ML20137L562

Safety Evaluation Report Related to the Operation of Millstone Nuclear Power Station,Unit No. 3.Docket No. 50-423.(Northeast Nuclear Energy Company)

ML20137L562
Person / Time
Site:	Millstone
Issue date:	11/30/1985
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-1031, NUREG-1031-S03, NUREG-1031-S3, NUDOCS 8512030475
Download: ML20137L562 (83)

Text

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NUREG-1031 Supplement No. 3 Safety Evaluation Report related to the operation of Millstone Nuclear Power Station,

' Unit No. 3

' Docket No. 50-423

Northeast Nuclear Energy Company U.S. Nuclear Regulatory

• Commission

< Offico of Nuclear Reactor Regulation November 1985 p" ~%,

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., o 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 Federal Regulations, and Nuclear Regulatory Commission issuances.

Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, journal and peri'odical 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.

1 NUREG-1031 l Supplement No. 3 l l

Safety Evaluation Report related to the operation of Millstone Nuclear Power Station, Unit No. 3 Docket No. 50-423 Northeast Nuclear Energy Company i

1 U~S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation  :

November 1985

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ABSTRACT This report supplements the Safety Evaluation Report (NUREG-1031) issued in July 1984, Supplement 1 issued in March 1985, and Supplement 2 issued in September 1985 by the Office of Nuclear Reactor Regulation of the U.S. Nuclear

' Regulatory Commission with respect to the application filed by Northeast Nuclear Energy Company (applicant'and agent for the owners) for a license to operate Millstone Nuclear Power Station, Unit No. 3 (Docket 50-423). The facility is located in the Town of Waterford, New London County, Connecticut, on the north shore of Long Island Sound.

This supplement provides more recent information regarding resolution or updat-ing of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report.

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I TABLE OF CONTENTS

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Pgte iii ABSTRACT.........................................................

1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT............... 1-1 1.1 Introduction........................................... 1-1

1. 5 Outstanding Items...................................... 1-1

1. 6 Confirmatory Items..................................... 1-2 4 1. 7 License Condition Items................................ 1-2 2 SITE CHARACTERISTICS........................................ 2-1 i

2.5 Geology and Seismology................................. 2-1

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2.5.2 V i b ra to ry G ro und Mo ti o n . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 i 2.5.4 Stability of. Subsurface Materials and Foundations................................. 2-4 J 2.5.5 Stability of S1 opes............................. 2-5 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS..... 3-1

3.5 Missile Protection..................................... 3-1 3.5.1 Missile Selection and Description............... 3-1 i

I 3.8 Design of Seismic Category I Structures................ 3-1 3.8.1 Concrete Containment............................ 3-1 i 3.8.4 Other Seismic Category I Structures. . . . . . . . . . . . . 3-3 4 REACT 0R..................... .............................. 4-1 4

0 4.2 Fuel Design............................................ 4-1 t 4.2.3 Design Evaluation............................... 4-1 4.4 ' Thermal-Hydraulic 0esign............................... 4-1 4.4.4 0perating' Abnormalities......................... 4-1 4.4.5 Loose Parts. Monitoring System................... 4-2 4-3 4.5 Reactor Materials......................................

4.5.1 Control Rod Drive Structural Materials.......... 4 i l

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l TABLE OF CONTENTS (Continued)

Page 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS................ 5-1 l 5.2 Integrity of Reactor Coolant Pressure Boundary. . . . . . . . . 5-1 5.2.3 Reactor Coolant Pressure Boundary Materials..... 5-1 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing.......................... 5-3 5.4 Component and Subsystem Design......................... 5-6 5.4.2 Steam Generators................................ 5-6 6 ENGINEERED SAFETY FEATURES.................................. 6-1 6.2 Containment Systems.................................... 6-1 6.2.1 Containment Functional Design.................. 6-1 6.2.3 Secondary Containment Functional Design......... 6-2 6.2.5 Combustible Gas Control System.................. 6-3 6.2.7 Fracture Prevention of Containment P re s s u re B o u nda ry. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 8 ELECTRIC POWER SYSTEMS...................................... 8-1 8.2 Offsite Electric Power System.......................... 8-1 8.2.2 Compliance With GDC 17.......................... 8-1 8.3 Onsite Power Systems................................... 8-2 8.3.1 Onsite AC Power Ssytem's Compliance With GDC 17..................................... 8-2 8.3.3 Common Electrical Features and Requirements..... 8-3 9 AUXILIARY SYSTEMS........................................... 9-1 9.5 Other Auxiliary Systems................................ 9-1 9.5.3 Lighting System................................. 9-1 9.5.4 Emergency Diesel Engine Fuel Oil Storage and Transfer System............................. 9-2 9.5.5 Emergency Diesel Engine Cooling Water System.... 9-4 9.5.6 Emergency Diesel Engine Starting System......... 9-5 9.5.7 Emergency Diesel Engine Lubricating Oil System...................................... 9-6 9.5.8 Emergency Diesel Engine Combustion Air Intake and Exhaust System....................... 9-8 Millstone 3 SSER 3 vi

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TABLE OF CONTENTS (Continued)

Pale 10 STEAM AND POWER CONVERSION SYSTEM........................... 10-1

10.4 Other Features......................................... 10-1 10.4.2 Main Condenser Evacuation System...............

10-1 10.4.3 Turbine Gland Sealing System................... 10-1 11 RADI0 ACTIVE WASTE MANAGEMENT................................ 11-1 11.4 Solid Waste Management System.......................... 11-1 11.5 Process and Effluent Radiological Monitoring and Sampling Systems....................................... 11-1 11.5.2 Evaluation Finding............................. 11-1 i 13 CONDUCT OF OPERATIONS....................................... 13-1 6

d 13.6 Physical Security Plan................................. 13-1 i 13.6.1 Physical Security Organization................. 13-1 13.6.2 Physical 8arriers.............................. 13-1 13.6.3 Access Requirements............................ 13-2 13.6.4 Detection Aids................................. 13-3 13.6.5 Communications................................. 13-3

13.6.6 Test and Maintenance Requirements.............. 13-3 13.6.7 Response Requirements.......................... 13-4 l 13.6.8 Employee Screening Program..................... 13-4 1

14 INITIAL TEST PR0 GRAM........................................ 14-1 I 15 ACCIDENT ANALYSES........................................... 15-1 i

l 15.4 Reactivity and Power Distribution Anomalies............ 15-1 l 15.4.3 Rod Cluster Control Assembly Malfunctions...... 15-1 1

i 16 TECHNICAL SPECIFICATIONS.................................... 16-1

! 17 QUALITY ASSURANCE........................................... 17-1 17.1 General................................................ 17-1 4 17.2 Organization........................................... 17-1 17.4 Conclusion............................................. 17-1 Millstone 3 SSER 3 vii

TABLE OF CONTENTS (Continued)

Page APPENDICES APPENDIX A CONTINUATION OF CHRON0 LOGY OF THE NRC STAFF RADIOLOGICAL REVIEW 0F-THE MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 APPENDIX B BIBLIOGRAPHY APPENDIX D ABBREVIATIONS

- APPENDIX F NRC STAFF CONTRIBUTORS LIST OF TABLES

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l 1.1 Listing of Outstanding Items (revised from SER Table 1.3)... 1 i 1.2 Listing of Confirmatory Items (revised from SER Table 1.4).. 1-4 1.3 Listing of License Conditions (revised from SER Table 1.5).. 1-9 2.1 Critical Component for Plant Damage States and Their Fragilities................... ..................... ....... 2-6 2.2 Fragilities of Different Plant Damage States................ 2-7 16.1 Technical Specification Items (revised from SER Table 16.1). 16-1

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l l 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction l

In July 1984 the U.S. Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER)(NOREG-1031) on the application filed by Northeast Nuclear Energy Company (NNEC0, the applicant), acting as agent and representative for the owners for a license to operate Millstone Nuclear Power Station, Unit No. 3, Docket No. 50-423. The SER was supplemented in March 1985 by Supplement 1 (SSER 1) and in September 1985 by Supplement 2 (SSER 2); these documented the resolution of several outstanding and confirmatory items and license conditions in further support of the licensing activities. The present report, Supplement 3 to that SER (SSER 3), provides more recent information regarding the resolution or updating of some of the outstanding and confirmatory items and license con-ditions identified in the SER and its supplements.

Each of the following sections or appendices is numbered the same as the corre-sponding SER section or appendix that is being updated. Each section is supple-mentary to and not in lieu of the discussion in the SER, unless otherwise noted.

Appendix A continues the chronology of the staff's actions related to the pro-cessing of the Millstone 3 application. Correspondence between the applicant and the NRC staff is listed chronologically in this appendix. Appendix 8 lists references cited in this report.* Appendix D contains abbreviations used in this supplement, and Appendix F lists principal staff members who contributed to this supplement.

Copies of this SER supplement are available for inspection at the NRC Public Document iacm at 1717 H Street, N.W., Washington, D.C., and at the local Public Document toom of the Waterford Public Library, Rope Ferry Road, Route 156, Waterfort, Connecticut.

The NRC Project Manager for Millstone 3 is Ms. Elizabeth L. Doolittle.

Ms. Doolittle may be contacted by writing to her at the Division of Licensing, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555.

1. 5 Outstanding Items The staff identified certain outstanding items in the SER that had not been resolved with the applicant. The status of these items is listed Table 1.1 (an updated version of SER Table 1.3) and is discussed further in the sections of this report as indicated. If the staff has completed its review of an item, the notation " closed" so indicates. The staff will complete its review of un-resolved items before the operating license is issued. Resolution of each of these unresolved items will be discussed in future supplements to the SER.

• Availability of all material cited is described on the inside front cover of this report.

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Millstone 3 SSER 3 1-1

1.6 Confirmatory Items The staff identified confirmatory items in its SER that required additional information to confirm preliminary conclusions. The status of these items is listed in Table 1.2 (an updated version of SER Table 1.4) and is discussed fur-ther in the sections of this report as indicated.

If the staff has completed its review of an item, the notation " closed" so indicates.

1. 7 License Condition Items In Section 1.7 of the SER, the staff identified seven license conditions. These

, include several issues that must be resolved by the applicant as a condition for issuance of an operating license, and other issues to be resolved in the longer term (will be cited in the operating license issued), to ensure that NRC requirements are met during plant operation.

The current status of license conditions is given in Table 1.3 (an updated ver-sion of SER Table 1.5).

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Table 1.1 Listing of outstanding items (revised from SER Table 1.3)

Item Status Saction*

(1) Internally generated missiles Closed (SSER 1)

(2) Diesel generators Under review (3) Protection against postulated pipe Under review breaks outside containment (4) Loading combinations- Closed (SSER 1)

(5) Design and construction of Closed (SSER 1) component supports (6) Inservice testing of pumps and Under. review valves (7) Equipment qualification Under review i (8) Flow measurement capability Under review 4.4.4.2 i (9) Loose parts detection program Closed (SSER 3) 4.4.5 (10) Subcompartment analysis Under review (11) Mass and energy release analysis Changed to confir-matory item (71)

(SSER 2)

(12) Volumetric inspection of Class 2 Closed (SSER 2) components (13) Power-operated relief valve and Closed (SSER 3) 8.'3.3.4 block valve, compliance with NUREG-0737 (14) Fire protection Under review (15) Functional capability of ac and dc Closed (SSER 3) 9.5.3 emergency lighting i (16) Shift technical advisor training Under review program and operating experience for startup (17) Emergency Plan Under review (18) Limitation on overtime Closed (SSER 2)

(19) Q list Closed (SSER 3) 17 (20) Detailed Control Room Design Review Under review

• Section of this supplement where item is discussed.

Millstone 3 SSER 3 1-3

l Table 1.2 Listing of confirmatory items (revised from SER Table 1.4) 1 Item Status Section*

(1) Plant's seismic capability beyond Closed (SSER 3) 2.5.2.7.2 design basis (2) Dynamic loading Closed (SSER 3) 2.5.4.3.2 (3) Liquefaction potential Closed (SSER 3) 2.5.4.4 i

(4) Shoreline slope Closed (SSER 3) 2.5.5.1

. (5) Turbine maintenance program Closed (SSER 3) 3.5.1.3 (6) Barrier design procedures Closed (SSER 1)

(7) Inservice examination of all pipe Awaiting welds in break exclusion area information (8) Jet impingement effects Awaiting information (9) Ultimate capacity of containment Closed (SSER 1)

(10) Design of spent fuel racks Closed (SSER 3) 3.8.4 (11) Program evaluation related to TMI Awaiting Item II.D.1 information (12) Predicted cladding collapse time Deleted (SSER 1, Appendix H)

(13) Fuel assembly mechanical response Closed (SSER 3) 4.2.3.3 i

(14) Margins itemized in WCAP-8691 Closed (SSER 3) 4.4.4.1 (15) Thermal-hydraulic analyses to Under review

support N-1 loop operation (16) Control rod drive structural Closed (SSER 3) 4.5.1 materials (17) ASME Code cases for Section III, Closed (SSER 2)

! Class I, components l (18) Yield strength of austenitic Closed (SSER 3) 5.2.3 l stainless steels in reactor coolant pressure boundary

• Section of this supplement where item is discussed.

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Table 1.2,(Continued)

. Item Status Section*

(19) Onsite demonstration'of ultrasonic Closed (SSER 3) 5.2.4 inspection

! (20) Preservice inspection program review Awaiting l and relief requests information i

(21) Preservice and inservice inspection Closed (SSER 3) 5.4.2.2.1 of_ steam generators j (22) Containment liner weld channel Closed (SSER 2) venting

'(23) Maximum external differential Under review- 6.2.1.1 i pressure on-containment

! (24) Minimum containment pressure for Closed (SSER 2) emergency core _ cooling system (25) Procedures for actuating hydrogen Closed (SSER 3) 6.2.5

. recombiner i

(26) Secondary enclosure building- Under review 6.2.3 I (27) Sump flow approach velocity Under review i

I (28) Compliance with GDC 51 Closed (SSER 3) 6.2.7  ;

! (29) Cable separation in nuclear steam Closed (SSER 1) supply system process cabinets

! (30) Design modification for automatic Closed (SSER 2) reactor trip using shunt coil j

trip attachment

• l (31) Reactor coolant pump underspeed Under review

! . trip ,

(32) Conformance with BTP ICSB-26 Closed (SSER 1)

(33) Test of engineered safeguard P-4 Closed (SSER 1) interlock  :

I (34) Steam generator level control and ' Closed (SSER 2).

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• Section of this supplement where item is discu: sed.

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Table 1.2 (Continued)

Item Status Section* '

(35) Confirmatory test related to IE Closed (SSER 2)

Bulletin 80-06 (36) Control building isolation reset Closed (SSER 2)

'(37) Power lockout feature for Closed (SSER 1) motor-operated valves (38) Failure mode and effects analyses Closed (SSER 1) of engineered safety features actuation system (39) Non-Class 1E control signals to Closed (SSER 2)

Class IE control circuits (40) Sequencer deficiency repe.t Closed (SSER 2)

(41) Balance-of piant instrumentation Closed (SSER 2) and control system testing capability (42) Instrument accuracy related to Closed (SSER 2)

Positions [ Attachments] 4, 5, and 6, NUREG-0737, Item II.F.1 (43) Description and analysis Closed (SSER 1) demonstrating compliance with GDC 5 (44) Physical separation of offsite Closed (SSER 3) 8.2.2.1 circuits within a common right of way (45) Physical sapsration of offsite Closed (SSER 3) 8.2.2.2 circuits between switchyard and Class IE system (46) Generation rejection scheme Closed (SSER 3) 8.2.2.5 (47) Description and analysis Closed (SSER 1) demonstrating compliance with GDC 17 (48) Description and analysis Closed (SSER 1) demonstrating compliance with GDC 18

• Section of this supplement where item is discussed.

Millstone 3 SSER 3 1-6

i Table 1.2 (Continued)

Item Status Section*

(49) Positive statement of compliance Closed (SSER 1) with BTP PSB-1 (50) Compliance with Position 1 of Closed (SSER 3) 8.3.1.3 BTP PSB-1 (51) Adequacy of station electric Under review 8.3.1.5 distribution system voltage (52) Routing of power cables in the Deleted (SSER 1, ,

cable spreading area Appendix H)

(53) Battery charger and transformer Closed (SSER 3) 8.3.3.3.10 used as isolation devices (54) Design criteria of associated Deleted (SSER 1, circuits from isolation device to Appendix H) ~*

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(55) Core damage procedure (II.B.3, Closed (SSER 1)

Criterion 2)

(56) Control of concrete dust Closed (SSER 3) 9.5.4.1(2)

(57) Qualification of engine-mounted Closed (SSER 3) 9.5.4.1(4) ,

control panels (58) 7-day fuel oil of storage of each Under review

. diesel' generator 1 (59) Airborne radioactivity monitoring Closed (SSER 3) 10.4.2,  ;

10.4.3 (60) Process control program for Closed (SSER 3) 11.4 solidification of wet wastes (61) TMI Action Plan Item II.F.1.1 Closed (SSER 3) 11.5 (62) TMI Action Plan Item I.C.1 - Under review procedures generation package nuclear' steam supply system (63) Physical Security Plan Closed (SSER 3) 13.6

• Section of this supplement where item is discussed.

