ML20154C299
| ML20154C299 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/31/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1031, NUREG-1031-S05, NUREG-1031-S5, NUDOCS 8603050091 | |
| Download: ML20154C299 (76) | |
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NUREG-1031 Supplement No. 5 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 Office of Nuclear Reactor Regulation January 1986 j%f"
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
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- 2. The Superintendent of Documents, U.S. Government Printing Of fice, Post Office Box 37082, Washington, DC 20013-7082
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, it is not intended to be exhaustive.
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- pro
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NUREG-1031 Supplement No. 5
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_ -. . - r Safety Evaketion Report related to the operation of Millstone Nuclear Power Station, Unit.No. 3 Docket No. 50-423 Northeast Nuclear Energy Company
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O.5. Nuclea Regulatory Commission Office of Nuclear Reactor Regulation ,
January 1986
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ABSTRACT 1
I This report supplements the Safety Evaluation Report (NUREG-1031) issued in :
July 1984, Supplement 1 issued in March 1985, Supplement 2 issued in September 1985, Supplement 3 issued in November 1985, and Supplement 4 issued in November '
1985 by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory j Commission with respect to the application filed by Northeast Nuclear Energy
- Company (licensee 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 j Long Island Sound.
The supplement provides more recent information regarding resolution of license conditions identified in the SER. Because of the favorable resolution of the items discussed in this report, the staff concludes that Millstone Nuclear Power
- Station, Unit No. 3, can be operated by the licensee at power levels greater than 5% without endangering the health and safety of the public.
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CONTENTS P
P_a.ge ABSTRACT......................................................... iii i
1 INTRCDUCTION AND GENERAL DESCRIPTION OF PLANT............... 1-1 1.1 Introduction........................................... 1-1 1.5 Outstanding Items...................................... 1-1 l 1. 6 Confirmatory Items..................................... 1-1 l 1.7 License Condition Items................................ 1-2 3 DESIGN CRITERIA FOR STRUCTURE, SYSTEMS, AND COMPONENTS...... 3-1 3.6 Protection Against Dynamic Effects Associated With The Postulated Rupture of Piping........................... 3-1 3.6.2 Determination of Rupture Locations and Cynamic Effects Associated With the Postulated Rupture of Piping....................................... 3-1 3.6.2.1 Bases for the Elimination of Arbitrary Intermediate Pipe Breaks............... 3-3 3.6.2.2 Staff Evaluation of the Bases for the Elimination of Arbitrary Breaks. . . . . . . . 3-4 3.6.2.3 Conclusion............................. 3-9 3.9 Mechanical Systems and Components...................... 3-9 3.9.2 Dynamic Testing and Analysis of Systems, '
Components, and Equipment....................... 3-9 3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical i Equipment.............................................. 3-11 l 4 REACT 0R..................................................... 4-1 4.4 Thermal-Hydraulic Design............................... 4-1 4.4.4 Operating Abnormalities......................... 4-1 4.4.4.2 Crud Deposition and Flow Measurement Uncertainty............................ 4-1 4.4.8 Instrumentation for Detection of Inadequate Core Cooling.................................... 4-2 Millstone 3 SSER S v i
CONTENTS (Continued)
.P, age 4.4.8.1 Clarification of Requirements.......... 4-2 4.4.8.2 Inadequate Core Cooling
. Instrumentation System Design.......... 4-2 1 4.4.8.3 Subcooling Margin Monitor.............. 4-2 4-3 d
