuclear Power Station related to Hurricane Gloria. The licensee's intent was to maximize the extent to which the insights derived from assessing these events can be appropriately factored into the staff's on going efforts to resolve Unresolved Safety Issue (USI) A-44, " Station Blackout." As reported in Licensee Event Reports (LER) 85-018-00 and 85-014-00, Millstone Unit Nos.1 and 2 experienced loss-of-offsite power (LOOP) as a result of Hurricane Gloria.

The following is abstracted from the licensee's submittal:

On September 26, 1985, the Millstone site hurricane action plan was implemented.

The hurricane action plan included a checkout of the Emergency Response Facili-ties (ERF), the selection of two Station Emergency Organization (SEO) shifts and successful testing of the emergency onsite AC power sources. At 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> on September 26, 1985, the National Weather Service declared a hurricane watch for Connecticut and the licensee's on-call emergency organizations for the Millstone site, the Haddam Neck Plant and the Corporate Emergency Operations Center were notified to report to duty stations at 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> on September 27, 1985.

By 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> on September 27, 1985,'all Millstone ERFs were manned and ready.

- At 0745 hours0.00862 days <br />0.207 hours <br />0.00123 weeks <br />2.834725e-4 months <br />, based on predictions that the storm would reach the Millstone site between 1600 and 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, the decision was made to bring all units at Millstone and Haddam Neck to a shutdown condition.

- At 0815 hours0.00943 days <br />0.226 hours <br />0.00135 weeks <br />3.101075e-4 months <br /> each corporate EOC Manager was requested to prepare t

contingency plans for dealing with a possible loss of communications and/or loss of offsite power sources for each unit.

j - At 0830 hours0.00961 days <br />0.231 hours <br />0.00137 weeks <br />3.15815e-4 months <br /> the Governor declared a " State of Emergency" in Connecticut.

- At Millstone, all power was secured to nonessential plant areas at 0843 hours0.00976 days <br />0.234 hours <br />0.00139 weeks <br />3.207615e-4 months <br />. The hurricane was being tracked east of New Jersey at 0845 hours0.00978 days <br />0.235 hours <br />0.0014 weeks <br />3.215225e-4 months <br /> and was moving north at 30 mph. The winds at Millstone were determined to be 29 mph (lower level) and 40 mph (upper level of approximately 142 feet) at 0915 hours0.0106 days <br />0.254 hours <br />0.00151 weeks <br />3.481575e-4 months <br />.

) Millstone Unit 1 SSER 1 1 Appendix C l

- Twelve-hour shift rotations were established for all Corporate E0C functions.

The National Weather Service issued a tornado watch for all of Connecticut for the hours between 1000 and 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />.

- In order to assure the availability of service water following the storm, preventative measures were taken to protect the integrity of the service water system during the storm.

The power level at 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br /> was 25% for Unit 1 and 43% for Unit 2.

- At 1040 hours0.012 days <br />0.289 hours <br />0.00172 weeks <br />3.9572e-4 months <br /> the Millstone meteorological conditions included 40 mph winds at the lower level and 50 mph at the upper level. The speed of the eye of the storm was estimated at 40 mph.

The Millstone Unit No. 1 gas turbine was successfully tested at 1045 hours0.0121 days <br />0.29 hours <br />0.00173 weeks <br />3.976225e-4 months <br />.

Millstone Unit No. 2 was taken off-line at 1112 hours0.0129 days <br />0.309 hours <br />0.00184 weeks <br />4.23116e-4 months <br />.

Millstone Unit No. I was taken off-line at 1140 hours0.0132 days <br />0.317 hours <br />0.00188 weeks <br />4.3377e-4 months <br />.

The Millstone E0F shifted to emergency power at 1216 hours0.0141 days <br />0.338 hours <br />0.00201 weeks <br />4.62688e-4 months <br />.

- At 1220 hours0.0141 days <br />0.339 hours <br />0.00202 weeks <br />4.6421e-4 months <br /> the winds at Millstone were determined to be 49 mph at the lower level and 57 mph at the upper level.

Millstone Unit No.' 2 shut down at 1227 hours0.0142 days <br />0.341 hours <br />0.00203 weeks <br />4.668735e-4 months <br />. Both units were shut down by 1255 hours0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.775275e-4 months <br />.

Details of LOOP Events At Millstone Unit Nos. I and 2, the first or " preferred" source of offsite power is supplied via each of the unit's reserve station service transformers (RSST). At Millstone Unit No. 1, an alternate source of offsite power is via the Flanders line, a distribution line originating in the Flanders Substation, approximately 5 miles from the Millstone site, and terminating at Millstone Unit No. 1. On September 27, 1985, at 1028 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91154e-4 months <br />, the Flanders line to Unit No. I was intermittently lost.