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Table 1.2 (Continued)

Item Status Section* l (64) Initial test program Closed (SSER 3) 14 (65) Reactor coolant pump trip during Under review loss-of-coolant accident (66) TMI Action Plan Item III.D.1.1 Awaiting information (67) Analysis of dropped control rod Closed (SSER 3) 15.4.3 (68) Steam generator tube rupture Under review (69) No failure in emergency core Deleted (SSER 2) cooling system (ECCS) is not most limiting case in evaluating ECCS (70) QA program commitments Closed (SSER 3) 17 (71) Mass and energy release analysis Under review

• Section of this supplement where item is discussed.

Millstone 3 SSER 3 1-8

Table 1.3 Listing of license conditions (revised from SER Table 1.5)

Item Status Section*

(1) Instrumentation for monitoring post- Under review accident conditions, RG 1.97, Rev. 2, requirements (2) Compliance with NUREG-0612 (" Heavy Load Closed (SSER 2)

Handling")

(3) Installation of postaccident sampling Unchanged system (4) Sediment control during fuel oil storage Closed (SSER 3) 9.5.4.2 tank refill (5) Moisture in air start system Revised 9.5.6 (6) Preheating of rocker arm lubrication Closed (SSER 3) 9.5.7 oil system (7) Blockage of access hatch in diesel Closed (SSER 3) 9.5.8 generator exhaust system (8) Plant-specific analyses utilizing Added (SSER 2)

NOTRUMP (TMI Item II.K.3.31)

• Section of this supplement where item is discussed.

Millstone 3 SSER 3 1-9

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4 2 SITE CHARACTERISTICS .

I 2.5 Geoloqy and Seismology 2.5.2 Vibratory Ground Motion ,

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2.5.2.7.2 Probabilistic Estimates of Ground Motion at Millstone ,

IntheMillstone3SER,thestaffreaffirmedtheseismicdesignbasisfNewmdrk-type response spectrum anchored at 0.17 g) approved for Millstone 3 at tie coh-struction permit stage. However, there was sufficient uncertainty associat6J with the causes of the 1982, body-wave-magnitude 5.75, New Brunswick earthquake ,

that the staff made a. limited evaluation of the ground motion resulting from reoccurrence of that size of earthquake in the vicinity of Millstone 3. In order to address this uncertainty, the staff utilized the preliminary _ insights gained from the Millstone 3 Probabilistic Safety Study (PSS) to conclude in the SER that the contributions to core melt from the seismic hazard for peak accel- Ay erations less than 0.30 g are small and that differences of 40% to 50% in ground motion at accelerations less than or equal to about 0.17 g (safe shutdown earth-quake (SSE) level) to 0.25 g (assumed ground motion level for nearby magnitude 5.75 event) are not significant when viewed from the perspective of risk. How-ever, the staff also stated in the SER that as confirmation of the conclusion drawn, the staff will require the applicant to utilize the results of the PSS to document the seismic capability, at acceleration up to 0.25 g, with high con-fidence of low probability of failure for individual controlling failure modes of structures and equipment. In addition, the applicant was required to assess plant fragilities for various acceleration levels considering those risk scenar- y los that include the majority of seismic risk to the plant.

By a letter dated December 6,1984, the applicant submitted a raport tit 1t.d "A ProgramToDeterminetheCapabilityoftheMillstone3Nuclearc[owerPlantTo Withstand Seismic Excitation Above the Design SSE" (Ravindra et al., 1984) for the staff review. The report contained information on the following aspects:

(1) identification of dominant contributors to seismic risk (2) evaluation of the high-confidence, low-frequency-of-failure accelerations for critical structures and equipment (3) evaluation of the high-confidence, low-frequency-of-failure fccelerations for the dominant plant damage states , s  ;

(4) evaluation of the frequencies of occurrence of significant plani. damage states from seismic events (5) evaluation of the contributions or various acceleration ranges to the frequencies of occurrence of significant plant damage states 3

Millstone 3 SSER 3 2-1

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(6) investig'ation of the sensitivity of the results of the Millstone 3 PSS to variation in the assumption or analytical models employed for the seismic fragility and hazard development The plant damage states for which the seismically initiated accident sequences have been shown to be important contributors are (Ravindra et al., 1984):

TE - transient (caused by loss of offsite power) with failure of onsite emergency power or reactor coolant system (RCS) heat removal AE - large loss-of-coolant accident (LOCA) with failure of safety injection and containment quench sprays SE - small LCCA or seismic anticipated transient without scram (ATWS) with failure of safety injection and containment quench sprays 1 V3 - LOCA with containment bypass The staff's Boolean expressions (relating the plant damage states to the component failures) discussed in the staff evaluation of the Millstone 3 PSS (NRC, NUREG-1152) are almost identical to those presented in the applicant's report. (The major difference is that the staff assigns the reactor coolant pump seal failure on station blackout to plant damage state SE, not to TE).

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The components that dominate the seismic risk for each damage state are shown in Table 2.1 along with their median acceleration capacities, uncertainties, and high-confidence / low probability failure (characterized as failure frequency less than 5% with 95% confidence) capacities. It should be noted that the Boolean expressions for the plant damage states contain some components whose capacities may be lower than those shown in Table 2.1; however, failures of these components by themselves are not critical. Design and construction errors and relay chatter are not explicitly considered in the capacity deter-mination or plant system analysis.

On the basis of the structure / component fragilities discussed above, the appli-cant calculated plant damage state fragilities / margin as shown in Table 2.2.

" The staff review of the structural / component capacities and plant damage state fragilities indicated that, except for the loss of offsite power, critical structures and compe n nts identified in the report have high-confidence, low-frequency-of-failure accelerations of at least 0.3 g. However, the staff noted that the applicant's analysis did not explicitly address the consequences of V seismically induced liquefaction of beach sands nor the stability of the beach sand slope under events greater than SSE which may prevent the intake structure from conveying the cooling water for the safe shutdown. In addition, the staff requested that the applicant evaluate the fragility analyses for the emergency generator enclosure (EGE) building and buried service water piping system con-sidering as-built foundation support conditions and variations in the assumed ratio of peak ground velocity to peak ground acceleration.

i The staff met with the applicant on May 14, 1985 to discuss the above issues and by a lytter dated July 12, 1985, the applicant provided results of its analyses f6r the staff review. To address the liquefaction issue, the applicant examined a seismically induced flow slide for three sections through the slope i

s Millstone 3 SSER 3 2-2

and dredged channel. A final post-flow slope was assumed on the basis of obser-vations from the 1964 Alaska earthquake. Examination of the post-flow channel elevation and the service water pump inlet elevation indicates that a seismically induced flow slide into the intake channel,will not adversely affect the supply of water required for cooling safety-related systems. Although the applicant's analyses did not address' specific seismic events above SSE, on the basis of the staff's past review experience and the applicant's study, the staff concludes that the flow slide of beach sand slope resulting from seismic events up to 0.3 g would not be a significant seismically induced failure mode.

The results of revised fragility analysis of EGE building and service water piping system indicate that changes in the median acceleration capacities and high confidence, low frequency of failure levels are minor and, therefore, impacts on the plant damage state frequencies are minimal.

The applicant also presented the plant damage state frequencies obtained by convolving plant damage state' fragi7ity curves with the seismic hazard curves.

(See staff evaluation of the Millstone 3 PSS (NRC, NUREG-1152) for further dis-

cussion on plant damage state frequencies.) From these calculations, the applicant

! also extracted information about which ranges of acceleration contribute most significantly to the overall frequency of occurrence of the damage state. For the plant damage state TE, which is the predominant contributor to the core-melt frequency (in the NRC staff analysis, NUREG-1152, SE is the predominant contrib-utor), it was observed that the contribution to 95% confidence frequency of occurrence and the median frequency of occurrence from the acceleration ranges below 0.3 g is small.

The applicant, conducted c number of sensitivity studies to examine the influence of assumptions made in the component fragility analysis and in different seismic

. hazard models (NRC, NUREG/CR-3756) on the plant damage state frequencies. These studies, in general, indicate that variations considered do not significantly

'itipact'the plant's capt.bility to withstand seismic events greater than SSE.

On the basis of the review of the above information and discussions with the applicant, the staff finds the following:

(1) Given the use of the same. set of hazard curves, the applicant's analysis is in general agreement with that done by the staff in its review of the PSS.

The applicant's results appear somewhat more conservative than the staff's for a given set of hazard curves because of neglect of overlap between dif-ferent plant damage states in the applicant's work (NRC, NUREG-1152). Al-though neglect of overlap increases the overall core damage frequency, it diminishes the percent contribution of the lower acceleration range since plant damage state overlap is most significant at higher accelerations.

, (2) The applicant's plant damage state Boolean expressions (derived from the PSS fault trees) are almost identical to those developed by the staff.

The applicant's Boolean expressions are slightly more conservative than the staff's, and the staff would place some failures in different plant damage states. However, the differences are minor.

(3) In general, the critical structures and components at Millstone 3 have high-confidence, low-frequency-of-failure accelerations of at least 0.3 g.

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l (Relay chatter and design / construction errors are not considered in this evaluation. See NUREG-1152 (NRC) for more discussion on relay chatter issue.)

(4) The dominant plant damage state TE at Millstone 3 has high-confidence, low-frequency-of-failure acceleration greater than 0.25 g.

(5) The analysis of contributions of different peak ground accelerations ranges to the plant damage state frequencies confirms that the contribution of earthquakes of up to 0.3 g is not significant.

On the basis of the above findings, the staff concludes that Millstone 3 and its critical structures and components possess significant margins beyond the design-basis SSE and there is a hign confidence that the frequency of the plant damage is significantly low at acceleration levels less than 0.25 g. Therefore,

. confirmatory issue (1) regarding the plant's seismic capability beyond design basis, i.e., reoccurrence of earthquake the size of 1982 New Brunswick event in the vicinity of the plant, is considered resolved.

2.5.4 Stability of Subsurface Materials and Fouisdations 2.5.4.3 Foundation Stability 2.5.4.3.2 Dynamic Loading In the Millstone 3 SER, in a letter dated June 26, 1984, the applicant committed to perform additional analysis incorporating 'the as-built soil / foundation condi-tion to confirm the design of the emergency generator enclosure (EGE) building.

During the meeting of May 14, 1985, the applicant indicated that during con-struction it had reexamined the locations of field density tests and discovered errors in locating some of the test samples. The applicant stated that correc-tions to the FSAR will be made in a future amendment to reflect the relocation.

The relocation of those test samples confirms that the majority of the strip footings of the EGE building were founded on basal till, and that one strip of footing was founded on a few feet of structural fill over till. The presence of this small amount of structural fill is judged to have insignificant effect on the seismic response of the EGE building.

The staff accepts the applicant's explanation that errors were probably made during consti-uction in locating some of the test samples and concurs with the applicant's assessment that the presence of a smal, amount of structural fill would have no impact on the EGE building. The soil-structure-interaction anal-ysis performed for the EGE building presented in the Final Safety Analysis Re-I port (FSAR) is acceptable to the staff and, therefore, confirmatory item (2) is resolved.

2.5.4.4 Liquefaction Potential As stated in the Millstone 3 SER, the applicant's analyses performed in May 1984

indicated that the beach sand deposits might be liquefiable under safe shutdown j earthquake (SSE) condition and the effects of liquefied beach sands on the func-tionality of the pumphouse need to be asscssed.

i Millstone 3 SSER 3 2-4

The applicant has performed an additional analysis assuming that liquefaction of beach sand deposits would take place under SSE condition. The staff has re-viewed this information which shows that the effect of the liquefied beach sand deposits would not impair the functionality of the pumphouse.

l The staff concurs with the applicant's assessment that a liquefaction-induced

! flow slide of beach sand deposits will not adversely affect the supply of emer-l gency cooling water. Confirmatory item (3) is resolved.

2.5.5 Stability of Slopes 2.5.5.1 Shoreline Slope In the Millstone 3.SER, the staff stated that additional information is required from the applicant to justify the adequacy of the retaining-wall design.

By letter dated October 18, 1984 and in a Stone & Webster Engineering Corpora-tion report transmitted May 15, 1985, the applicant has provided additional in-formation to justify adequacy of the retaining-wall design. The staff has re-viewed this information and concludes that the design of the retaining wall is adequate. Confirmatory item (4) is resolved.

On the basis of the information submitted, the staff concludes that: (1) the t

liquefaction of beach sand deposits under SSE condition would not affect the safety function of the pumphouse, (2) the foundation model used for the seis-mic analysis of the EGE building is acceptable, and (3) the west retaining-wall design is acceptable.

Millstone 3 SSER 3 2

< 1 l

Table 2.1 Critical component for plant damage states and their fragilities Median High confidence, Plant accelera- Uncertainty low frequency, damage Component / structure tion, g failure level, g !

state failure mode (50%-50%) Random modeling (95%-5%)

TE- Loss of offsite power 0.20 0.20 0.25 0.10 Emergency generator 0.88 0.20 0.46 0.30 l enclosure building wall footing failure Diesel generator oil 0.91 0.24 0.43 0.30 cooler-anchor bolt failure Control building 1.00 0.24 0.43 0.39 diaphragm Service water pumphouse 1.30 0.24 0.49 0.39 sliding Engineered safe guard 1.70~ 0.23 0.43 0.58 feature building. Fail-ure of shear wall near basemat AE Reactor coolant 1.59 0.48 0.51 0.31 4

system piping SE Reactor vessel core 0.99 0.21 0.33 0.35 geometry distortion Control rod drive system 1.00 0.30 0.38 0.33 (failure to scram)

V3 Containment cranewall 2.20 0.39 0.38 0.62 failure i

! Millstone 3 SSER 3 2-6

i i

! Table 2.2 Fragilities of different plant damage states High confidence, Median low frequency of acceleration failure level, g Plant damage state g (50%-50%) (95%-5%)

V3 LOCA with containment bypass 2.05 0.60 AE Large LOCA with early core melt 1.22 0.45 SE Small LOCA or ATWS with early 0.77 0.40 core melt TE Transient (loss of offsite power) 0.61 0.26 with early core melt Millstone 3 SSER 3 2-7

3 DESIGN CRITERIA FOR STRhCTURES, SYSTEMS, AND 3.5 Missile Protection 3.5.1 Missile Selection and Description 3.5.1.3 Turbine Missiles 3.5.1.3.3 Summary In Supplement 1 to the Millstone 3 SER. it was stated that the applicant agrees to submit for NRC approval, within 3 yurs of obtaining an operating license, a turbine system maintenance program based on the manufacturer's calculations of missile generation probabilities with the option of conducting on independent review and analysis if so desired.

If review of the applicant's program or independent review and analysis results in the need for additional information or modifications, the applicant will be required to submit this information or make the appropriate modifications at that time. Confirmatory item (5) is, therefore, resolved.

3.8 Design of Seismic Category I Structures 3.8.1 Concrete Containment Millstone 3 experienced a construction fire in the reactor containment which damaged the liner plate. When the fire-damaged containment liner plate was re-moved, an area approximately 5 ft x 3 ft was identified as missing stud anchors to the concrete containment wall. Subsequent examinations disclosed that sev-eral other areas in the liner also were missing studs.

The applicant issted "Nonconformance and Disposition Report 2581" on May 18, 1983, and subsequently the issue was closed out in February 1984 concluding that, based on detailed analyses, "the as-built location of studs in the af-fected area outlined by this Nonconformance and Disposition Report is accept-able." The applicant submitted a report and accompanying calculations for staff concurrence. The staff reviewed the report and a summary of that review follows. In particular, the staff evaluated analytical calculations of liner-stud interaction and resultant stresses.

Calculations were performed to demonstrate that the liner and anchor studs at or near the boundary of the area where studs are missing do not exceed the allowables stated in the FSAR when subjected to accident condition loadings.

Two models were used to analyze the liner and studs; one model was formulated with the ANSYS finite element program and the other was formulated using large

deflection theory of plates and manual calculations. The ANSYS model is a three-dimensional model of the as-built liner containing the 3-ft x 5-ft area; missing studs were centered at elevation 65 ft 0 in.

l l

Millstone 3 SSER 3 3-1

r-The loading combination applied to both models is a pressure of -6.7 psi and a temperature change of 120.5 Fahrenheit degrees. The loading would result when the containment spray system acted with the containment purge air system to lower the containment pressure. This combination represents the most severe set of conditions for a plate without anchor studs.

To account for irregularities from a perfect cylinder, an out-of-roundness sur-vey was performed in the field and the results were used as input for the models.

A 10-ft x 12-ft area of the liner was represented in the ANSYS model. The model had an approximate radius of 840 in. The stud locations and a radial profile of the liner at elevation 65 ft 0 in. as determined from field surveys were used as input for the model. The model extended three to four studs in all directions beyond the area missing studs.

Triangular plate elements with both membrane and bending capabilities were used to represent the liner. Two mesh sizes were used to model the liner. A mesh of triangles with 3-in. sides was used in the area lacking studs and went two studs beyond on all sides. The outer edges of the model contained elements with 6-in. sides. The small mesh was used in the critical area so that the concrete and the liner profile could be defined more accurately.

Gap elements were used to model the concrete at every node of the liner elements. The gap elements were basically spring elements which had a high stiffness in compression but no stiffness in tension. The purpose of these elements was to restrict the deflection of the liner toward the concrete, but to allow the liner free movement away from the concrete.

The studs were represented by nonlinear force-deflection elements. These were spring elements with I degree of freedom. A force deflection curve was input to model the change in stiffness with load based on experimental results. Three elements were required at each stud node to model the axial stiffness and the shear stiffness of the stud in the horizontal and vertical directions.