4.4.8.4 Core Exit Thermocouples ...............
4.4.8.5 Reactor Vessel Level Instrumentation System................................. 4-3 4.4.8.6 Evaluation and Conclusions............. 4-3 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS................ 5-1 5.2 Integrity of Reactor Coolant Pressure Boundary. . . . . . . . . 5-1 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing.......................... 5-1 6 ENGINEERED SAFETY FEATURES.................................. 6-1 6.2 Containment Systems.................................... 6-1 6.2.1 Containment Functional Design................... 6-1 6.2.1.1 Containment Structure.................. 6-1 6.2.1.2 Subcompartment Analyses................ 6-2 6.2.1.4 Mass and Energy Release Analysis for
- Postulated Secondary System Pipe Ruptures............................... 6-2 6.6 Irservice Inspection of Class 2 and 3 Components....... 6-2 j 6.6.3 Evaluation of Ccmpliance With 10 CFR 50.55a(g).. 6-2
- 9 AUXILIARY SYSTEMS........................................... 9-1 9.5 O the r Aux i l i a ry Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1
- 9. 5.1 Fire Protection................................. 9-1 9.5.1.5 Fire Detection and Suppression......... 9-1 1
9.5.3 Lighting System................................. 9-1 j i
13 CONDUCT OF 0PERATIONS....................................... 13-1 13.1 Organizational Structure of Applicant. . . . . . . . . . . . . . . . . . 13-1 l 13.1.2 Plant Staff..................................... 13-1 i
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CONTENTS (Continued) 1' P. age t
i 13.5 Station Administrative Procedures...................... 13-5 i
13.5.2 Operating and Maintenance Procedures............ 13-5 l 13.5.2.3 Reanalysis of Transients and Acciaents;
' Development of Emergency Operating Procedures............................. 13-5 i
.i 15 ACCIDENT ANALYSES........................................... 15-1 I 15.3 Decrease in Reactor Coolant Flow Rate.................. 15-1
- 15.3.3/15.3.4 Reactor Coolant Pump Rotor Seizure and Shaft Break............................. 15-1 15.6 Decrease in Reactor Coolant Inventory.................. 15-1 1
i 15.6.3 Steam Generator Tube Rupture.................... 15-1 17 QUALITY ASSURANCE........................................... 17-1 l 17.6 Indepe.1 dent Design Verification........................ 17-1 17.6.4 Staff Assessment................................ 17-1 i 17.6.4.3 Hazard Analysis Program................ 17-1 18 HUMAN FACTORS ENGINEERING................................... 18-1 l 18.2 Safety Parameter Display System........................ 18-1 i APPENDICES
< A CONTINUATION OF CHRONOLOGY OF THE NRC STAFF RADIOLOGICAL REVIEW l OF THE MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3
- B REFERENCES ;
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4 F NRC STAFF CONTRIBUTORS l N PRESERVICE INSPECTION RELIEF REQUEST EVALUATION t
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l CONTENTS (Continued)
TABLES (
P,, age 1.1 Listing of outstanding items (revised from SER Table 1.3).................................................. 1-3
- 1. 2 Listing of confirmatory items (revised from SER Table 1.4).................................................. 1-5
- 1. 3 Listing of license conditions (revised from SER Table 1.5).................................................. 1-10 15.1 Locked rotor accident assumptions........................... 15-3 15.2 Locked rotor accident dose calculations. . . . . . . . . . . . . . . . . . . . . 15-3 l
Millstone 3 SSER 5 viii
1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction In July 1984, the U.S. Nuclear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER)(NUREG-1031) on the application filed by Northeast Nuclear Energy Company (NNECO, the licensee), 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 1 (SSER 1), in September 1985 by Supplement 2 (SSER 2), in November 1985 by Sup-
- plement 3 (SSER 3) and in November 1985 by Supplement 4 (SSER 4); these docu-mented the resolution of several outstanding and confirmatory items and license conditions in further support of the licensing activities. The present report, ,
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Supplement 5 to the SER (.SSER 5), provides more recent information regarding the resolution or updating of some of the outstanding and confirmatory items and license conditions identified in the SER and its supplements, and supports the license for operation at power levels greater than SL
- Each of the following sections or appendices is numbered the same as the cor-responding SER section or appendix that is being updated. Each section is sup-plementary 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 processing of the Millstone 3 application. Correspondence between the li-censee and the NRC staff is listed chronologically in this appendix. Appen-dices B and D list references and abbreviations, respectively. Appendix F lists i principal staff members who contributed to this supplement. Appendix N adds to i
the SER the staff's evaluation of the licensee's request for relief from certain ASME Code requirements.
, Copies of this SER supplement are available for inspection at the NRC Public Document Room at 1717 H Street, N.W., Washington, D.C., and at the local Public Document Room of the Waterford Public Library, Rope Ferry Road, Route 156, Waterford, Connecticut.
The NRC Project Manager for Millstone 3 is Ms. Elizabeth L. Doolittle. Ms. Doo-little may be contacted by writing to her at the Division of PWR Licensing-A, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555.
- 1. 5 Outstanding Items
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The staff identified certain outstanding items in the SER that had not been re-
, solved with the licensee. The status of these items is listed Table 1.1 (an i updated version of SER Table 1.3). As Table 1.1 indicates, none of these items i
are considered open items.