- At 1250 hours0.0145 days <br />0.347 hours <br />0.00207 weeks <br />4.75625e-4 months <br /> there was voltage fluctuation on the 345-kV line supplying the switchyard.

- At 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> the Millstone Unit No. 3 RSST was sparking. .

- At 1307 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.973135e-4 months <br /> Millstone Unit No. 2 was proceeding to natural circulation. l

- At 1317 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.011185e-4 months <br /> Millstone Unit No. 2 manually disconnected from the RSST. Both of the Unit No. 2 emergency diesel generators automati-cally started and loaded.

Millstone Unit 1 SSER 1 2 Appendix C

At 1334 hours0.0154 days <br />0.371 hours <br />0.00221 weeks <br />5.07587e-4 months <br /> Hillstone Unit No. 1 lost normal power and both the emergency diesel generator and the gas turbine autocatically started and loaded (additionally, both of the Unit No. 3 emergency diesel generators automatically started).

At 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> the Millstone lower level winds were 49 mph and the upper level winds were 59 mph.

Recovery From LOOP Events The Station Emergency Organization had been activated and in place since 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> on the morning of the storm to ensure that all actions taken were performed under a coordinated and planned effort. Extra personnel were kept at the station to provide assistance and all non-essential personnel were sent home well before the peak of the storm hit the area. A relief schedule was prepared and put into effect that provided for adequate relief for those remaining at the station during the storm.

Millstone Unit No. 1 The Millstone Switchyard was reenergized at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> on September 27, 1985.

The 23-kV Flanders line into Unit No. I was reenergized at 1705 hours0.0197 days <br />0.474 hours <br />0.00282 weeks <br />6.487525e-4 months <br /> on September 27, 1985; however, operators elected to stay on emergency AC power.

This decision was based on the excellent performance of all 3 units' energency AC power sources and the stable configuration of the plant. This allowed the unit to stay with the emergency power source until the RSST was energized via the switchyard. Unit No. 1 energized its RSST at 0910 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.46255e-4 months <br /> on September 28, 1985, following a complete washdown of switchyard and station insulators.

While the unit relied on onsite power for approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, offsite power, if needed, could have been restored via the Flanders line within 315 hours0.00365 days <br />0.0875 hours <br />5.208333e-4 weeks <br />1.198575e-4 months <br />.

The staff noted, during a November 14, 1985 Commission briefing on the resolution of USI A-44, that this decision allowed for power to be restored in a prudent fashion. It is possible that power could have been restored sooner, but the licensee took the time to clean the salt off the insulators in the switchyard and perform extensive tests on breakers before attempting to restore power.

Licensee actions were timely and appropriate.

The staff also indicated that NNEC0 had implemented corrective actions follow-ing a similar storm, Hurricane Belle, in 1976. Following Hurricane Belle, NNECO assessed the results of a lack of effective rainfall during a storm which would cause a buildup of salt spray in the switchyard. As a result of the assessment, NNECO:

Installed salt monitors in the switchyard and Installed new equipment in the switchyard to increase creep path, i.e., increase resistance to ground. Specifically, NNECO a) installed the largest commercially available glass insulators in the switchyard; b) replaced switchyard circuit breakers to provide better insulation capability; and c) replaced transformer bushings between the unit and the switchyard.

Millstone Unit 1 SSER 1 3 Appendix C

These modifications and precautionary actions taken prior to the event enabled NNECO to respond to the recent LOOP event at the Millstone units in a prudent, deliberate, and coordinated manner without jeopardizing the safety and health of either the public or company employees.

As noted by members of the Advisory Committee on Reactor Safeguards (ACRS) dur-ing the November 19, 1985 Subcommittee meeting in Waterford, Connecticut (refer-ence pages 207 through 209 of.the meeting transcript), the advance warning and actions taken prior to a severe storm arrival lead to conservatisms in a proba-bilistic risk assessment, and perhaps, these events should be categorized in a fashion different from other than LOOP. events. Prior to and during the reactor shutdown, precautionary steps were taken which included laying out supply hoses to bring alternate cooling water to a diesel generator or the instrument air compressors, installing sand bags around doorways, closing floodgate doors, and installing life lines between outdoor buildings to ensure personnel could move safely between buildings when necessary.