The manual calculations consisted of two separate steps which required iteration to converge to an edge displacement consistent for both the laterally deflected liner plate and the adjoining studs. In the first :;tep, lateral deflection of a rectangular plate clamped on all four sides ana' 3abjected to in place displace-ments was computed as a function of the amplitude of the initial distortion, the temperature increase, transverse pressure loading, and assumed in plane edge displacements. Using small strain and large displacement theory, the Rayleigh-Ritz method was applied with displacement function compatible with the first mode shape of a buckled plate.

The second step was a strip analysis in which the loading consisted of displace-ments experienced by the cracked concrete wall and the embedded ends of the studs under the loading conditions of interest. A stiffness analysis with six plate elements and five studs extending horizontally or vertically away from the plate without studs was then performed with the unknown displacements equal to the l movement of the stud / liner joint. Equilibrium equations at each stud were for-mulated using the linear behavior of the adjoining plates, the known force (shear)-slip relationship for the studs, and a constant reaction force of the Millstone 3 SSER 3 3-2

laterally displaced liner plate based upon an assumed in plane edge displace-ment. The computed edge displacement was' compared to the assumed displacement and the steps were repeated until a close agreement was reached.

The results from ANSYS and manual calculation were compared and found to be in good agreement. In addition they were within allowable values specified in the ASME Boiler and Pressure Vessel Code, Division 2, and Millstone 3 FSAR cri-teria. This provides a reasonable assurance that the liner will maintain leak-tight integrity for all loading conditions including design-bases accident.

3.8.4 Other Seismic Category I Structures In the Millstone 3 SER, it was noted that the applicant had informed the staff that the design of the spent fuel pool racks complied with the current staff acceptance criteria [ Standard Review Plan (SRP) Section 3.8.4, Appendix 0]. The staff's intention to review the information which will confirm such compliance was also indicated in the SER.

By a letter dated May 20, 1985, the applicant has now provided details of its analysis / design of the spent fuel racks. In particular, these details include: (1) description of the fuel rack assembly; (2) models and procedures for seismic analysis; (3) loads and load combinations for structural analysis; (4) sliding and overturning analysis; and (5) structural acceptance criteria.

The staff review of this information indicates that the applicant's seismic analysis is consistent with the staff acceptance criteria accounting for the nonlinearities resulting from the gap between the fuel cell and the fuel assembly, the boundary conditions of the fuel rack support locations and energy losses at the support locations. Thus, the nonlinear model accounts for. fuel to rack impact loading, support pad liftoff, hydrodynamic forces, and the nonlinearity of sliding friction interfaces. The sliding and overturning analysis indicates that the impact between adjacent rack modules, and rack module and pool wall is prevented. The factor of safety against overturning is much greater than the staff acceptance criteria. The load combinations and acceptance criteria are in accordance with the earlier staff position paper, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," and they also meet the requirements of SRP Section 3.8.4, Appendix D, using the 1980 Edit. ion, Winter 1982 Addendum of subsection NF of the ASME Boiler and Pressure Vessel Code,Section III, Division 1.

On the basis of the above findings, the staff concludes that the design of the spent fuel pool racks at Millstone 3 complies with the intent of the staff acceptance criteria and, therefore, confirmatory item (10) is resolved.

f Millstone 3 SSER 3 3-3

4 REACTOR 4.2 Fuel Design 4.2.3 Design Evaluation 4.2.3.3 Fuel Coolability Evaluation (4) Structural Damage From External Forces In Supplement 1 to the Millstone 3 SER, the staff agreed with the applicant that the Millstone 3 deviation from the approved bounding Westinghouse seismic re-sponse curve appeared to be a secondary effect. However, the staff stated that it would complete the review pending the applicant's further analysis of seismic and loss-of-coolant accident (LOCA) loads on fuel assemblies.

By a letter dated February 1, 1985, the applicant provided the results of a combined seismic-and-LOCA loads analysis including asymmetric blowdown load using the approved methodology described in Westinghouse document WCAP-9401.

The results show that the combined loads on grids and on non grid components were less than the allowable strengths for Millstone 3.

Therefore, the staff concludes that the applicant has demonstrated acceptable results for fuel assemblies under combined seismic-and-LOCA conditions, and confirmatory item (13) is resolved.

4.4 Thermal-Hydraulic Design 4.4.4 Operating Abnormalities 4.4.4.1 Fuel Rod Bowing A significant parameter which affects the thermal-hydraulic design of the core is rod-to-rod bowing within fuel assemblies. The staff has approved the Westinghouse methods for predicting the effects of rod bow on departure from nucleate boiling (DNB) (WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation").

The FSAR stated that there is a 9.1% margin to accommodate full- and low-flow DNB ratio (DNBR) penalties that result from fuel rod bowing. In a previous SER supplement (SSER 1), the staff stated that the applicant should verify that (1) the breakdown of this margin into individual factors is consistent with WCAP-8691 and (2) this margin (in whole or part) was not used in any other analysis.

Per the staff's request, the applicant inserted into the Bases of the Technical Specifications the markdown of the generic margins that were used to offset the reduction in DNBR due to rod bowing. Also, in a letter dated May 2, 1985 the applicant stated that: "the DNBR margin used to offset the worst case rod bow penalty is not used in any other analysis"; therefore, the applicant's use of available margins to of fset rod bow penalties is acceptable, and confirmatory item (14) is resolved.

Millstone 3 SSER 3 4-1

r 4.4.4.2 Crud Deposition In response to. question Q492.4 and subsequently Q492.7, the applicant stated that the reactor coolant system (RCS) flow measurement is based upon performing a precision heat balance flow measurement at the beginning of each fuel cycle and using the result to calibrate the RCS elbow tap flow indicators. In a  ;

letter dated September 19, 1984, the applicant described the inspection of the venturis before startup of each cycle via ports located upstream and downstream of the venturis. The applicant stated that cleaning will be done by hydrolasing when required. These inspection ports will be installed during the first refueling outage. The applicant stated that if the venturis are not inspected, an additional 0.1% will be added to the total RCS flow measurement I uncertainty. This is acceptable. '

In a letter dated July 15, 1985, the applicant stated that the method used in determining error of an instrument loop is the statistical combination of the groups of components in an instrument which are statistically independent.

Errors that are not statistically independent are combined arithmetica11y.

The applicant stated that vendor technical manuals and drawings were used as sources of uncertainty. The applicant's results were as follows:

Four loops Three loops The total uncertainty in determining core power equals............................... 10.44% 0.50%

Total RCS flow uncertainty based on a precision heat balance equals ............. 12.31% 2.32%

The accuracy of the RCS elbow tap flow indicators in determining tota' flow is.... 10.54% 0.54%

Combining the uncertainty of the above gives total RCS flow measurement uncertainty equal to ................. ................ 12.37% 2.37%

If venturis are not verified clean, then total RCS flow measurement uncertainty equals..................................... 12.47% 12.47%

However, the applicant has not provided enough detail for the staff to deter-mine that this analysis is valid. For example, the three-loop value should be higher than the two-loop value. The applicant should provide a breakdown of the total flow uncertainty into its components similar to the analysis provided in the applicant's February 16, 1984, letter on the same subject.

4.4.5 Loose Parts Monitoring System The applicant has provided a description of the loose parts monitoring system (LPMS) which will be used at Millstone 3. The design will consist of 12 active

, instrumentation channels, each comprising a piezoelectric accelerometer (sensor) l and signal conditioning equipment. Sensors are fastened mechanically to the l reactor coolant system (RCS) at each of the following o"tential loose parts collection regions:

Millstone 3 SSER 3 4-2

(1) reactor pressure vessel--upper head region (2) reactor pressure vessel--lower head region (3) each steam generator--reactor coolant inlet region The system will be capable of detecting a metallic loose part that weighs from 0.25 to 0.30 pound impacting within 3 feet of a sensor and having a kinetic energy of 0.5 foot pound on the inside surface of the RCS pressure boundary.

In a letter dated August 26, 1985, the applicant submitted, in response to staff question Q492.5, a report describing operation of the system hardware and imple-

mentation of the loose part detection program.

In that report, the applicant agreed to install a second sensor on each steam generator by the end of the first refueling outage. That is acceptable. Also, the applicant's discussion regarding the LPMS qualification for an operating basis earthquake (0BE) is acceptable. The discussion of the calibration of the LPMS every refueling or every 18 months, whichever is greater, and the frequency of the operability checks are also acceptable. Outstanding item (9) is resolved.

However, the staff will require the applicant to include a Technical Specifica-tion on the operability of the LPMS similar to the generic Westinghouse Tech-nical Specification as shown in Technical Specification Section 3/4.3.3.9.

4.5 Reactor Materials 4.5.1 Control Rod Drive Structural Materials The staff concludes that the control rod drive mechanism structural materials are acceptable and meet the requirements of General Design Criteria (GDC) 1, 14, and 26 (Appendix A, 10 CFR 50) as well as 10 CFR 50.55a. This conclusion is based on the applicant hav bg demonstrated that the properties of materials selected for the control red drive mechanism components exposed to the reactor coolant satisfy Appendix I of Section III of the ASME Boiler and Pressure Ves-sel Code, and Parts A, B, and C of Section II of the Code and conform with the staff p* ition that the yield strength of cold-worked austenitic stainless steel should not exceed 90,000 psi. Conformance with the applicable Code cases listed in Regulatory Guide (RG) 1.85, " Materials Code Case Acceptability, ASME Section III, Division 1," is addressed in SER Section 5.2.1.2.

In addition, the controls imposed on the austenitic stainless steel of the mech-anisms satisfy, to the extent practical, the recommendations of RG 1.31, " Con-trol of Ferrite Content in Stainless Steel Weld Metal," and RG 1.44, " Control of the Use of Sensitized Stainless Steel." The staff has reviewed and found acceptable the alternate method of control of ferrite content by testing of the purchased material and the modification of testing procedures to evaluate weld-ments for stress corrosion cracking. The use of Standard ASTM A708 as an alter-native test method to ASTM A262 Practice A or E for sensitivity to intergranular corrosion has been reviewed by the staff and accepted. Bending the test speci-mens over a 4-t* mandrel rather than a 1-t mandrel is acceptable to the staff.

• f = thickness of the material.

Millstone 3 SSER 3 4-3

Yield strength stresses will be achieved in the 4-t bend specimen, and the 4-t bend is a test requirement for qualification of weld procedures in Section IX of the Code. These adjustments are made because of the variabilities in properties that occur in a weldment. The applicant has confirmed that the tempering tem-peratures and aging temperatures of heat-treatable materials in the control rod drive mechanism are specified to eliminate the susceptibility to stress corro-sion cracking in reactor coolant. The fabrication and heat treatment practices performed provide assurance that stress corrosion cracking will not occur during the design life of the components. The compatibility of all materials used in the control rod system in contact with the reactor coolant satisfies the cri-teria of Articles NB-2160 and N8-3120 of Section III of the Code. Cleaning and cleanliness controls are in accordance with ANSI Standard N45.2.1-1973, " Clean-ing of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants," and RG 1.37, " Quality Assurance Requirements for Cleaning Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants."

Confirmatory item (16) is, therefore, resolved.

Millstone 3 SSER 3 4-4

5 REACTOR C0OLANT SYSTEM AND CONNECTED SYSTEMS 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.3 Reactor Coolant Pressure Boundary Materials t

l The staff concludes that the plant design is acceptable and meets the require-i ments of General Design Criteria (GDC) 1, 4, 14, 30, and 31 (Appendix A to 10 CFR 50); the requirements of Appendices B and G to 10 CFR 50; and the re-quirements of 10 CFR 50.55a. This conclusion is based on the staff's review of the FSAR.

The materials used for construction of components of the reactor coolant pres-sure boundary (RCPB) have been identified by specification and found to be in conformance with the requirements of Section III of the ASME Boiler and Pres-sure Vessel Code. Compliance with the above Code provisions for materials specifications satisfies the quality standards requirements of GDC 1 and 30, and 10 CFR 50.55a.

The materials of construction of the RCPB exposed to the reactor coolant have been identified and all of the materials are compatible with the primary coolant water, which is chemically controlled in accordance with appropriate Technical Specifications. This compatibility has been proven by extensive testing and satisfactory performance. This includes satisfying, to the extent practical, the recommendations of Regulatory Guide (RG) 1.44, " Control of the Use of Sensitized Stainless Steel." Where the recommendations of the regulatory guide were not followed, the alternative approaches taken have been reviewed by the staff and are acceptable (see Section 4.5.1 of this supplement).

General corrosion of all materials in contact with reactor coolant is negligible, and accordingly, general corrosion is not of concern. Compatibility of the materials with the coolant and compliance with the Code provisions satisfy the requirements of GDC 4 relative to compatibility of components with environmental conditions.

The materials of construction for the RCPB are compatible with the thermal insulation used in these areas. The thermal insulation used on the RCPB is either the reflective stainless steel type or is made of nonmetallic compounded materials that meet most of the recommendations of RG 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steels." The use of standard commercial packaging with receipt inspection for damage, as an alternative approach to the special packaging recommendations in the guide, is acceptable to the staff.

Conformance with the above recommendations satisfies the requirements of GDC 14 and 31 relative.to prevention of failure of the RCPB.

The ferritic steel tubular products and the tubular products fabricated from austenitic stainless steel have been found to be acceptable by nondestructive examinations in accordance with provisions of the ASME Code,Section III.

Compliance with these Code requirements satisfies the quality standards require-ments of GDC_1 and 30, and 10 CFR 50.55a.

Millstone 3 SSER 3 5-1

r The fracture toughness tests required by the ASME Code, augmented by Appendix G to 10 CFR 50, provide reasonable assurance that adequate safety margins against nonductile behavior or rapidly propagating fracture can be established for all pressure-retaining components of the RCP8. The use of Appendix G to the ASME Code,Section III, and the results of fracture toughness tests performed in accordance with the Code and NRC regulations in establishing safe operating procedures, provide adequate safety margins during operating, testing, main-tenance, and postulated accident conditions. Compliance with these Code pro-visions and NRC regulations satisfies the requirements of GDC 31 and 10 CFR 50.55a regarding prevention of fracture of the RCP8.

l The applicant has taken alternative approaches to the recommendations of RG 1.50, " Control of Preheat Temperature for Welding Low Alloy Steels." The alternative approaches taken by the applicant are that (1) welding procedures are qualified within the preheat temperature range (minimum limit plus 50 Faren-heit degrees) rather than at the minimum preheat temperature and (2) preheat temperatures are maintained for an extended period of time rather than until the start of post-weld heat treatment. The staff concludes that these alter-native approaches will not have a significant effect on the propensity for hydrogen cracking (the concern of RG 1.50) and will not cause other hazards.

Accordingly, the staff accepts these alternative approaches. The controls used provide reasonable assurance that components made from low-alloy steels will not crack during fabrication. If cracking does occur, the required Code inspec-tions should detect such flaws. These controls satisfy the quality standards requirements of GDC 1 and 30, and 10 CFR 50.55a.

RG 1.34, " Control of Electroslag Weld Properties," is not applicable because the electrcslag welding process was not used on RCPB components.

The controls imposed on welding ferritic and austenitic steels under conditions of limited accecsibility satisfy, to the extent practical, the recommendations of RG 1.71, " Welder Qualification for Areas of Limited Accessibility." The applicant's contractors maintain close supervisory control of the welders, and reoccurrence of welding situations in production are adequate to ensure that the most skilled welders are used in areas of licited accessibility. The staff concludes that as such welds are inspected, qualification of the welders making acceptable welds occurs automatically under the Code. These controls satisfy the quality standards requirements of GDC 1 and 50, and 10 CFR 50.55a. The controls imposed on weld cladding of low-alloy steel components by austenitic stainless steel are in accordance with the recommendations of RG 1.43, " Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components." These controls provide assurance that practices that could result in undercladding cracking will be restricted. The controls also satisfy the quality standards require-ments of GDC 1 and 30, and 10 CFR 50.55a.

The controls to avoid stress corrosion cracking in RCP8 components constructed of austenitic stainless steels limit yield strength of cold-worked austenitic stainless steels to 90,000 psi maximum and satisfy, to the extent practical, the recommendations of RG 1.44, " Control of the Use of Sensitized Stainless l Steel," and meet the recommendations of RG 1.37, " Quality Assurance Pequire-l ments for Cleaning of Fluid Systems and Associated Components of Water Cooled

! Nuclear Plants." The alternate approaches taken by the applicant to RG 1.44 were reviewed by the staff and are acceptable (see Section 4.5.1 in this supplement).

Millstone 3 SSER 3 5-2

The controls followed during material selection, fabrication, examination, pro-tection, sensitization, and protection from contamination, provide reasonable assurance that the RCPB components of austenitic stainless steels are in a metallurgical condition that minimizes susceptibility to stress corrosion cracking during service. These controls meet the requirements of GDC 4 rela-tive to compatibility of components with environmental conditions and require-ments of GDC 14 relative to prevention of leakage and failure of the RCPB.

The controls imposed during welding of austenitic stainless steels in the RCPB satisfy, to the extent practical, the recommendations of RGs 1.31, 1.34, and 1.71. The alternate approaches taken by the applicant were reviewed by the staff and are acceptable (see Section 4.5.1 in this supplement).

These controls provide reasonable assurance that welded components of austenitic stainless steel did not develop microfissures during welding and have high structural integrity. These controls meet the quality standards requirements of GDC 1 and 30, and 10 CFR 50.55a, and satisfy the requirements of GDC 14 rela-tive to prevention of leakage and failure of the RCPB. Confirmatory item (18) is, therefore, resolved.