- 1. 6 Confirmatory Items The staff identified confirmatory items in the SER that required additional information to confirm preliminary conclusions. The status of these items is I
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Ifsted in Table 1.2 (an updated version of SER Table 1.4). As Table 1.2 indi-
, cates, none of these items are considerect emfirmatory items.
- 1. 7 License Condition Items In Sect' ion 1.7 of the SER, the staff identified seven license conditions. These included several issues to be resolved by the licensee as a condition for issu-ance of an operating license, and other issues to be resolved in the longer term to ensure that NRC. requirements are met during plant operation.
The license conditions are listed in Table 1.3 (an updated version of SER Table 1.5). If the staff has removed the license condition, the notation
" closed" so indicates. If the license condition has been revised, that is noted, too, in Table 1.3. The section of this supplement in which the change j is reported is listed in Table 1.3.
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Table 1.1 Listing of outstanding items (revised from SER Table 1.3)
Item Status Section*
(1) Internally generated missiles Closed (SSER 1)
(2) Diesel generators Closed (SSER 4)
(3) Protection against postulated pipe Changed to License breaks outside containment Condition 12 (SSER 4)
(4) Loading combinations Closed (SSER 1)
(5) Design and construction of Closed (SSER 1) component supports (6) Inservice testing of pumps and Closed (SSER 4) valves (7) Equipment qualification Changed to License Condition 13 (SSER 4)
(8) Flow measurement capability Closed (SSER 4)
(9) Loose parts detection program Closed (SSER 3),
(10) Subcompartment analysis Closed (SSER 4)
(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) block valve, compliance with NUREG-0737 (TMI Action Plan)
(14) Fire protection Changed to License Condition 14 (SSER 4)
(15) Functional capability of ac and dc Closed (SSER 3) emergency lighting (16) Shift technical advisor training Changed to License program and operating experience Condition 15 for startup (SSER 4)
(17) Emergency Plan Closed (SSER 4)
Millstone 3 SSER 5 1-3
Table 1.1 (Continued)
Item Status Section*
(18) Limitation on overtime Closed (SSER 2)
(19) Q list Closed (SSER 3)
(20) Detailed Control Room Design Review Closed (SSER 4)
, Section of this supplement where item is discussed.
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l Table 1.2 Listing of confirmatory items (revised from SER Table 1.4) l Item Status Section*
(1) Plant's seismic capability beyond Closed (SSER 3) design basis (2) Dynamic loading Closed (SSER 3)
(3) Liquefaction potential Closed (SSER 3)
(4) Shoreline slope Closed (SSER 3)
(5) Turbine maintenance program Closed (SSER 3)
(6) Barrier design procedures Closed (SSER 1)
(7) Inservice examination of all pipe Deleted (SSER 4) welds in break exclusion area (8) Jet impingement effects Deleted (SSER 4)
(9) Ultimate capacity of containment Closed (SSER 1)
(10) Design of spent fuel racks Closed (SSER 3)
(11) Program evaluation related to TMI Closed (SSER 4)
Action Plan Item II.D.1 (12) Predicted cladding collapse time Deleted (SSER 1, Appendix H)
(13) Fuel assembly mechanical response Closed (SSER 3)
(14) Margins itemized in WCAP-8691 Closed (SSER 3)
(15) Thermal-hydraulic analyses to Changed to License support N-1 loop operation Condition 11 (SSER 4)
(16) Control rod drive structural Closed (SSER 3) materials (17) ASME Code cases for Section III, Closed (SSER 2)
Class I, components (18) Yield strength of austenitic Closed (SSER 3) stainless steels in reactor coolant pressure boundary (19) Onsite demonstration of ultrasonic Closed (SSER 3) inspection Millstone 3 SSER 5 1-5
Table 1.2 (Continued)
Item Status Section*
(20) Preservice inspection program review Closed (SSER 4) and relief requests (21) Preservice and inservice inspection Closed (SSER 3) of steam generators (22) Containment liner weld channel Closed (SSER 2) venting (23) Maximum external differential Closed (SSER 4) pressure on containment (24) Minimum containment pressure for Closed (SSER 2) emergency core cooling system (25) Procedures for actuating hydrogen Closed (SSER 3) recombiner (26) Secondary enclosure building Closed (SSER 4)
(27) Sump flow approach velocity Closed (SSER 4)
(28) Compliance with GDC 51 Closed (SSER 3)
(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 trip attachment (31) Reactor coolant pump underspeed Closed (SSER 4) trip (32) Conformance with BTP ICSB-26 Closed (SSER 1)
(33) Test of engineering safeguard P-4 Closed (SSER 1) interlock (34) Steam generator level control and Closed (SSER 2) protection (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 Millstone 3 SSER 5 1-6
Table 1.2 (Continued) l Item Status Section*
(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 1E control circuits (40) Sequencer deficiency report Closed (SSER 2)
(41) Balance-of plant instrumentation Closed (SSER 2) and control system testing capability (42) Instrument accuracy related to Closed (SSER 2)
Positions [ Attachments] 4, 5, and 6, TMI Action Plan Item II.F.1 (43) Description and analysis Closed (SSER 1) ((
demonstrating compliance with GDC 5 (44) Physical separation of offsite Closed (SSER 3).