As the storm reached its peak, it became evident that, because of a lack of any effective rainfall, a heavy buildup of salt spray was taking place as evidenced by an increased frequency of arcing on outside transformers, switchyard trans-mission lines, and circuit breakers. Steps were taken to bring the units off-line. All the Millstone emergency onsite AC power sources successfully started and loaded and ran until prLdent plant actions were completed to allow for ros-toration of normal offsite power. If necessary, Millstone Unit No. 1 could have had offsite power restored within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Since more rapid restoration of offsite power was not vital, NNECO elected to pursue a more deliberate and thorough cleaning and checking restoration process. This approach was in the best interest of personnel safety of company employees.

The advance notification associated with severe weather events of this kind pormits advance precautionary actions not usually credited by the staff or in plant probabilistic safety studies.

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Millstone Unit 1 SSER 1 4 Appendix C

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• • • • • December 10, 1985 Honorable Nunzio J. Palladino Chaiman U. S. Nuclear Regulatory Comission Washington, D. C. 20555

SUBJECT:

ACRS REPORT ON A FULL-TERM OPERATING LICENSE FOR THE MILLSTONE NUCLEAR POWER STATION, UNIT 1 During its 308th meeting, December 5-7, 1985, the Advisory Comittee on Reactor Safeguards completed its review of the application by the Northeast huclear Energy Company (Licensee) for conversien of the provisional cperating license (POL) for its Millstone Nuclear Power Station, Unit I to a full-tenn operating license (FTOL). The ACRS Subcommittee on Millstone Point Units 1-3 toured Unit 1 on November 18, 1985 and met in Waterford, Conn., on November 18-19, 1985 to consider the application. During our review, we had the benefit of discussions with representatives of the Licensee and the NRC Staff.

We also had the benefit of the documents referenced. The Comittee most recently discussed and reported on this plant in a letter dated December 13, 1982 relating to the Systematic Evaluation Program (SEP) review of Millstone, Unit 1.

Millstone, Unit I received a POL in October 1970 and began comercial operaticn in December of the same year. The Licensee applied for an FTOL in 1972, but review of this application was deferred by the NRC Staff in 1975, along with several other FTOL reviews. In 1978, Millstone, Unit I was included in Phase II of the SEP because much of the review needed for the FTOL was similar in scope to that for the SEP.

In the Comittee's letter reporting on the results of the SEP as applied to Millstone, Unit 1, we indicated that our review of the FTOL would be deferred until the NRC Staff had completed its actions on the SEP issues that were still pending and on the Unresolved Safety Issues (USI) and TMI Action Plan items. The Safety Evaluation Report (SER) issues have been resolved to the satisfaction of the NRC Staff in the manner reported in Supplement I to the Integrated Plant Safety Assess-ment Report for Millstone, Unit 1. The status of the USI and TMI Action Plan items for Millstone, Unit I has been discussed by the NRC Staff in its SER related to the FTOL for Millstone, Unit 1. We believe that the procedures and schedules that have been agreed to for these items are satisfactory.

This is the first plant to use the Integrated Safety Assessment Program (ISAP) as a device for identifying and scheduling the resolu-tion of regulatory items. This program requires a plant-specific Millstone Unit 1 SSER 1 1 Appendix E l

Honorable Nunzio J. Palladino D2cemb2r 10, 1985 probabilistic safety analysis, as well as an evaluation of plant operating experience and licensee perfemance. The licensee.has taken an active and dedicated part in this endeavor and the program appears to be working well.

Our review of tFe nperating experience at Millstone, Unit 1 indicates that the plant is being satisfactorily operated. This conclusion is further supported by the recent Systematic Assessment of Licensee Performance reports for Millstone, Unit 1 in which all activities reviewed were classed in either Category 1 or 2.

The Committee believes that there is reasonable assurance that 'the Millstone Nuclear Power Station," Unit I can continue to be operated at power levels up to 2011 MWt under a full-tem operating license without undue risk to the health and safety of the public.

Sincerely, David A. Ward Chaiman

References:

1. Northeast Nuclear Energy Company, " Final Safety Analysis Report.

Facility Description and Safety Analysis Report, Millstone Nuclear Generating Station, Unit 1," Volumes 1-3

2. Northeast Nuclear Energy Company, " Application for Conversion from Provisional Operating License to Full-Tem Operating License, Millstone Nuclear Power Station, Unit 1," dated September 1, 1972

3. U. S. Nuclear Regulatory Comission, " Integrated Plant Safety Assessment. Systematic Evaluation Program, Millstone Nuclear Power Station, Unit 1," USNRC Report NUREG-0824, dated February 1983 and Supplement 1, dated November 1985

4. U. S. Nuclear Regulatory Comission, " Safety Evaluation Report Related to the Full-Tem Operating License for Millstone Nuclear g Power Station, Unit 1," USNRC Report NUREG-ll43, dated October 1985 Millstone Unit 1 SSER 1 2 Appendix E

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