5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing This evaluation resolves SER confirmatory item (19) concerning the demons?. cation of the adequacy of ultrasonic inspection techniques for cast stainless steel pipe. The applicant'has not yet submitted its request for relief from the ASME Code preservice inspection (PSI) requirements; therefore, confirmatory item (20),

the PSI review, is still incomplete.

5.2.4.3 Evaluation of Compliance With 10 CFR 50.55a(g)

The NRC staff and the nondestructive testing industry are in general agreement that performing ultrasonic testing of cast stainless steel is extreme N diffi-cult because of the poor acoustical properties of the materials of cons:ruction.

The staff has, therefore, adopted the practice of requesting confirmatt./y demon-strations to determine if the particular material of construction at a given plant has adequate acoustical properties to permit a valid examination with state-of-the-art instrumentation.

(1) Ultrasonic Examination Demonstrations On November 9, 1984, members of the NRC Headquarters and Region I staff and their consultants from Pacific Northwest Laboratories and Oak Rf dge National Laboratory observed a demonstration at the plant site of thL impability of the applicant's ultrasonic examination procedure and instrumentation to detect actual flaws and artificial reflectors.

After the staff and its consultants observed the examination of four welds during the demonstration on November 9, 1984, the staff reached the preliminary conclusion that the ultrasonic transducer was not penetrating the weld from the elbow side and the examination of the pipe was marginal because of the high electronic noise relative to the signal that must be interpreted. This was based on the observation that there was lack of a strong and consistent back-wall reflection around the circumference with the 0 transducer.

Millstone 3 SSER 3 5-3

The applicant disagreed that a persistent 0 back-wall signal is necessary to ensure that an adequate angle beam examination could be made or that this should be the determinant that a valid inspection is possible. The applicant also pointed out that during the same in plant demonstration, the ability of ultra-sonic procedure and equipment to detect significant flaws in fatigue-cracked specimens of cast stainless steel had been demonstrated. The staff observed this. The applicant has also provided evidence of penetration of the ultra-sound during the actual preservice examination by submitting data sheets in which signals from counterbore and inside diameter (ID) geometry are noted.

As a result of the November 9, 1984 demonstration, the staff submitted a detailed request fer additional infornution. The applicant supplied this information in letters of May 7, July 1, and July 2,1985. NRC Headquarters l and the Region I staff reviewed these responses. To resolve these matters, a Regional Inspector was sent to the site for a further demonstration on June 6, 1985.

The NRC Region I Inspection Report, No. 50-423/85-22, details that three repre-sentative samples of Millstone's centrifugally cast piping containing 15%

throughwall depth fatigue cracks were examined with the defects verified. In the in plant demonstration, one of the same welds tested in the November 9, 1984 demonstration was again ultrasonically examined. It was reported that an ID reflector was detected, but that transducer wobble, beam spread, and redirec-tion of the beam in the cast stainless grain structure caused its location to be incorrectly reported. This could perhaps explain ths disagreement about whether there was ultrasonic penetration during the November 9, 1984 demonstra-tion. The inspector concluded that the ultrasonic examination was adequate because a significant indication could have been detected using the applicant's

~

method of ultrasonic examination, if such a defect were present.

(2) Calibration Block The staff also questioned whether the one basic calibration block used by the applicant was adequate to establish the ultrasonic parameters for the inspec-tion of both the statically cast stainless steel elbows, the centrifugally cast stainless steel pipe, and also the ferritic steam generator nozzle-to-elbow welds. The applicant stated that the calibration block used for the PSI exami-nations meets the material and product form requirements of Section XI in that it was of the same specification (SA-351 Grade CF8A) as those materials in the main loop piping, which is one of the materials being joined by the weld (i.e.,

elbow to steam generator nozzle). With respect to the thickness of the cali-bration block, the wall thicknesses in the examination volumes of the pipes examined were within 1 inch of the calibration block thickness. The Regional Inspector determined that the calibration block was nominally 6 dB more attenua-tive than the cast pipe and 1 dB to 2 dB more attenuative than the cast elbow.

This means that the use of the block was generally twice as conservative in the scanning sensitivity as that required for the pipe and slightly more conserva-tive as that required for the elbow. This should adequately compensate for any nominal difference in thickness between the calibration block and the pipes examined.

l t

Millstone 3 SSER 3 5-4

(3) Future Inservice Inspections Northeast Nuclear Energy Co., as a member of the Westinghouse Owners Group, is participating in a development program designed to improve the current defect detection and characterization reliability for inspection of main coolant loop piping systems. The primary emphasis will be on field-usable inspection tech-niques and data processing systems. The program places emphasis on carefully prepared test samples which are designed to represent actual field conditions.

The applicant has stated it will perform examinations on main coolant loop welds using the best available technology. In addition, in preparing the Inservice Inspection Program, the applicant will select welds that show the best acoustical properties and have the best access for ultrasonic examination of the weld and required volume in accordance with staff positions.

5.2.4.4' Conclusions On the basis of the above review, the staff has reached the following conclu-sions regarding the ability to perform preservice and inservice inspections of the cast stainless steel pipe at Millstone 3:

(1) The examination procedure, instrumentation, and calibration standard meet the requirements of the ASME Boiler and Pressure Vessel Code,Section XI.

(2) The examination results were meaningful, i.e. , significant defects, if present, could have been detected in the volume required to be examined by the Code.

(3) The applicant's participation in the Westinghouse Owners Group primary coolant piping examination research and development programs and the appli-cant's commitment to use the best available technology to perform main coolan'. loop weld inspections provides assurance that future inservice inspections will be adequate.

(4) The basic objective of inservice inspections of the piping welds in the reactor coolant boundary is to perform a repetitive examination of a repre-sentative sample of welds in order to detect generic service-induced degradation. In addition to providing a baseline, the preservice examina-tions will identify those welds that will optimize the effectiveness of future inservice examinations. This will be implemented by the applicant's commitment to select welds that show the best acoustical properties and have ~the best access for ultrasonic examination of the weld and required volume for future Inservice Inspection Programs.

The staff has determined that the cast stainless steel pipe and elbow welds at Millstone 3, as demonstrated, have sufficiently good acoustical properties to permit a valid ultrasonic examination with state-of-the-art instrumentation.

Because the applicant has committed to perform future examinations using the best available technology and to select welds with the best acoustical proper-ties and access, the staff considers confirmatory item (19) regarding preservice examinations of welds in cast stainless piping to be resolved.

Millstone 3 SSER 3 5-5

5.4 Component and Subsystem Design 5.4.2 Steam Generators 5.4.2.2 Steam Generator Tube Inservice Inspection 5.4.2.2.1 Compliance With the Standard Review Plan The July 1981 edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (NUREG-0800) includes acceptance criteria recommending that the applicant perform inspections based on RG 1.83,

" Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes" and the applicable Standard Technical Specifications (NUREG-0452), " Standard Techni-cal Specifications for Westinghouse Pressurized Water Reactors."

When the Millstone 3 Safety Evaluation Report, NUREG-1031, was issued (July 1984), the applicant had not yet submitted its proposed Plant Technical Speci-fications. Therefore, the review of compliance with staff criteria was identi-fied as confirmatory item (21). The staff has now completed its review and finds that the applicant's Inservice Inspection Program complies with the recom-mendations in RG 1.83 and the applicable Standard Technical Specifications con-cerning provisions for a baseline examination, selection, and sampling of the tubes, inspection intervals, actions to be taken in the event defects are identified and reporting requirements. In addition, by letter of June 25, 1985, the applicant stated that to supplement the requirements for eddy current testing of steam generator tubes contained within the Technical Specifications, it would inspect the full length of the tubes for the initial 3% sample of an inservice inspection.

The staff concludes that the Inservice Inspection Program of steam generator

. tubes, confirmatory item (21), is acceptable and meets the inspection and test-ing requirements of GDC 32. This conclusion is based on the applicant follow-ing the recommendations in RG 1.83, and the Standard Technical Specifications (NUREG-0452), as reviewed by the staff and determined to be appropriate for this application.

o l

Millstone 3 SSER 3 5-6

l 6 ENGINEERED SAFETY FEATURES 1

6.2 Containment Systems 6.2.1 Containment Functional Design 6.2.1.1 Containment Structure The staff-indicated in the Millstone 3 SER that the applicant should provide a revised analysis of maximum external differential pressure since the original analysis was based on an assumed initial containment temperature representative of normal operation (100 F) rather than the limiting containment temperature oermitted by Technical Specifications (120 F). This reatter was classified as confirmatory item (23) pending receipt and review of the revised analysis.

In FSAR Amendment 14, the applicant provided information regarding a transient containment response analysis performed for Millstone 3, assuming inadvertent actuation of one train of the containment quench spray system. This analysis was performed using the CONTEMPT-LT computer program, and resulted in a minimum containment internal pressure of 8.03 psia at the time the quench spray was assumed to be manually deactivated (10 minutes). The' containment is designed for a minSum internal pressure of 8.0 psia.

The staff has reviewed the applicant's analysis and finds that the applicant's assumptions regarding initial containment conditions and quench spray system operation will tend to minimize the containment internal pressure. Further-more, the staff has performed a confirmatory analysis using the CONTEMPT-4 computer code, and assuming appropriate initial conditions, the actuation of both trains of the containment quench spray system, and no operator intervention.

The staff analysis produced a minimum containment internal pressure of 8.00 psia.

On the basis of the two analyses, the staff concludes that the Millstone 3 con-tainment can safely withstand the maximum external pressure resulting from an inadvertent spray actuation event.

As part of the review of the maximum external differential pressure, the staff has also considered the potential for excessive containment depressurization from operation of the containment vacuum pump system or the steam-jet air-ejector system. The contairiment vacuum pump system is operated periodically -

during normal operation to maintain the containment at or below the maximum permissible pressure; the steam-jet air ejector is used to evacuate the con-tainment during startup operations. The applicant has indicated that both of these systems are manually operated and are capable of reducing the containment air pressure to below the 8.0 psia design value if they are permitted to operate uninterrupted for a sufficient period of time. However, continued operation of the systems after establishment of the desired containment pressure is not con-sidered credible by the applicant on the basis that adequate control room indi-  ;

cation and administration controls are provided.

Millstone 3 SSER 3 6-1

On the basis of a preliminary review of the instrumentation and control provi-sions to preclude excessive containment depressurization, the staff has deter-mined that there are no containment-atmosphere low pressure alarms at Millstone 3.

(It should be noted that a statement in FSAR Section 9.5.10.3 incorrectly indicates that a low pressure alarm will annunciate in the main control room, notifying the operator that a low pressure condition exists; the applicant has indicated that it will correct the FSAR by deleting the statement). Hence, the control room operator would receive no direct indication if the containment pressure were reduced below the minimum permissible Technical Specification value (8.9 psia) or the design value (8.0 psia), other than through routine surveillance of the containment pressure at 12-hour intervals. Furthermore, the operator would not be aware of the operational status of the vacuum pump or air-ejector systems except for indicator lights on the main control board showing the position of the system isolation valves.

The staff will further evaluate the administrative controls regarding the use of the containment vacuum pump system and steam-jet air-ejector system before reaching a conclusion about the matter of n'aximum external differential pres-sure for Millstone 3. The staff will report the results of its review in a future supplement to the Millstone 3 SER.

6.2.3 Secondary Containment Functional Design In the Millstone 3 SER, the staff indicated that the supplementary leak col-lection and release system (SLCRS) satisfies the requirements of GDC 16 provided that (1) all openings in the secondary containment are under administrative control and (2) door position indicators and alarm capability are provided in the main control room. This matter was classified as confirmatory item (26) pending verification by the applicant of compliance with the above stipulations.

In subsequent communications with the staff, the applicant stated that status indication and alarms for all doors within the.SLCRS boundary at Millstone 3 will be provided at the security alarm station rather than in the main control room, and that the responsibility for reporting door status would be delegated to the Security Department. As part of this arrangement, all doors within the SLCRS boundary will be designated as security doors, and will be locked, alarmed, and access-controlled in accordance with 10 CFR 73.55 (see station security plan). To confirm that the control room has been advised of SLCRS door status by the Security Department, the applicant has committed to add a step to the appropriate plant operating procedure, and_to require approval from j the operation shift supervisor before blocking open secondary building doors l for temporary routing of hoses, cables, etc.

The staff has reviewed the information and commitments provided by the appli-cant regarding the secondary enclosure building. On the basis of its review, the staff concludes that the applicant's proposal to provide door position indicators and alarm capability in the security alarm station, and to delegate the responsibility for status reporting to the Security Department is acceptable provided that the applicant (1) identify the specific circumstances under which secondary building access. doors will be permitted to be maintained open during plant operation and (?) incorporate in the Technical Specifications or appro-priate plant procedures, additional controls to govern and limit the cumulative duration of such events. The basis for restricting the time that the access doors are open is twofold. Access to the enclosure building at Millstone 3 is Millstone 3 SSER 3 6-2

through single doors rather than dual doors in series, hence, the SLCRS effec- .

tiveness is compromised during enclosure building entry / exit, and whenever a I door remains temporarily open for routing of hoses, cables, etc. Also, location of access door position indicators and alarms at the security alarm station rather than in the main control room, as recommended in SRP Section 6.2.3, would necessitate a communications link between control room operators and security personnel in order to assess enclosure building door status, and could result in additional delays in taking appropriate action to close any doors l

that might be open. l l Should the applicant be unable to demonstrate that the likelihood of having an l open secondary building access door is sufficiently small, the staff will require additional analyses of response times for assessing door status, and the effect of open doors on secondary building performance and offsite releases, before reaching a conclusion on this matter. The staff will report the results of its continued review in a future supplement to the SER.

6.2.5 Combustible Gas Control System The staff indicated in the Millstone 3 SER that for initial containment atmosphere conditions which minimize the mass of air in the containment, the '

recombiners would need to be started earlier in the accident than assumed in the applicant's analyses, in order to maintain the containment hydrogen con-centration below 4 volume percent. Accordingly, the matter of recombiner actuation criteria was classified as confirmatory item (25) pending receipt and review of the plant procedures for actuating the recombiners.

By letter dated October 1, 1984, the applicant submitted the Emergency Operating Procedures Generation Package for Millstone, Unit 3. The procedures described therein call for recombiner actuation whenever the hydrogen concentration is greater than or equal to 0.5 volume percent. The applicant, by letter dated September 24, 1985, affirmed that this criterion is invoked in all sections of the procedures which involve recombiner actuation. The staff has performed a combustible gas control analysis for Millstone using the COGAP computer code, and assuming the appropriate limiting conditions for operation and the actuation of a single recombiner at 0.5 volume percent hydrogen. The staff's analysis i supports the applicant's conclusion that a single recombiner is sufficient to maintain the postaccident hydrogen concentration below 4 volume percent. On this basir, the staff concludes that confirmatory item (25) has been satisfac-torily resolved.

6.2.7 Fracture Prevention of Containment Pressure Boundary In previous input to the Safety Evaluation Report (NUREG-1031, July 1984), the staff indicated that ferritic materials that are used in_the containment pres-sure boundary will be reviewed to the fracture toughness criteria for Class 2 components identified in the Summer 1977 Addenda of Section III of the ASME Boiler and Pressure Vessel Ccde. For Class 2 components, the fracture toughness criteria in the Summer 1977 Addenda of Section III of the ASME Code permit the materials to be either Charpy V-notch tested at or below the lowest service tem-perature, evaluated to the nil-ductility transition temperature requirements'of Table NC-2311(a)-1 of the ASME Code, or evaluated using the fracture mechanics methods contained in Appendix G to the ASME Code.

Millstone 3 SSER 3 6-3

Ferritic materials that are in the Millstone 3 containment pressure boundary were procured to earlier fracture toughness criteria than those in the Summer.

1977 Addenda of the ASME Code. Hence, many materials were not Charpy V-notch tested at or below the lowest service temperature. To demonstrate that these materials meet the review criteria, the applicant used the fracture toughness data presented in NUREG-0577 and ASME Code Section III, Summer 1977 Addenda, Subsection NC. These data indicate that all materials meet the nil-ductility transition temperature criteria of Table NC-2311(a)-1 except for ferritic materials in the feedwater line.

The ferritic materials in the feedwater line were evaluated using the fracture mechanics methods in Appendix G to the ASME Code. The a bound reference stress intensity factor (26.78 ksi S.)pplicant used a lower for determining the allowable material fracture toughness. According to Appendix G, the reference stress intensity value used in the analysis would be applicable for ferritic material at 180 Fahrenheit degrees below the materials nil-ductility transition temperature. Additional fracture toughness data for materials with similar com-position and heat treatment as the Millstone 3 feedwater materials is reported j (Rolfe and Barsom, 1977). These data indicate that the reference stress inten- '

sity value assumed in the Appendix G fracture mechanics analysis is conservative.

. The crack sizes assumed in the evaluation were greater than what was permitted during the preservice examination of the component and allowed for flaw growth in service. The fracture mechanics analysis indicates that the ferritic mate-rials in the feedwater line would meet the safety margins recommended in Appen-dix G to the ASME Code. Additional fracture mechanics analysis performed by the applicant indicates that the critical crack size for brittle fracture would be greater than twice the depth used in the Appendix G analysis.

On the basis of review of the available fracture data and material fabrication histories, the use of correlations between metallurgical characteristics and material fracture toughness, and fracture mechanics analysis performed by the applicant, the staff concludes that the ferritic components in the Millstone 3 containment pressure boundary meet the fracture toughness requirements that are specified for Class 2 components by the 1977 Addenda of Section III of the ASME Code. Compliance with these Code requirements provides reasonable assurance that the Millstone 3 reactor centainment pressure boundary will behave in a nonbrittle manner, that the probability of rapidly propagating fracture will be minimized, and that the requirements of GDC 51 are satisfied. Confirmatory item (28) is resolved.