circuits within a common right of way (45) Physical separation of offsite Closed (SSER 3) circuits between switchyard and Class IE system (46) Generation rejection scheme Closed (SSER 3)
(47) Description and analysis Closed (SSER 1) demonstrating compliance with GDC 17 (48) Description and analysis Closed (SSER 1) demonstrating compliance with GDC 18 (49) Positive statement of compliance Closed (SSER 1) with BTP PSB-1 (50) Compliance with Position 1 of Closed (SSER 3)
BTP PSB-1 (51) Adequacy of station electric Closed (SSER 4) distribution system voltage (52) Routing of power cables in the Deleted (SSER 1, cable spreading area Appendix H)
Millstone 3 SSER 5 1-7
Table 1.2 (Continued)
Item Status Section*
(53) Battery charger and transformer Closed (SSER 3) used as isolation devices (54) Design criteria of associated Deleted (SSER 1, circuits from isolation device to Appendix H) load (55) Core damage procedure (THI Action Closed (SSER 1)
Plan Item II.B.3, Criterion 2)
I (56) Control of concrete dust Closed (SSER 3)
(57) Qualification of engine-mounted Closed (SSER 3) control panels (58) 7-day fuel oil storage of each Closed (SSER 4) diesel generator (59) Airborne radioactivity monitoring Closed (SSER 3)
(60) Process control program for Closed (SSER 3) solidification of wet wastes (61) THI Action Plan Item II.F.1.1 Closed (SSER 3)
(62) TMI Action Plan Item I.C.1 Changed to License procedures generation package Condition 10 nuclear steam supply system (SSER 4)
(63) Physical Security Plan Closed (SSER 3)
(64) Initial test program Closed (SSER 3)
(65) Reactor coolant pump trip during Closed (SSER 4) loss-of-coolant accident (66) TMI Action Plan Item III.D.1.1 Closed (SSER 4) 1 (67) Analysis of droppyd control rod Closed (SSER 3)
(68) Steam generator tube rupture Deleted (SSER 4)
(69) No failure in emergency core Deleted (SSER 2) cooling system (ECCS) is not most limiting case in evaluating EECS i
(70) QA program commitments Closed (SSER 3)
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i Table 1.2 (Continued)
Item Status Section*
(71) Mass and energy release analysis Changed to License Condition 9 (SSER 4)
- Section of this supplement where item is discussed.
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Table 1.3 Listing of license conditions (revised from SER Table 1.5)
Item Status Section*
(1) Instrumentation for monitoring post- Revised (SSER 4) accident conditions, RG 1.97, Rev. 2 requirements (2) Compliance with NUREG-0612 (" Heavy Closed (SSER 2)
Load Handlirg")
(3) Installation of postaccident sampling Closed **
system (4) Sediment control during fuel oil Closed (SSER 3) storage tank refill (5) Moisture in air start system Revised (SSER 3)
(6) Preheating of rocker arm lubrication Closed (SSER 3) oil system (7) 81c:kage of access hatch in diesel Closed (SSER 3) generator exhaust system (8) Plant-specific analyses utilizing Added (SSER 2)
NOTRUMP (TMI Item II.K.3.31)
(9) Mass and energy release analysis Closed (SSER 5) 6.2.1.4 (10) TMI Action Plan Item I.C.1 proce- Closed (SSER 5) 13.5.2 dures generation package nuclear steam supply system l (11) N-1 loop operation Unchanged (SSER 4)
(12) Protection against postulated pipe Closed (SSER 5) 3.11 breaks outside containment (13) Equipment qualification Closed (SSER 5) 3.11 (14) Fire protection Cloced (SSER 5) 9.5.1.5 (15) Shift technical advisor training Revised (SSER 5) 13.1.2 program and operating experience for startup (16) Seismic interaction program Closed (SSER 5) 3.9.2 (17) Hazards program Closed (SSER 5) 17
- Section of this supplement where item is discussed.