I Millstone 3 SSER 3 6-4

8 ELECTRIC POWER SYSTEMS 8.2 Offsite Electric Power System 8.2.2 Compliance With GDC 17 ,

8.2.2.1 Physical Separation of Offsite Circuits Within a Common Right-of-Way In a letter dated April 1, 1985, the applicant provided FSAR revisions as they will appear in FSAR Amendment 13. In this letter, the applicant stated that these revisions incorporate the response to NRC question Q430.4. This resolves confirmatory item (44) dealing with the inclusion of the information in the FSAR.

8.2.2.2 Physical Separation of Offsite Circuits Between Switchyard and Class 1E System In a letter dated April 1, 1985, the applicant provided FSAR revisions as they will appear in FSAR Amendment 13. In this letter, the applicant stated that these revisions incorporate the response to NF.? question Q430.5. This resolves confirmatory item (45) dealing with the inclusion of this information in the FSAR. During a site visit held on April 10 and 11, 1985, the staff observed that the A division cables from the normal station service transformer are routed in cable trays when they pass through the B division cable tunnel. This contradicts sheet 2 of FSAR Figure 8.3-7 which shows this cable in embedded con-duit. The applicant has indicated this figure will be corrected to resolve the discrepancy. This is acceptable.

8.2.2.5 Generator Rejection Scheme In the Millstone 3 SER, the staff stated that the surveillance and operability requirements for the generator rejection scheme (also termed the severe line outage detector, SLOD, scheme by the applicant) would be included in the tech-nical specifications. Subsequently, in a letter dated September 19, 1985, the applicant stated that the SLOD scheme is only one of many system features which in total give rise to a highly reliable transmission network. The applicant stated that although system reliability is largely affected by the quality and reliability of the hardware, it is also true that software elements are important to reliability, such as how VARs (volt amperes reactive) will be dispatched and how much spinning reserve and of what type in what location are available. This is handled by the Connecticut Valley Electric Exchange (CONVEX) which operates the transmission system (grid) within the State of Connecticut.

The applicant stated that CONVEX has operating procedures in p. lace, as well as its equivalent of limiting conditions for operation (complete with compensatory interim measures), to provide for any transmission system off-normal condition.

The applicant stated that no offsite Millstone switchyard transmission lines i will be removed from service when all three Millstone units are operating, except under forced outage conditions. If SLOD is not in service at that time, CONVEX will act to reduce Millstone site generation, within a matter of minutes, i Millstone 3 SSER 3 8-1

to preclude the possibility of an unstable condition arising. Similarly, if a line were forced out of service whenever SLOD is in service, CONVEX will act to optimize load flow and spinning reserve to prepare for the event that SLOD could operate and disconnect significant amounts of nuclear generation.

The staff finds that these provisions minimize, to the extent practical, the likelihood of the simultaneous failure of both offsite power sources and there-fore eliminate the need for Technical Specification limitations on the SLOD scheme. This closes confirmatory item (46).

8.3 Onsite Power Systems 8.3.1 Onsite AC Power System's Compliance With GDC 17 8.3.1.3 Description of Compliance With Position 1 of BTP PSB-1 In a letter dated April 1, 1985, the applicant provided FSAR revisions as they will appear in FSAR Amendment 13. One of these revisions includes the response to NRC question Q430.9 which the staff has previously reviewed and found accept-able. This resolves confirmatory item (50). As part of its review of the Millstone 3 Technical Specifications, the staff will ensure that the second level undervoltage protection setpoints are acceptable and the description of the logic matches that in the FSAR revision.

8.3.1.5 Adequacy of Station Electric Distribution System Voltage As part of the site visit on April 10 and 11, 1985, the staff reviewed tne results of the Millstone 3 voltage drop analysis and found the results to be acceptable; however, the staff noted that the grid sceltage limits necessary to maintain adequate plant voltages may not be outside of the normal grid voltage extremes. This could result in the grid being incapable of supplying adequate voltages to safety loads during periods when the grid is operating at its normal voltage extremes. The staff will pursue this item with the applicant and report its resolution in a future SER supplement.

NRC will verify the test results that substantiate the Millstone 3 voltage analysis. This item remains confirmatory pending completion of verification by the staff.

8.3.1.7 Diesel Generator Protective Relaying This item was identified as confirmatory in SER Section 8.3.1.7, but was not assigned a confirmatory item number in SER Table 1.4. The staff reviewed, with the applicant, Stone & Webster Engineering Corporation (SWEC) drawings 12179-ESK-8KK (Rev. 2),12179-ESK-8KF (Rev. 6),12179-ESK-5DS (Rev.11), and 12179-iSK-8KG (Rev. 6). The staff confirmed that the design for bypassing diesel gene ator protective relaying under accident conditions meets the staff position. This item is, therefore, considered complete.

l l 8.3.1.11 Diesel Generator Load Acceptance Test After Operation at No Load In SER Section 8.3.1.11, it was stated that the method by which the diesel gen-

! erator's no-load capability is considered i- the load acceptance tests would be i

Millstone 3 SSER 3 8-2

3 3

pursued with the applicant. In letters dated June 7,1985 and September 20, -

1985, the applicant provided information on how the deleterious effects of exteaded no-load operation will be minimized on the diesel generators. The  ;

staff evaluation of this response will be addressed in Section 9.5.4.1 of Supplement 4. _-

8.3.3 Common Electrical Features and Requirements _

m 8.3.3.1 Compliance With GDC 2 and 4 C 8.3.3.1.1 Submerged Electrical Equipment As a Result of a Loss-Of-Coolant G Accident In the Millstone 3 SER, the staff indicated that the applicant had identified 12 safety-related motor-operated valves that would be deenergized during normal c plant operat. ion. The staff further stated that it would require periodic veri- -_

fication in the Millstone 3 Technical Soecifications that these valves are in fact deenergized during normal plant or.eration. In a subsequent letter dated  ;'

September 19, 1985, the applicant provided a revised response on this subject which identified only four matcr-operated valves that required deenergization.

The staff has ensured that the requirement for deenergization of these four valves has been included in the Millstone 3 Technical Specifications. This is .-

acceptable. 8 -

8.3.3.3 Physical Independence - Compliance With GDC 17 8.3.3.3.2 Frequency of Cable Identification Markings This item was identified as confirmatory in SER Section 8.3.3.3.2, but was not 2 assigned a confirmatory item number in SER Table 1.4. The staff reviewed the -

cable's color code identification to determine that the 15-ft marking interval is sufficient to facilitate visual verificat. ion that the cables are installed in conformance with separation criteria. Becase the majority of cables were ""

continuously marked (solid color cable) so that they contrasted with black cables marked at approximately 15-ft intervals, the sta'f fcund that visual verification _

was not a problem. This item is, therefore, considered complete. -

8.3.3.3.8 Adequacy of Protection Provided Class IE Circuits From the Effects '

of Non-Class IE Circuits _

The applicant has performed tests and analyses to justify less than the minimum =

separation specified in IEEE Std. 384-1974 between Class 1E and non-Class 1E -

circuits. This is in accordance with Section 5.1.1.2 of IEEE Std. 384-1974 which allows the test and analysis approach. The tests and analyses that were _

performed are presented in Wyle Laboratory Test Report No. 47506-02 dated February 25, 1985. The following configurations were tested: m (1) A test in free air consisting of a fault cable inside a SWEC protective wrap (SILTEMP 188CH fabric), with three target cables in contact with _

outside of wrap. The purpose of the test was to demonstrate that a -

faulted cable enclosed within SWEC protective wrap does not affect exter- _

nal cables with 1-in. separation, which represents field installations '

of free air drops for cables going from Millstone 3 SSER 3 8-3

l (a) tray to tray (b) tray to conduit (c) conduit to conduit (d) tray / conduit to equipment (2) A test in free air consisting of a fault cable, in contact with a target cable which was wrapped in the SWEC protective wrap. The purpose of the test was to demonstrate that a faulted cable external to the SWEC protec-tive wrap does not affect the protected cable with 1-in. separation, which represents field installations of free air cable drops the same as in item 1 above.

(3) A test consisting of a dropout fault cable in the upper tray of a horizon-tal four-tray stack. The cable drops out of the upper tray, over the top of the covered tray below it, and proceeds down past the lower three trays.

The purpose of the test was to (a) Demonstrate the acceptability of a single, solid, nonventilated cable tray cover as a barrier with 1-in. separation from the tray cover.

(b) Demonstrate that a faulted dropout cable from a tray does not affect cables in trays below it.

M (4) A test consisting of a dropout target cable in a tray just below the upper tray of a horizontal four-tray stack. The cable drops out of the tray, runs along the ventilated cover of the covered tray below it, and proceeds down past the lower two trays. The fault cable is located just below the cover of the covered tray, and other target cables are located in the trays above and immediately below the covered tray. The purpose of the test was to (a) Demonstrate that a faulted cable in a tray with a single, ventilated cover does not affect cables in trays above it.

(b) Demonstrate that a faulted cable in a tray with a single, ventilated cover does not affect a dropout cable from a tray above it with 1-in. separation.

(5) A test consisting of two tests between a horizontal, four-tray stack and one vertical tray with 1-in. separation. The fault cable was placed in the vertical tray with four target cables in the horizontal trays during the first test. The positions of the fault cable and one of the target cables were interchanged for the second test. The purpose of the test was to demonstrate that, at a perpendicular crossing of a horizontal and vertical tray with a single cover on the vertical tray or a cover on the top and bottom on the horizontal tray (ventilated or nonventilated), a faulted cable in either tray does not affect the cable in the other tray.

(6) A test consisting of two tests between a horizontal cable tray and a parallel conduit mounted 1-in. above the fault cable at the top of the tray centerline. The fault cable was placed in the tray with three target cables in the conduit during the first test. The positions of the fault cable and tarcet cables were interchanged for the second test. The purpose of the test was to Millstone 3 SSER 3 8-4

(a) Demonstrate that a faulted cable enclosed in a conduit does not affect external cables with 1-in. separation. 4 (b) Demonstrate that a facited cable external to a conduit ancf with 1-in. separation does not affect cables internal to the conduit.

(7) A test between two horizontal conduits with no separatior. The fault cable was placed in the lower conduit with two target c o les in the upper conduit. The purpose of the test was to demonstrate that a faulted cable enclosed in a conduit does not affect cables in another conduit with 1/8-in. separation between conduits for low energy power, control, and in-strumentation circuits (applicant's "K," "C," and "X" service).

The staff has reviewed the test results applicable to the'above test configu-rations and finds them acceptable. The field instal;ations at Millstone 3 identified above are, therefore, also acceptable. These separations are applicable only between Class IE and non-Class 1E circuits.

8.3.3.3.10 Transformer Used as an Isolation Device In the Millstone 3 SER, the staff stated that by letters dated August 29, 1983, and June 12, 1984, the applicant provided results of tests and design provisions to ensure that non-Class 1E circuits are sufficiently isolated and will not cause unacceptable influence on any Class 1E circuits, and that the results of the staff review would be reported in a supplement to the SER. The staff has reviewed this information and finds that the results of tests and design pro-visions included the following items:

(1) The transformer is Class 1E and is protected by a fuse and a circuit breaker that are physically separated.

(2) The circuits from the transformer to the loads are protected by transformer.

outpet fuses and feeder circuit fuses.

(3) The output circuit of the transformer i:: run in dedicated conduit to the 120-volt, non-Class IE distribution panal.

(4) The loads are limited to control and instrument loads.

(5) The circuits from the 120-volt, non-Class 1E distribution panel are routed in raceways designated nonsafety; thus, circuits associated with redundant safety division are intermixed. The staff found this aspect of the design unacceptable.

(6) The test report, with respect to a bolted short on the output of the transformer demonstrated that associated Class IE circuits and power scpplies were not adversely affected.

(7) The test report, with respect to hot short, indicated electrical tran-sients may adversely affect Class 1E circuits. The staff fcund this aspect of the design unacceptable.

t Millstone 3 SSER 3 8-5

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1 Subsequently, bysletter dated July 18, 1984, the applicant committed to perform additional' testing to demonstrate the hot short capability of the isolation transformer. Given the reverse assumption that the isolation transformer

=x passes the additior.al testing, the staff concludes that the above design s provisions and thesisolation capability of the transformer meet the guidelines

. of Regulatory Guide (RG) 1.75 and are, therefore, acceptable. Given the assump-

• tion that the isolation transformer fails to pass the additional testing, the applicant has committed to either route the associated cables independently so that redundant associated cables are not intermixed or to remove the subject non-Class 1E circuits from Class 1E power sources. The staff concludes that either of these commitments would provide adequate protection for and indepen-dence between Class 1E circuits and are, therefore, acceptable.

During the staff's visit to the site en April 10 and 11, 1985, the applicant provided Appendix B of Test No. T3345BP002. The NRC staff found the results to be satisfactory. This item is, therefore, considered complete.

8.3.3.3.15 Ceordination of Breakers In the Millstone 3 $ER, the staff stated that for those series-connected circuit breakers used as isolation devices, periodic testing and calibration will be included in_the Millstone 3 Technical Specifications. The applicant, however, committed, by letter dated June 12, 1984, to periodically test and calibrate these devices to ensure that proper breaker coordination is maintained. This commitment on routine testing of the breakers acceptably resolves the staff's concern on breaker coordination and in accordance with staff practice on recent operating license reviews for similar situations, a technical specification covering testinq of these breakers will not be required.

8.3.3.3.16 Design Criteria of Associated Circuits From the Isolation Device to Load The staff confirmed that information presented by letter dated June 12, 1984 or by proposed FSAR Amendment 8 was included in FSAR Amendment 8 dated May 1984.

This item is, therefore, considered com rlete.

8.3.3.4 Compliance With the Guidelines of NUREG-0737 II.G.1, Emergency Power for Pressurizer Equipment The power supplies to the pressurizer power-operated relief valves (PORVs) and their associated block valves were originally taken from opposite power trains.

In the Millstone 3 SER, the staff indicated that this met the objective of TMI Action Plan Itam II.G.1 Clarification 2, but did not meet the recommendations of Branch Technical Position (BTP) RSB 5-2 for overpressurization protection while operating at low temperatures. In a letter dated April 11, 1985, the applicant committed to change this arrangement so that both series valves (PORV and block valve) will be powered from the same electrical division but from different power supplies. One PORV will receive power from 1 ~-volt vital direct current, and its associated block valve will receive pswer from the same train 480-volt alternating current emergency bus. The other PORV and block valve will have the same arrangement, but will be powered from the opposite division. This meets the requirements of both the TMI Action Plan (NUREG-0737) and BTP RSB 5-2. This resolves outstanding item (13).

Millstone 3 SSER 3 8-6

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8.3.3.7 Thermal Overload Protection Bypass In a letter dated September 19, 1985, the applicant provided justification for not conducting surveillance tasts on the bypasses of thermal overload (TOL) devices for motor-operated valves. The applicant explained that the engineered safety features (ESF) slave relay contact which bypasses the TOL function is the same contact that operates the valve for accident conditions and is, there-fore, tested by the same series of tests that verifies the operability of the valve actuation circuitry. The applicant proposes to change the wording for testing of the TOLs to correspond with that required for testing of the valve actuation circuitry so that one test would suffice for both requirements and an sdditional test would not be required. The staff finds this acceptable.

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Millstone 3 SSER-3 8-7

9 AUXILIARY SYSTEMS 9.5 Other Auxiliary Systems 9.5.3 Lighting System The Millstone 3 SER required the applicant to do the following:

For the auxiliary shutdown panel / purple switchgear roomi (1) The illuminatior. level provided by the purple emergency ac lighting system should be increased to a minimum of 10 ft-candles over the work area.

(2) Adequate ac lighting (a minimum of 10 ft-candles) should be provided in the auxiliary shutdown area from the other train of the ac light-ing system.

(3) The illumination level provided by the de lighting system should be increased to a minimum of 10 ft-candles in these areas of the purple switchgear room where work may be performed to restore ac power.

In letters dated June 29, 1984, and July 18, 1985, the applicant committed to items 1, 2, and 3.

For other safety-related area of the plant:

(1) Since the control room, the orange switchgear room, and the diesel generator room may require access during certain events so that ac power can be restored, the dc emergency lighting system illumination intensity shall be increased to a minimum of 10 ft-candles at those work stations where work Ir.uy be performed to restore ac power.

(2) The ac emergency lighting system illumination intensity shall be in-creased to a minimum of 10 ft-candles at the work station instead of an average of 10 ft-candles.

(3) The illumination intensity shall be increased to a minimum of 10 ft-candles at the panel surfaces ano at the work stations and 2 to 5 ft-l candles on the basis of the activity level for access and egress to safety-related plant areas.

In letters dated June 29, 1984 and July 18, 1986, the applicant addressed these concerns. The applicant committed to item 1. With regard to items 2 and 3, the applicant committed to storing portable battery-oowered lighting on site to

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[ supplement the emergency lighting, and provide the fe..owing for access / egress l lighting:

- Access / egress routes between manned work stations will be illuminated to 0.5 ft-candle average maintained.

Millstone 3.SSER 3 9-1

Slight hazards will be illuminated to 0.5 ft-candle minimum at the center point of the hazard.

High hazards will be illuminated to 2 ft-candles at the center point of the hazard.

The staff has evaluated the information and finds it acceptable; therefore outstanding item (15) is resolved.

On the basis of its review, the staff concludes that the various lighting sys-tems provided at Millstone 3 are in conformance with the standards, criteria, and design basis; can perform their design function; and, therefore, are accep!.abl e.

9.5.4 Emergency Diesel Engine Fuel Oil Storage and Transfer System 9.5.4.1 Emergency Diesel Engine Auxiliary Support Systems (General)

(2) Concrete Dust Control The Millstone 3 SER stated the following on concrete dust control:

It is the staff's position that, before initial startup, the con-crete floor and walls shall be painted with an appropriate paint or treated to minimize the generation of concrete dust. In a letter dated May 17, 1984, the applicant has committed to treat the floor slab with an appropriate sealant to preclude generation of concrete dust. The staff requires that sealant be applied before initial startup.

In a letter dated January 24, 1985, the applicant committed to complete the above by plant startup. Confirmatory item (56) is closed.

(4) Vibration of Instruments and Controls In a letter dated January 24, 1985, the applicant modified a commitment made in g a letter dated May 8, 1984 and found acceptable in the Millstone 3 SER. The applicant has now committed to the following for qualifying the engine-mounted instrumentation and controls for vibration:

(a) Actual vibrational levels of the equipment will be measured to confirm that they are within the tolerances specified as accept-able by the equipment manufacturers. The vibration levels will be measured during preoperational or qualification testing of the diesel generator units.

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(b) Equipment within the panels will undergo in-house vibration test-ing to ensure that it will remain operable, under actual equipment vibration levels, throughout the 18-month calibration period.

(c) Equipment which can not be qualified by one of the above.methoo will be replaced by items that can be qualified.

Millstone 3 SSER 3 9-2

(d) The engine skid-mounted panels will be removed from the engine skid and mounted as freestanding floor panels.

The applicant shall keep the staff advised about the qualification method being pursued. Vibration measurements and the complete quali-fication package will be submitted for staff approval. The program is to be completed by the end of the first refueling outage.

The above additional information expands on the original commitment; therefore, the staff continues to find the program acceptable. Confirmatory item (57) is closed pending review of the results of the above tests by the staff.

9.5.4.2 Emergency Diesel Engine Fuel Oil Storage and Transfer System With regard to sediment control, the Millstone 3 SER stated the following as license condition (4):

The design of the system as described above allows the diesel gener-ator day tank to be filled from either fuel oil storage tank. Thus, fuel oil can be drawn from one fuel oil storage tank while the other tank is being filled and then allowed to stand for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while the sediment settles out. This is an acceptable procedure for sedi-ment contrcl in the fuel oil system. Therefore, the staff requires that the plan'. operating procedures be modified to incorporate this filling procedure."

In a letter dated January 24, 1985, the applicant provided the following infor-mation about sediment control in his fuel oil storage tanks:

Correspondence with diesel fuel oil suppliers has indicated that turbulence caused by incoming fuel w1uld not be sufficient to dis-turb an exis+,ing sediment bed if the initial fuel oil level is greater than 4 feet. This level is based on anticipated sediment levels, pumping rate of fuel delivery trucks, and existing tank design. The fill line is such that it is terminated upon penetra-tion of the storage tank top. Therefore, any disruption of the fuel oil present in the tank will occur at the fluid surface and not in the area of the sediment layer.

The applicant will require in the plant operating procedures that refilling operations are started prior to the tank dropping below the 50% (5.2 feet) level. This requirement will ensure that un-acceptable sediment concentrations are not realized.

In the event that filling does not commence prior to reaching this 50% point, an adequate settling period will be provided for the recently filled tank, with transfer being accomplished form the alternate tank. Information provided by fuel oil suppliers indi-cates that a_1-hour settling period per foot of final product height is generally utilized. Therefore, the allowed settling time will be based on the final tank height at the conclusion of the filling operation. The settling period will be provided in the plant oper-ating procedures.

Millstone 3 SSER 3 9-3

The staff has evaluated the sediment control justification and associated operating procedures and finds them acceptable, and hence removes the license condition.

On the basis of its review, the staff concludes that the emergency diesel engine fuel oil storage and transfer system meets the requirements of GDC 2, 4, 5, and 17; the recommendations of NUREG/CR-0660; and the guidance of the cited RGs, SRP Section 9.5.4, and industry codes and standards. Thus, it can perform its design safety function and, therefore, is acceptable.

Piping and Associated Components The applicant was requested to define and provide the industry standards to which the emergency diesel engine fuel oil storage and transfer system piping and components were designed. In a letter dated June 29, 1984, the applicant provided the standards to which the engine-mounted piping was designed. The applicant stated that this piping and the associated components such as valves, fabricated headers, fabricated special fittings, and the like, are designed, manufactured, and inspected in accordance with the manufacturer's standards which are equivalent to ASME Boiler and Pressure Vessel Code Section III, Class 2 requirements as well as the guidelines and requirements of ANSI Stan-dard N45.2, " Quality Assurance Program Requirements for Nuclear Facilities,"

and 10 CFR 50, Appendix B. The engine-mounted piping and associated components are intentionally overdesigned (subjected to low working stresses) for the ap-plication, thereby resulting in high operational reliability. The staff finds the design of the engine-mounted piping and components, as stated, acceptable.

On the basis of its review, the staff concludes that the emergency diesel engine fuel oil storage and transfer system, with regard to piping design, meets the requirements of GDC 2, 4, 5, and 17 (Appendix A to 10 CFR 50); the recommendations of NUREG/CR-0660; and the guidance of the cited RGs, SRP sections, and industry codes and standards. Thus, it can perform its design safety fur.ction and, therefore, is acceptable.

9.5.5 Emergency Diesel Engine Cooling Water System Piping and Associated Components The applicant was requested to define and provide the industry standards to -

which the emergency diesel engine cooling water system piping and components l were designed. In a letter dated June 29, 1984, the applicant provided the standards to which the engine-mounted piping was designed. The applicant stated that this piping and the associated components such as valves, fabricated headers, fabricated special fittings, and the like are designed, manufactured, and in-spected in accordance with the manufacturer's standards which are equivalent to ASME Boiler and Pressure Vessel Code Section III, Class 2 requirements as wil as the guidelines and requirements of ANSI Standard N45.2, " Quality Assurance Program Requirements for Nuclear Facilities," and 10 CFR 50, Appendix 8. The engine-mounted piping and associated components are intentionally overdesigned (subjected to low working stressas) for the application, thereby resulting in high operational reliability. The staff finds the design of the engine-mounted piping and components, as stated, acceptable.

I 9-4 Millstone 3 SSER 3

On the basis of its review, the staff concludes that the emergency diesel engine cooling water system, with regard to piping design, meets the require-ments of GDC 2, 4, 5, and 17 (Appendix A to 10 CFR 50); the recommendations of NUREG/CR-0660; and the guidance of the cited RGs, SRP sections, and industry codes and standards. Thus, it can perform its design safety function and, therefore, is acceptable.

9.5.6 Emergency Diesel Engine Starting System In the Millstone 3 SER, the staff accepted the applicant's justification for de-laying the installation of the air dryers on the diesel generators until first refueling with the license condition (5): The air dryers shall be installed at the first opportunity but no later than before startup of the first refueling.

With regard to the second part of the license condition, namely blowdown of the air receivers and inspection of inline filters, the applicant by letters dated May 4, 1984 and January 7, 1985 provided additional information. The staff reviewed the information and found it acceptable and hence removes this portion of license condition (5).

The Millstone 3 SER stated the following:

Operating experience at two nuclear power plants has shown that dur-ing periodic surveillance testing of a standby diesel generator, ini-tiation of an emergency start signal (LOCA or LOP) resulted in the failure of the diesel to start and perform its function because of depletion of the starting air supply from repeated activation of the starting relay. This event occurred as the result of inadequate pro-cedures and from a failure in engine starting and control circuit logic to address.a built-in time delay relay to ensure the engine comes to a complete stop before attempting a restart. During the period that the relay was timing out, fuel to the engine was blocked while the starting air was uninhibited. This condition with repeated start attempts de-pleted starting air and rendered the diesel generator unavailable until the air system could be repressurized. This is an unacceptable operat-ing condition. The applicant was asked to review his procedures and/

or control system logic to ensure that this event will not occur at Millstone Uni,t 3.

In a letter dated June 29, 1985, the applicant stated that it reviewed the diesel control scheme and concluded that the event described above was not applicable to Millstone 3.

On the basis of its review, the staff concludes that the emergency diescl engine air-starting system meets the requirements of GDC 2, 4, 5, and 17; the guidance of the cited RGs and SRP Section 9.5.6; and the recommendations of I

NUREG/CR-0660 and industry codes and standards. Thus, it can perform its i design safety function and is, therefore, acceptable.

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! Piping and Associated Components-l The applicant was requested to define and provide the industry standards to which the emergency diesel engine starting system piping and components were designed.' In a letter dated June 29, 1984, the applicant provided the standards Millstone 3 SSER 3 9-5

to which the engine-mounted piping was designed. The applicant stated that this piping and the associated components such as valves, fabricated headers, fabricated special fittings, and the like, are designed, manufactured, and inspected in accordance with the manufacturer's standards which are equivalent to ASME Boiler and Pressure Vessel Code Section III, Class 2 requirements as well as the guidelines and requirements of ANSI Standard N45.2, " Quality Assur-ance Program Requirements for Nuclear Facilities," and 10 CFR 50, Appendix B.

The engine mounted piping and associated components are intentionally over-designed (subjected to low working stresses) for the application, thereby resulting in high operational reliability. The staff finds the design of the engine-mounted piping and components, as stated, acceptable.

Commensurate with the safety function performed by the air-starting system, the Millstone 3 SER required the following:

(1) All air-starting engine-mounted piping and components that are pressurized to high-energy pressures (275 psig or greater) during standby, starting, and/or operation will be designed seismic Category I, ASME Code,Section III, Class 3 (Quality Group C).

(2) All high-energy air-starting piping will be adequately restrained to pre-vent damage to other diesel generator piping, components, and equipment from pipe whip. Note: Seismic restraints and supports may not be adequate as pipe whip restraints.

In a letter dated April 11, 1985, the applicant provided its justification for not performing a high-energy line analysis of the air-starting system or classi-fying the entire air-starting system as ASME Code Section III, Class 3 (Quality Group C). The staff reviewed the applicant's justification for not performing a high-energy line analysis for the diesel generator air-starting system. On the basis of the available criteria and guidelines on high-energy failure, the staff finds the air-starting system for Millstone 3 acceptable as designed.

On the basis of its review, the staff concludes that the emergency diesel engine starting system, with regard to piping design, meets the requirements of General Design Criteria (GDC) 2, 4, 5, and 17 (Appendix A to 10 CFR 50); the recommenda-tions of NUREG/CR-0660; and the guidance of the cited RGs, SRP sections, and industry codes and standards. Thus, it can perform its design safety function and, therefore, is acceptable.

9.5.7 Emergency Diesel Engine Lif,ricating Oil System In the Millstone 3 SER, the staff found the justification provided by the ap-plicant for not preheating the rocker arm lubricating oil system acceptable.

However, the staff imposed license condition (6) based on the applicant's jus-tification: Upon actuation of the diesel generator low room temperature alarm, the room air temperature shall be' increased to 50 F or greater, or this may result in diesel generator being placed in a limiting condition for operation.

In a letter dated January 24, 1985, the applicant provided additional informa-tion. The applicant stated plant operating procedures will include the actions that would be taken to increase diesel generator room temperature upon actuation of low room temperature alarm. The staff finds this acceptable and removes license condition (6). This issue ~is closed.

Millstone 3 SSER 3 9-6

The Millstone 3 SER stated the following:

It is stated that the rocker arm lubricating oil reservoir level is monitored for high level and the level is maintained by a level con-trol valve. No mention is made of a reservoir low-level alarm. A failure of the level control valve to maintain lubricating oil level in the rocker arm reservoir could result in inadequate or no lubri-cating oil for the rocker arms, leading to diesel generator unavail-ability and/or failure. This is an unacceptable condition.

In letters dated May 17, 1984 and May 2, 1985, the applicant provided addi-tional information. The applicant stated that The rocker arm lube oil reservoir level will be checked, in accord-ance with the manufacturer's recommendations, prior to any manual start, biweekly on engines in standby and daily on operating engines.

The staff finds the above acceptable. This issue is closed.

On the basis of its review, the staff concludes that the emergency diesel engine lubricating oil system meets the requirements of GDC 2, 4, 5, and 17; the guidance of the cited RGs and SRP Section 9.5.7; and the recommendations of NUREG/CR-0660 and industry codes and standards. Thus, it can perform its design safety function and is, therefore, acceptable.

Piping and Associated Components The applicant was requested to define and provide the industry standards to which the emergency diesel engine lubricating oil system piping and components were designed. In a letter dated June 29, 1984, the applicant provided the standards to which the engine-mounted piping was designed. The applicant stated that this piping and the associated components such as valves, fabricated headers, fabricated special fittings, and the like, are designed,_ manufactured, and inspected in accordance with the manufacturer's standards which are equi-valent to ASME Boiler and Pressure Vessel Code Section III, Class 2 require-ments as well as the guidelines and requirements of ANSI Standard N45.2,

" Quality Assurance Program Requirements for Nuclear Facilities," and 10 CFR 50, Appendix B. The engine-mounted piping and associated components are intention-ally overdesigned (subjected to low working stresses) for the application, thereby resulting in high operational reliability. The staff finds the design of the engine mounted piping and components, as stated, acceptable.

On the basis of its review, the staff concludes that the emergency diesel engine lubricating oil system, with regard to piping design, meets the require-ments of GDC 2, 4, 5, and 17 (Appendix A to 10 CFR 50); the recommendations of NUREG/CR-0660; and the guidance of the cited RGs, SRP sections, and industry codes and standards. Thus, it can perform its design safety function and, therefore, is acceptable.

Millstone 3 SSER 3 9-7

9.5.8 Emergency Diesel Engine Combustion Air Intake and Exhaust System In the Millstone 3 SER, the staff found the intent of the operating procedure for the access hatch which is part of the diesel engine exhaust system accept-able. However, the staff proposed license condition (7) requiring the following to be included in the Plant Technical Specifications.

(1) In the event of a tornado alert or an ice storm, snow storm, or freezing rain storm forecast, the access hatch in the emergency diesel generator combustion exhaust system shall be opened and shall remain open until the event has passed.

(2) At least once a year, the access hatch shall be opened to verify opera-tion of the hatch, inspected for corrosion of parts (hinges, locking mechanisms, etc.), and maintained in an operable status by replacement of corroded parts, properly lubricated, pointed, etc.

In a letter dated January 7, 1985, the applicant provided additional informa-tion. The applicant stated that the design function of the access hatch is to provide an alternate exhaust path in the event that the exhaust stack is damaged by a tornado missile. The applicant also stated that because of the meteorolog-ical climate at the Millstone site, it is highly improbable that a tornado alert will occur coincidentally with a freezing rain or snow condition. Therefore, the requirement that the access hatch be opened when an ice storm, snow storm, or freezing rain storm forecast is received is felt to be unnecessary. The staff agrees with the applicant. The applicant will require, in an abnormal operating procedure, that the hatch be periodically inspected and if significant accumulation is observed, corrective action will be taken to ensure the hatch remains operable. The applicant will require that the access hatch be opened in the event of a tornado alert.

With regard to the second staff requirement, the applicant will require that measures be taken to ensure hatch operability be addressed in the preventive maintenance procedures. The access hatch will be opened at least once a year, inspected for corrosion of parts, and maintained in an operable status.

The staff finds the above acceptable and hence removes license condition (7).

This issue is closed.

Piping and Associated Components The applicant was requested to define and provide the industry standards to which the emergency diesel engine combustion air intake and exhaust system piping and components were designed. In a letter dated June 29, 1984, the ap-plicant provided the standards to which the engine-mounted piping was designed.

The applicant stated that this piping and the associated components such as valves, fabricated headers, fabricated special fittings, and the like, are designed, manufactured, and inspected in accordance with the manufacturer's standards which are equivalent to ASME Boiler and Pressure Vessel Code Sec-tion III, Class 2 requirements as well as the guidelines and requirements of ANSI Standard N45.2, " Quality Assurance Program Requirements for Nuclear Facilities," and 10 CFR 50, Appendix B. The engine-mounted piping and asso-ciated components are intentionally overdesigned (subjected to low working Millstone 3 SSER 3 9-8

I

. stresses) for the application, thereby resulting in high operational reliabil-ity. The staff finds the design of the engine-mounted piping and components, as stated, acceptable. l l

On the basis of its review, the staff concludes that the emergency diesel en-gine. combustion air intake and exhaust system, with regard to piping design, meets the requirements of GDC 2, 4, 5, and 17 (Appendix A to 10 CFR 50); the recommendations of NUREG/CR-0660; and the guidance of the cited RGs, SRP sec-tions, and industry codes and standards. Thus, it can perform its design safety function and, therefore, is acceptable.

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i Millstone 3 SSER 3 9-9

10 STEAM AND POWER CONVERSION SfSTEM l

10.4 Other Features l

10.4.2 Main Condenser Evacuation System In the Millstone 3 SER, the staff stated that the applicant has committed to provide an acceptable Offsite Dose Calculation-Manual (0DCM) which will describe

"; a detailed methodology for calculating noble gas, iodine, and particulate re-leases from the mechanical vacuum pump operation. In response to this commit-ment the applicant has provided calculational methods in the ODCM submitted with its letter dated June 19, 1985.