- Letter from J. F. Opeka (NNECO) to V. Noonan (NRC), dated January 24, 1986.
Millstone 3 SSER 5 1-10
l 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 1
3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping 3.6.2 Determination of Rupture Locations and Dynamic Effects Associated With the Postulated Rupture of Piping In the " Background" to Branch Technical Position (BTP) MEB 3-1 as presented in Standard Review Plan (SRP) Section 3.6.2 (NUREG-0800), the staff position on pipe break postulation acknowledged that pipe rupture is a rare event that may only occur under unanticipated conditions such as those that might be caused by pos-sible design, construction, or operation errors, unanticipated loads, or un- t anticipated corrosive environments. The BTP MEB 3-1 pipe break criteria were '
intended to utilize a technically practical approach to ensure that an adequate level of protection had been provided to satisfy the requirements of General Design Criterion (GDC) 4 (10 CFR 50, Appendix A). Specific guidelines were developed in BTP MEB 3-1 to define explicitly how the requirements of GDC 4 were to be implemented. The SRP guidelines in BTP MEB 3-1 were not intended to be absolute requirements, but represent viable approaches considered to be acceptable by the staff.
The SRP provides a well-defined basis for performing safety reviews of light-water reactors. The uniform implementation of design guidelines in BTP MEB 3-1 ensures that a consistent level of safety will be maintained during the licens-ing process. Alternative criteria and deviations from the SRP are acceptable provided an equivalent level of safety can be demonstrated. Acceptable reasons for deviations from SRP guidelines include changes in emphasis of specific guidelines as a result of new developments from operating experience or plant-unique design features not considered when the SRP guidelines were developed.
The SRP presents the most definitive basis available for specifying NRC's de-sign criteria and design guidelines for an acceptable level of safety for light-water-reactor facility reviews. The SRP guidelines resulted from many years of experience gained by the staff in establishing and using regulatory requirements in the safety evaluation of nuclear facilities. The SRP is part of a continuing regulatory standards development activity that not only documents current meth-ods of review, but also provides a basis for an orderly modification of the review process when the need arises to clarify the content, correct any errors, or modify the guidelines as a result of technical advancements or an accumula-tion of operating experience. Proposals to modify the guidelines in the SRP are considered for their impact on matters of major safety significance.
The staff has recently received a request from the licensee to consider an alternate approach to the guidelines in SRP Section 3.6.2, BTP MEB 3-1, regard-ing the postulation of intermediate pipe breaks (letter dated January 14, 1986).
For all high-energy piping systems identified in the January 14 letter, the licensee proposes to eliminate from design considerations those breaks generally referred to as " arbitrary intermediate breaks" (AIBs) which are defined as those break locations which, based on piping stress analysis results, are below the Millstone 3 SSER 5 3-1
stress and fatigue limits specified in BTP MEB 3-1, but are selected to provide a minimum of two postulated breaks between the terminal ends of a piping system.
The licensee has stated that occupational radiation exposure during inspection, maintenance, and repair will be reduced over the life of the plant. The licensee is re' questing approval of alternative pipe break criteria to eliminate the re-quirement to evaluate the jet impingement effects from six AIB locations and to eliminate the need to design and protect from the ef fects of AIBs during future plant modifications. However, the licensee has stated that the elimination of AIBs will not impact the environmental qualification of safety-related equipment.
The break postulation for environmental effects is performed independently of break postulation for pipe whip and jet impingement.