The main condenser air removal vacuum pumps are operated only during plant startup operation. The calculational methods specified in the ODCM utilize noble gas activity measured at the main condenser air ejector before shutdown

' to calculate the airborne noble gas releases during subsequent startup opera-tion of the mechanical vacuum pumps and actual grab samples from the mechani-cal vacuum pump exhaust will be used for calculating iodine and particulate releases. The staff finds the calculational methods specified in the ODCM to be acceptable.

10.4.3 Turbine Gland Sealing System a

In the Millstone 3 SER, the staff stated that the applicant has committed to provide an acceptable Offsit, Dose Calculation Manual (0DCM) which will describe a detailed methodology for calculating noble gas, iodine, and particulate re-leases from the turbine gland seal system operations. In response to this com-mitment, the applicant has provided calculational methods in the ODCM submitted with its letter dated June 19, 1985.

The turbine gland sealing system utilizes three steam sources: main, auxiliary, and/or extraction steam. The sys. tem is normally operated with extraction steam.

During low load operation (startdp and shut.oown), steam is taken from the main steamlines ahead of the turbine stop valves. When main steam is unavailable, the gland steam seal system is operated with auxiliary steam (nonradioactive).

The calculational methods utilize noble gas activity measured at the main con-denser air ejector using a ratio of steam flow rate into the main condenser to that into the gland seal condenser for estimating noble gas releases from the turbine gland sealing system. Iodine and particulate releases are based on those measured in the steam generator blowdown water using a carryover factor in the steam generator and a decontamination factor in the gland seal condenser.

The staff finds the calculational methods specified in the ODCM to be acceptable.

Confirmatory item (59) is, therefore, resolved.

Millstone 3 SSER 3 10-1

11 RADI0 ACTIVE WASTE MANAGEMENT 11.4 Solid Waste Management System Process Control Program In the Millstone 3 SER, the staff concluded that the proposed solid radwaste system at Millstone 3 is acceptable provided that the applicant provide an acceptable process control program (PCP) complete with a compliance program to meet the requirements set forth in 10 CFR 61.

The applicant submitted a~ process control program dated June 1985 for Millstone Units 1, 2, and 3. The PCP has not been revised from that submitted for Millstone Units 1 and 2 except to indicate that it is,also applicable to Millstone 3.

The applicant states in the PCP that all solidified radioactive wastes will meet the requirements set forth in 10 CFR 20 and 61. The staff finds, therefore, the Millstone PCP to be acceptable on an interim basis. The acceptability of this process control program is based on currently available guidelines, but a future revision should address full compliance with 10 CFR 61 when revised PCP guidance becomes available from NRC.

Confirmatory item (60) is resolved.

11.5 Process and Effluent Radiological Monitoring and Sampling Systems 11.5.2 Evaluation Findings Item II.F.1, Attachment 1, Noble Gas Effluent Monitor In the Millstone 3 SER, the staff concluded that the high range roble gas moni-toring systems to be ir-talled at Millstone 3 meet the requirements of clarifi-cation items (1), (2), (3), and (4), and Table II.F.1-1 in NUREG-0737 and meet the intent of the guidelines in RG 1.97, Rev. 3.

The staff further concluded, however, that the applicant should provide addi-tional information on the monitor calibration method, detector energy response characteristics, and calculational method to be used for converting instrument readings to release rate as a function of time after an accident.

The applicant has provided this additional information in its letter dated June 6, 1985. In it, the applicant states that the monitors will be calibrated using four check sources in the actual detector geometry during each refueling outage or every 18 months.

The staff has determined that the applicant has the capability to develop time-dependent correction factors that could be used in the Millstone 3 offsite dose calculational procedure. The correction factors will take into account (1) an Millstone 3 SSER 3 11-1

expected isotopic. composition Of neble gases as a function of time after an ac-cident, (2) detector counting Officiencies for each noble gas isotope, and (3) noble gas isotopic gamma energies. These correction factors can be used to con-vert the monitor readouts to actual radioactive gaseous releases.

The. staff finds acceptable the applicant's method of calibration and calcula-tional method to be used for converting instrument readings to release rates as a function of time after an accident. Therefore, the staff concludes the noble gas monitors installed at Millstone 3 F.eet the requirements of TMI Action Plan Item II.F.1, Attachment 1 and the guidelines provided in RG 1.97, Rev. 3. Con-firmatory item (61) is resolved.

Millstone 3 SSER 3 11-2 1

l 13 CONDUCT OF OPERATIONS 13.6 Physical Security Plan The Northeast Nuclear Energy Company acting as agent for the Northeast Utilities has filed with the Nuclear Regulatory Commission for the Millstone Nuclear Power Station Unit 3 the following security plans which have since been amended:

Physical Sew. ity Plan Contingency Plan Guard Training and Qualification Plan This supplement to the Millstone 3 SER summarizes how the applicant has provided for meeting the requirements of 10 CFR 73. This material is composed of a basic analysis that is available for public review, a protected Appendix, and a pro-tected response-force-size worksheet.

On the basis of a review of the subject documents and visits to the site, the staff has concluded that the protection provided by the Northeast Utilities against radiological sabotage at the Millstone site meets the requirements of 10 CFR 73. Accordingly, the protection provided will ensure that the health and safety of the public will not be endangered.

13.6.1 Physical Security Organization To satisfy the requirements of 10 CFR 73.55(b), Northeast Utilities has provided a physical security organization that includes a Security Shift Supervisor who ,

is on site at all times and who has the authority to direct the physical protec-tion activities. To implement the commitments made in the physical security, guard training and qualification plan, and the safeguards contingency plan, written security procedures specifying the duties of the security organization members are available for inspection. The training program and critical secur-ity tasks and duties for the security organization personnel are defined in the

" Millstone Nuclear Power Station Guard Training and Qualification Plan," which meets the requirements of 10 CFR 73, Appendix B, for the training, equipping, and qualification of the security organization members. The physical security plan and the training program provide commitments that preclude the assignment of any individual to a security-related duty or task before the individual has been trained, equipped, and qualified to perform the assigned duty in accordance with the approved guard training and qualification plan.

13.6.2 Physical Barriers In meeting the requirements of 10 CFR 73.55(c), the applicant has provided a

( protected area barrier which meets the definition of 10 CFR 73.2(f)(1). An isolation zone of at least 20 feet, to permit observation of activities along the barrier, is provided on both sides of the barrier with the exception of the locations listed in the protected Appendix. The staff has reviewed those locations and determined that the security measures in place are satisfactory and continue to meet the requirements of 10 CFR 73.55(c).

Millstone 3 SSER 3 13-1

Illumination of 0.2 ft-candle is maintained for the isolation zones, pro-tected area barrier, and external portions of the protected area. In areas where illumination of 0.2 ft-candle cannot be maintained, special procedures are applied as described in the Appendix.

The protected area is patrolled at random intervals to detect the presence of unauthorized persons, vehicles, and materials.

Identification of Vital Areas The Appencix contains a discussion of the applicant's vital area program and identifies those areas and items of equipment determined to be vital for pro-tection purposes. Vital equipment is located within vital areas which are located within the' protected area and which require passage through at least two barriers, as defined in 10 CFR 73.2(f)(1) and(2), with certain exceptions, to gain access to vital equipment. The staff has reviewed those exceptions and has determined that the barriers are sufficiently substantial to meet the intent of the two-barrier requirement.

Except as noted in the Appendix, vital area barriers are separated from the pro-tected area barrier. The control room and central alarm station are provided with multi purpose walls, doors, ceilings, floors, and windows. On the basis of these findings and the analysis set forth in the Appendix, the staff has concluded that the applicant's program for identifying and protecting vital equipment satisfies the regulatory intent. However, this program is subject to onsite validation by the staff in the future, and to subsequent changes if changes are found necessary.

13.6.3 Access Requirements In accordance with 10 CFR 73.55(d), all points of personnel and vehicle access to the protected area are controlled. The individual responsible for control-ling the final point of access into the protected area is located in a multi purpose structure. As part of the access control program, vehicles (except under emergency conditions), pers'onnel, packages, and materials entering the protected area are searched for explosives, firearms, and incendiary devices by electronic search equipment and/or by physical search.

I Vehicles admitted to the protected area, except applicant-designated vehicles, are controlled by escorts. Applicant-designated vehicles are limited to onsite station functions and remain in the protected area except for operational maintenance, repair, security, and emergency purposes. Positive control over these vehicles is maintained by personnel authorized to use the vehicles or by the escort personnel.

A photo-badge / key-card system, utilizing encoded information, identifies individuals that are authcrized unescorted access to protected and vital areas and is used to control access to thest areas. Individuals not authorized unescorted access are issued non photo badges that indicate an escort is

required. Access authorizations are limited to those individuals who have a need for access to perform their duties.

i Millstone 3 SSER 3 13-2

Unoccupied vital areas are locked and alarmed. Access to the reactor contain-ment is positively controlled to ensure that only authorized individuals are permitted to enter. In addition, all doors and personnel / equipment hatches into the reactor containment are locked and alarmed. Keys, locks, combinations, and related equipment are changed on an annual basis. In addition, when an individual's access authorization has been terminated because of lack of reliability or trustworthiness, or for poor work performance, the keys, locks, combinations, and related equipment to which that person had access are changed.

13.6.4 Detection Aids In satisfying the requirements of 10 CFR 73.55(e), the applicant has installed

-intrusion-detection systems at the protected area barrier, at entrances to vital areas, and at all emergency exits. Alarms from the intrusion detection system annunciate within the continuously manned central alarm station located in the protected area and within a secondary alarm station also located in the protected area. In addition, the central alarm station is constructed so that the walls, floors, ceilings, doors, and windows are bullet-resistant. The alarm stations are located and designed in such a manner so that a single act cannot interdict the capability of calling for assistance or responding to alarms. The central alarm station contains no other functions or duties that would interfere with its alarm response function.

The i. M sion-detection systems transmission lines and associated alarms annunciation hardware are line-supervised and tamper-indicating. Alarm annunciators indicate the type of alarm and its location when activated. An automatic indication of when the alarm system is on standby power is provided in the central alarm station.

13.6.5 Communications As required in 10 CFR 73.55(f), the applicant has provided for the capability of continuous communications between the central and secondary alarm station operators, guards, watchmen, and armed response personnel through the use of a conventional telephone system, and a security radio system. In addition, direct communication with the local law enforcement authorities is maintained through the use of a conventional telephone system and a two-way FM radio link.

All non portable communication links, except the conventional telephone system, are provided with an uninterruptible emergency power source backed up by diesel generators.

13.6.6 Test and Maintenance Requirements In meeting the requirements of 10 CFR 73.55(g), the applicant has established a program for testing and maintaining all intrusion alarms, emergency alarms, communication equipment, physical barriers, and other security-related devices or equipment. Equipment or devices that do not meet the design performance cri-teria or have failed to otherwise operate will be compensated for by appropriate compensatory measures as defined in the " Millstone Nuclear Power Station Physical Security Plan" and in onsite procedures. The compensatory measures defined in these plans will ensure that the effectiveness of the security system is not re-duced by failures or other contingencies affecting the operation of the security-related equipment or structures.

Millstone 3 3SER 3 13-3

Intrusion-detection systems are tested for proper performance at the beginning and end of any period that they are used for security. Such testing will be conducted at least once every seven days.

Communication systems for onsite communications are tested at the beginning of each security shift. Offsite communications are tested at least once each day.

Audits of the security program are conducted once every 12 months by the North-east Utilities Services Company (NUSCO) System Security Staff, which is inde-pendent of site security management and supervision. The audits, focusing on the effectiveness of the physical protection provided by the onsite security organization in implementing the approved security program plans, include, but are not limited to, a review of the security procedures and practices, system testing and maintenance programs, and local law enforcement assistance agreements.

The NUSCO System Security Staff prepares a report documenting its findings and recommendations and submits it to the applicant, Northeast Nuclear Energy Company (NNECO), for review and necessary action.

13.6.7 Response Requirements In meeting the requirements of 10 CFR 73.55(h), the applicant has provided for armed responders immediately available for response duties on all shifts con-sistent with the requirements of the regulations (protected Appendix). Consi-derations used in support of this number are listed in the protected Appendix.

In addition, liaison with local law enforcement authorities to provide addi-tional response support in the event. of security events has been established and documented.

The applicant's safeguards contingency plan for dealing with thefts, threats, and radiological sabotage events satisfies the requirements of 10 CFR 73, Appendix C. The plan identifies appropriate security events that could ini-tiate a radiological sabotage event and identifies the applicant's preplanning, response resources, safeguards contingency participants, and coordination activities for each identified event. Through this plan, upon the detection of abnormal presence or activities within the protected or vital areas, response activities using the available resources would be initiated. The response activities and objectives include the neutralization of the existing threat by requiring the response force members to interpose themselves between the adver-sary and the objective, instructions to use force commensurate with that used by the adversary, and authority to request sufficient assistance from the local law enforcement authorities to maintain control over the situation.

13.6.8 Employee Screening Program In meeting the requirements 00 10 CFR 73.55(a) to protect against the design-basis threat as stated in 10 CFR 73.1(a)(1)(ii), Northeast Utilities has pro-vided for an employee-screening program. Personnel who successfully complete the employee-screening program or its equivalent may be granted unescorted access to protected and vital areas at the Millstone site. All other personnel requiring access to the site are escorted by persons authorized and trained for escort duties and who have successfully completed the employee-screening program.

Millstone 3 SSER 3 13-4 l

The employee-screening program is based on accepted industry standards and in-cludes a background investigation, psychological evaluation, and a continuing observation program.

The plan also provides for a "grandfcther clause" exclusion which allows recog-nition of a certain period of trustworthy service with the utility or contrac-tor as being equivalent to the overall employee-screening program. The staff has reviewed the applicant's screening program against the accepted industry standards (ANSI N18.17-1973) and has determined that the Northeast Utilities program is acceptable. Confirmatory item (63) is resolved.

4 i

I i

i Millstone 3 SSER 3 13-5

---a , - .. .

14 INITIAL TEST PROGRAM In the Millstone 3 SER, the staff stated that resolution of issues identified during the initial test program review were submitted by letters dated April 19, 1984, and May 15, 1984, and that these reoponses were found acceptable. The responses have been included in an FSAR amendment and, therefore, confirmatory item (64)'is resolved.

Millstone 3 SSER 3 14-1

15 ACCIDENT ANALYSES 15.4 Reactivity and Power Distribution Anomalies 15.4.3 Rod Cluster Control Assembly Malfunctions In Section 15 of the Safety Evaluation Report (SER) for Millstone Unit 3, the

staff indicated that a potential controller problem existed for the dropped control rod event which could lead to the imposition of operating restrictions.

It also indicated anticipating receiving a detailed analysis showing that if the transient occurs, thermal limits would not be exceeded. In addition,'the SER indicated that '..estinghouse has developed a sol'ition for the problem via a new methodology for analyzing the event and has documented it in a topical re-port (WCAP-10298); this report and its methodology were evaluated and approved by the staff. The evaluation was enclosed in the NRC memorandum to F. Miraglia from L. Rubenstein, March 2, 1983, " Review of the Westinghouse Report ' Dropped Rod Methodology for Negative Flux Rate Trip Plants'." The solution requires a reactor cycle specific analysis showing that departure from nucleate boiling (DNB) limits will not be exceeded. The Millstone 3 FSAR has been revised in Amendment 12 to include a discussion of this analysis, and the results for cycle 1 operation indicate that DNB limits will be met for this cycle for both N and N-1 loop operation. Thus, operating limits will not be necessary for cycle 1. Each future reload cycle will require similar cycle-specific analysis as part of the normal reload analysis. Confirmatory item (67) is resolved.

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l l

Millstone 3 SSER 3 15-1

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16 TECHNICAL SPECIFICATIONS As a result of informacion contained in this supplement, the following corrections to SER Table 16.1, Technical Specification items, are made:

Item (8) and Item (l'-) are deleteo as Technical Specification requirements.

Table 16.1 hchnical Specification items (revised from SER Tatle 16.1)

Item Status SER Section (1) Securing of watertight docks into service

• 2.4.5 water cubicles (2) Intake water temperature monitoring

• 2.4.11.2 (3) Containment isolation valves in purge / vent

• 6.2.6 system test every 6 months (4) Periodic testing to ensure control room

• 6.4 leaktightness (5) ESF atmosphere cleanup system flow rate

• 6.5.1 (6) Periodic surveillance of the battery

• 8.3.2.1 float charge (7) Deenergizing of 12 safety-related motor-

• 8.3.3.1.1 operated valves during normal plant operation (8) Periodic testing and calibration of Deleted in 8.3.3.3.15 interrupting devices SSER 3 J

(9) Fuel oil quality and tests conformance

• 9.5.4.2 to Regulatory Guide 1.137 (10) Lubricating oil storage

• 9.5.7 (11) Capability to transfer lubricating oil

• 9.5.7 from storage to diesel generator (12) Lubricating oil inventory in storage

• 9.5.7 (13) Opening of access hatch in emergency Deleted in 9.5.8

diesel generator combustion exhaust system SSER 3 l (14) Open access hatch once per year

• 9.5.8 l

(15) Main steam stop and control valves and

• 10.2 reheat valves (16) Verification that containment isolation Added in 6.2.4 valves are closed SSER 2 (17) Operability of the loose parts monitoring Added in 4.4.5 system SSER 3

• Unchanged from Millstone 3 SER.

Millstone 3 SSER 3 16-1

17 QUALITY ASSURANCE 17.1 General FSAR Amendment 13 updated FSAR Section 17.2 so that it now references Revision 6 l of the " Northeast Utilities Quality Assurance Program Topical Report" (NU-QA-1) instead of an earlier revision. Revision 6 of NU-QA-1 has been reviewed and

! found acceptable by the staff.

17.2 Organization The applicant's organization for quality assurance (QA) is basically that reported in Section 17 of the Millstone 3 SER. However, in SER Figure 17.1: (a) the block shown as Vice President Purchasing and Materials Management, (b) the block shown as Vice President System Transmission & Distribution Engineering & Operations is now Vice President Transmission & Distribution Engineering & Operations, and (c) the Betterment Construction QA block is now under the Supervisor Construction QA block instead of the Supervisor Design and Operations QA block.

The staff finds the applicant's organization for QA continues to be acceptable.

17.4 Conclusion The staff's conclusion reported in tt' 'tillstone 3 SER--that the Northeast Util-ities description of the QA program for operations is in compliance with appli-cable NRC regulations, meets the requirements of Appendix 8 to 10 CFR 50, and is acceptable--is unchanged.

The applicant has acceptably revised either the FSAR or its topical report on QA to include the applicable responses to staff questions on QA and, as noted in Section 17.1 above, has updated the FSAR QA commitment to Revision 6. This closes confirmatory item (70), "QA Program Commitments," discussed on page 17-4 of the SER.

Outstanding item (19), "Q list," is closed, since differences between the staff and the applicant regarding the Q list have been eliminated.

Millstone 3 SSER 3 17-1

APPENDIX A CONTINUATION OF CHRONOLOGY OF THE NRC STAFF RADIOLOGICAL REVIEW OF THE MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 i August 8, 1985 Letter from applicant concerning seismic interaction program.

August 12, 1985 Letter from applicant concerning proposed revisions to Draft Technical Specifications.

August 12-16, 1985 Representatives of the NRC staff and its consultants from 4

EG&G Idaho audited equipment qualification files at Millstone 3. (Summary issued September 19, 1985.)

August 13, 1985 Letter to applicant requesting additional information the staff requires to prepare the final draft of the Millstone 3 Technical Specifications.

I August 13, 1985 Letter from applicant concerning Supplement No. 1 to NUREG-0737, Control Room Design Review.

i

August 15, 1985 Letter from applicant concerning Emergency Operating j Procedures Generation Package Response to NRC Comments.

J August 16, 1985 Letter to applicant transmitting Proof and Review copy of Draft Technical Specifications.

August 16, 1985 Letter from applicant requesting deviations from BTP CMEB 9.5-1.

August 20, 1985 Letter from applicant responding to NRC staff's requests for additional information regarding Draft 2 to Revision 0 of the Millstone Nuclear Power Station Emergency Plan.

I i

August 21, 1985 Letter from applicant concerning Technical Specifications.  ;

August 21, 1985 Representatives from NRC and Northeast Utilities meet in Bethesda, Md., to discuss applicant's response to Millstone 3 SER confirmatory item (28), Fracture Preven-tion of Containment Pressure Boundary (GDC 51).

August 26, 1985 Letter from applicant transmitting a response to SER open l

item (9).

i August 26, 1985 Letter from applicant transmitting Emergency Plan Imple-menting Procedures. ,

Millstone 3 SSER 3 1 Appendix A

j August 29, 1985 Letter from applicant requesting four deviations from BTP CMEB 9.5-1.

August 30, 1985 Letter from applicant transmitting the Environmental Pro-4 tection Plan.

i

, September 5, 1985 Letter from applicant responding to Materials Engineering Branch SER confirmatory item (28), GDC 51.

September 6, 1985 Representatives from NRC and Northeast Utilities meet in Bethesda, Md. , to discuss schedule for completion of Millstone 3 Draft Technical Specification review.

September 9, 1985 Letter from applicant concerning ATWS Rule Schedule required by 10 CFR 50.62(d).

September 10, 1985 Representatives from NRC and Northeast Utilities meet in Bethesda, Md., to review schematics and FSAR logic dia-j grams for N-1 loop operation.

, September 10, 1985 Letter to applicant requesting additional information.

September 12, 1985 Letter from applicant transmitting Draft 3 to Revision 0 of the Millstone Nuclear Power Station Emergency Plan.

September 13, 1985 Letter from applicant transmitting additional information concerning SER confirmatory item (61).

September 17, 1985 Letter from applicant transmitting additional information on the Seismic Interaction Program.

September 18, 1985 Letter from applicant requesting an exemption from ASME Code requirements for radiographic inspections.

September 19, 1985 Letter from applicant transmitting additional information on Technical Specifications, Proof and Review.

September 19, 1985 Letter to applicant transmitting two copies of Supple-ment 2 to NUREG-1031 (Safety Evaluation Report).

September 20, 1985 Letter from applicant transmitting additional information on revised response to SER open item (2.2) per telephone

. conversation with staff.

September 20, 1985 Letter from applicant concerning Technical Specifications, Proof and Review.

September 20, 1985 Letter from applicant concerning hot participation experience.

September 20, 1985 Letter from applicant responding to NRC comments on Revi-sion 0 to the Physical Security Plan.

Millstone 3 SSER 3 2 Appendix A

September 23, 1985 Letter from applicant concerning Technical Specifications, Proof and Review.

September 23, 1985 Letter from applicant concerning Material Engineering Branch Preservice Inspection Program - Class 3.

September 24, 1985 Letter to applicant concerning use of ASME Code Case N-249-4.

September 24, 1985 Letter from applicant concerning Technical Specifications, Proof and Review.

September 25, 1985 Representatives from NRC and Northeast Utilities meet in Bethesda, Md., to discuss applicant's responses to open items resulting from the site audit held on Auaust 19-23, 1985.

September 25, 1985 Letter from applicant concerning Technical Specifications, Proof and Review.

September 25, 1985 Letter to applicant concerning Preliminary Safety Evalua-tion to be included in SER Supplement No. 3.

September 26, 1985 Letter from applicant concerning Technical Specifications, Proof and Review.

September 26, 1985 Letter from applicant responding to Questions 220.39 through 220.41, Containment Liner Review.

September 26, 1985 Letter to applicant concerning failure to submit informa-tion requested by the staff in a timely manner.

September 26, 1985 Letter from applicant transmitting FSAR Amendment 15.

September 26, 1985 Letter from applicant concerning revised response to Question 492.7.

September 26, 1985 Letter to applicant responding to Materials Engineering Branch SER confirmatory item (16).

September 26, 1985 Letter from applicant concerning applicant's application for exemption from Appendix J.

September 27, 1985 Letter to applicant transmitting 20 copies of SER Supple-ment 2 (NUREG-1031).

September 27, 1985 Letter from applicant concerning Project Key Events Schedule.

September 27, 1985 Letter from applicant concerning Seismic Interaction Program.

Millstone 3 SSER 3 3 Appendix A

September 30, 1985 Letter from applicant concerning Technical Specifications, Proof and Review.

September 30, 1985 Letter from applicant transmitting a response to NRC Question 260.59.

September 30, 1985 Letter from applicant transmitting a response to SER confirmatory item (26).

September 30, 1985 Letter.from applicant responding to SER confirmatory item (7).

0ctober 1, 1985 Letter from applicant responding to SER open item (14.3) request for deviations from BTP CMEB 9.5-1.

October 1, 1985 Letter from applicant transmitting responses to SER open item (14.2) and (14.7). .

October 1, 1985 Letter from applicant responding to SER confirmatory item (66).

October 1, 1985 Letter from applicant transmitting the Final Inservice Test-Program for Pumps and Valves.

October 1, 1985 Letter from applicant transmitting a revised response to Question 480.37.

October 1, 1985 Letter from applicant concerning Safety, Relief and Block Valve Adequacy Report.

October 1, 1985 Letter from applicant transmitting a revision to Millstone 3 Boron Dilution Analysis.

October 2, 1985 Representatives from NRC and Northeast Nuclear Energy Company meet in Bethesda, Md., to discuss activities to be completed before the low power license is issued.

(Summary issued October 8, 1985.)

October 4, 1985 Letter from applicant responding to NRC Question 480.7.

October 4, 1985 Letter from applicant concerning Technical Specifications, Proof and Review (B11785).

October 4, 1985 Letter from applicant concerning Technical Specifications, Proof and Review (811779).

October 7, 1985 Letter from applicant concerning Technical Specifications, Equipment Protective Devices.

October 7, 1985 Letter to applicant concerning preliminary safety evalua-tion to be included in SER Supplement 3 for Millstone 3.

7.

' Millstone 3 SSER 3 4 Appendix A

October 7, 1985 Letter from applicant concerning design verification activities.

October 7, 1985 Letter from applicant concerning Technical Specifications, Proof and Review.

October 8, 1985 Letter from applicant transmitting responses to Seismic Qualification Review Team (SQRT) audit.

October 9, 1985 Letter from applicant concerning Technical Specifications, Proof and Review.

October 9, 1985 Representatives from NRC and Northeast Utilities meet in Bethesda, Md., to discuss the Millstone 3 Seismic Inter-action Program.

October 10, 1985 Letter from applicant transmitting responses to pump and valve operability review team (PVORT) audit question:,.

October 15, 1985 Letter from applicant transmitting an application for schedular exemption from GDC 2.

October 15, 1985 Letter from applicant transmitting revised response to SER confirmatory item (27).

October 16, 1985 Letter to applicant requesting additional information on Technical Specifications 2282.400-568, 221.180-127, and 2472.110-611.

October 17, 1985 Letter to applicant concerning Emergency Operating Pro-cedures and Operator Training at Millstone 3.

October 17, 1985 Representatives from NRC and Northeast Utilities meet in Bethesda, Md., to discuss information needed to complete the staff's review of the DCROR.

October 17, 1985 Letter to applicant concerning Probabilistic Safety Study for Millstone 3.

October 18, 1985 Letter to applicant transmitting the Final Draft Technical Specifications for Millstone 3.

October 21, 1985 Letter from applicant concerning plant readiness for fuel load and request for operating license.

October 23, 1985 Letter to applicant transmitting the results of safety parameter display system audit for Millstone 3.

October 24, 1985 Representatives from NRC and Northeast Utilities meet in Bethesda, Md., to discuss the applicant's Seismic Inter-action Program at Millstone 3.

Millstone 3 SSER 3 5 Appendix A

October 25, 1985 Letter to applicant requesting additional information concerning Licensee Qualification Branch and reouesting a reply by November 1, 1985.

October 28, 1985 Letter to applicant transmitting draft license to Millstone 3.

October 28-29, 1985 Representatives from NRC and Northeast Nuclear Energy Company meet at the Millstone 3 site to permit NRC manage-ment to assess the operational readiness of Millstone 3.

Millstone 3 SSER 3 6 Appendix A

APPENDIX B BIBLIOGRAPHY Northeast Nuclear Energy Company, "Probabilistic Safety Study," submitted by letter dated July 27, 1983, from W. G. Counsil, NNECo., to B. J. Youngbloud, NRC,

Subject:

Millstone Nuclear Power Station, Unit 3.

-- , " Millstone Nuclear Power Station Physical Security Plan," submitted by letter dated May 31, 1985, from J. F. Opeka, NNECo., to B. J. Youngblood, NRC.

-- , letter dated June 5, 1985, from J. F. Opeka, NNECo., to B. J. Youngblood, NRC,

Subject:

" Quality Assurance Program Topical Report," Rev. 6.

-- , letter dated June 19, 1985, from J. F.0peka, NNECo., to B. J. Youngblood, NRC,

Subject:

Process Control Program.

-- , " Millstone Nuclear Power Station Guard Training and Qualification Plan," March 31, 1981.

-- , " Millstone Nuclear Power Station Contingency Plan," June 8, 1982.

Ravindra, M. K., and others, Structural Mechanics Associates, "A Program to Determine the Capability of the Millstone 3 Nuclear Power Plant to Withstand Seismic Excitation Above the Design SSE," NTS/SMA 206.01-R2, November 1984.

Polfe, S. T., and J. M. Barsom, " Fracture and Fatigue Control in Structures, Applications of Fracture Mechanics," Prentice-Hall, Englewood Cliffs, N. J.,

1977.

Stone & Webster Engineering Corporation " Evaluation of Anchor Stud Spacing, Containment Structure Steel Liner," NERM-59,1984.

-- , " Stability and Design Evaluation of West Retaining Wall," transmitted May 15, 1984.

U.S. Nuclear Regulatory Commission, memorandum from L. Rubenstein to F. Miraglia,

" Review of the Westinghouse Report 'Oropped Rod Methodology for Negative Flux Rate Trip Plants," March 2, 1983.

-- , Office Technical Position Paper, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Application."

-- , NUREG-0452; " Standard Technical Specifications for Westinghouse Pressurized Water Reactors," June 1978 (revised periodically).

-- , NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," October 1979.

Millstone 3 SSER 3 1 Appendix B

-- , NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

-- , NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," July 1981.

-- , NUREG-1152, " Millstone 3--Risk Evaluation Report," to be published; draf t, August 1985.

-- , NUREG/CR-0660, " Enhancement of On-Site Emergency Diesel Generator Reliability," February 1979.

-- , NUREG/CR-3756, " Seismic Hazard Characterization of the Eastern United States: Methodology and Interim Results for Ten Sites," April 1984.

-- , Region I, letter dated August 11, 1985, from S. Ebneter, NRC, to J. F. Opeka, NNECo. ,

Subject:

Inspection Report 50-423/85-22.

Westinghouse, WCAP-8691, " Fuel Rod Bow Evaluation," Revision 1, November 9, 1981.

-- , WCAP-9401, " Verification Testing and Analysis of the 17 x 17 Optimized Fuel Assembly," August 1981 (proprietary); WCAP-9402 (nonproprietary version).

-- , WCAP-10298, " Dropped Rod Methodology for Negative Flux Rate Trip Plants," January 20, 1982.

Wyle Laboratory, Test Report No. 47506-02, submitted by letter dated March 22, 1985, from W. G. Counsil, NNEC0, to 8. J. Youngblood, NRC,

Subject:

Millstone Nuclear Power Station, Unit 3, Test Report on Electrical Separation Verification Testing, February 25, 1985.

Millstone 3 SSER 3 2 Appendix B

APPENDIX D ABBREVIATIONS ASME American Society of Mechanical Engineers ATWS anticipated transient (s) without scram BTP branch technical position CFR Code of Federal Regulations CONVEX Connecticut Valley Electric Exchange DNB departure from nucleate boiling DNBR departure from nucleate boiling ratio ECCS emergency core cooling system EGE emergency generator enclosure ESF engineered safety features FSAR Final Safety Analysis Report GDC general design criteri(on)(a)

ID inside diameter IE Office of Inspection and Enforcement LOCA loss-of-coolant accident LOP loss of offsite power LPMS loose parts monitoring system NNECO Northeast Nuclear Energy Company (applicant)

NRC U.S. Nuclear Regulatory Commission NUSCO Northeast Utilities Services Company OBE operating-basis earthquake ODCM Offsite Dose Calculation Manual PCP process control program PORV power-operated relief valve PSI preservice inspection PSS Probabilistic Safety Study QA quality assurance RCPB reactor coolant pressure boundary RC5 reactor coolant system RG regulatory guide Millstone 3 SSER 3 1 Appendix 0

SER Safety Evaluation Report SLCRS supplementary leak collection and release system SLOD severe line outage detector SRP Standard Review Plan SSE safe shutdown earthquake SSER Safety Evaluation Report supplement SWEC Stone & Webster Engineering Corp.

TMI Three Mile Island TOL thermal overload VARs volt amperes reactive

' Millstone 3 SSER 3 2 Appendix D

APPENDIX F NRC STAFF CONTRIBUTORS Name Title Branch W. Belke Quality Assurance Engineer-- Quality Assurance Nuclear J. Chen Geotechnical Engineer Structural and Geotechnical Engineering N. Chokshi Structural Engineer Structural and Geotechnical Engineering H. Conrad Senior Materials Engineer Materials Engineering B. Elliot Materials Engineer Materials Engineering C. Gaskin Safeguards Engineer Power Reactor Safeguards Licensing R. Giardina Mechnical Engineer Power Systems A. Gill Nuclear Engineer Core Performance J. Lazevnick Electrical Engineer Power Systems J. Lee Nuclear Engineer Meteorology and Effluent Treatment R. Palla Containment Systems Engineer Containraent Systems D. Smith Materials Engineer Materials Engineering J. Spraul Quality Assurance Engineer Quality Assurance J. Tsao Materials Engineer Materials Engineering Millstone 3 SSER 3 1 Appendix F

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NUREG-1031 BIBLIOGRAPHIC DATA SHEET Supplement No. 3 i r .. w.,

1 TITLE .s3 SueTeTL Safety Eva ation Report Related to the operation of fiillstone il lear Power Station, Unit No. 3

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The Safety Evaluation Report issued in A us 984 provided the results of the NRC staff review of Northeast fluclear Energy Compan 's pplication for a license to operate the Millstone fluclear Power Station, Unit No. . Supplement No. I to that renort, issued in fiarch 1985 updated the information cont .d in the Safety Evaluation Report and addressed the ACRS Report issued on Septe er 0, 1984. Supplement flo. 2 issued in September 1985 updated the information c taint- in the Safety Evaluation Renort and Supplement No.1 and addressed prior unr solved tems.

This Supplement, No. 3, provides more 2 cent infor tion renarding resolution or updating of sone of the open and confirmatory f ens and lice e conditions identified in the Safety Evaluation Report.

The facility is located in Waterfor Township, New Con County, Connecticut.

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