In the early 1970s when the pipe break criteria in BTP MEB 3-1 were first drafted, the advantages of maintaining low stress and usage factor limits were clearly recognized, but it was also believed that equipment in close proximity to the piping throughout its run might not be adequately designed for the environmental consequences of a postulated pipe break if the break postulation proceeded on a purely mechanistic basis using only high stress and terminal end breaks. As ce {
pipe break criteria were implemented by the industry, the impact of the pipe break criteria became apparent on plant reliability and costs as well as on plant safety. Although the overall criteria in BTP MEB 3-1 have resulted in a viable methed which assures that adequate protection has been provided to satisfy the requirements of GOC 4, it has become apparent that the particular criterion requiring the postulation of arbitrary intermediate pipe breaks can be overly restrictive and may result in an excessive number of pipe rupture protection devices which do not provide a compensating level of safety.
At the time the BTP MEB 3-1 criteria were first drafted, high-energy leakage cracks were not being postulated. In Revision 1 to the SRP (NUREG-0800), the concept of using high-energy-leakage cracks to mechanistically achieve the en-vironment desired for equipment qualification was introduced to cover areas which are below the high-stress / fatigue-limit break criteria and which would otherwise not be enveloped by a postulated break in a high energy line. In the proposed elimination of AIBs, the staff believes that the essential design re-quirement of equipment qualification is not only being retained but is being improved, since all safety related equipment is to be qualified environmentally and, furthermore, certain elements of construction which may lead to reduced reliability are being eliminated.
In addition, some requirements which have developed over the years as part of the licensing process have resulted in additional safety margins that overlap the safety margin provided in the pipe break criteria. For example, the cri-teria in BTP MEB 3-1 include margins to account for the possibility of flaws that might remain undetected in construction and to account for unanticipated piping steady-state vibratory loadings not readily determined in the design process. However, inservice inspection requirements for the life of the plant to detect flaws before they become critical, and staff positions on the vibra-tion monitoring of safety-related and high-energy piping systems during pre-operational testing, further reduce the potential for pipe failures occurring from these causes.
Because of the recent interest expressed by the industry to eliminate the AIB criteria and, particularly, in response to the submittals provided by several utilities including Northeast Nuclear Energy Company (NNECO), the staff has Millstone 3 SSER 5 3-2
reviewed the BTP MEB 3-1 pipe break criteria to determine where such changes may be made.
3.6.2.1 Bases for the Elimination of Arbitrary Intermediate Pipe Breaks In a letter dated January 14, 1986, the licensee presented its request for the elimination of AI8s ard the technical bases for its proposal. There is a general consensus in the nuclear industry that current knowledge and experience support the conclusion that designing for the arbitrary intermediate pipe breaks is not justif f ed. The reasons for this conclusion are discussed in the paragraphs that follaw.
(1) 02 erating Experience Does Not Support Need for Criteria The combined operating history of commercial nuclear plants (extensive operating experience in more than 80 operating U.S. plants and a number of similar plants overseas) has not shown the need to provide protection from the dynamic effects i of AI8s.
(2) Piping Stresses Well Below ASME Code A110wables Currently, AIBs are postulated to provide a minimum of two pipe breaks at the two hi Jhest stress locations between piping terminal ends. Consequently, AIBs are postulated at locations in the piping system where pipe stresses and/or cumulative usage factors are well below ASME Code allowables. Such postulation necessitates tne installation and maintenance of complicated mitigating devices to af ford protection f rom dynamic ef fects such as pipe whip and/or jet impinge-ment.
When these selected break locations have stress levels only slightly greater than tne rest of the system, installation of mitigating devices lends little to enhance overall plant safety.
(3) Unanticipated Thermal Expansion Stress Unanticipated stresses from restraint of thermal expansion can be introduced into the piping system if pipe rupture protection devices come into contact with the pipes. The potential for this happening is greater than that for mechanistic failure at an arbitrary break point. To prevent a consequent de-crease in the overall reliability of the pipe system, an additional as-built verification step is involved in the design process for each installed pipe whip restraint. Elimination of AI8s would significantly reduce the ef fort in-volved in designing and installing pipe rupture protection devices.
(4) Improved Inservice Inspection Pipe whip restraints are normally located adjacent to or surrounding the welds at changes in pipe direction. Access during plant operation of inservice inspec-tion activities can be improved by eliminating congestion created by these pipe rupture protection devices and the supporting structural framing associated with arbitrary pipe breaks.
(5) Substantial Lost Savings and Reduction in Radiation Exposure The estimated cost savings in design, material, and construction over the 40 year plant life is estimated to be well in excess of $1 million and the Mll' stone 3 SSER 5 3-3
savings in personnel radiation exposure in excess of 100 person-rem for Mill-stone 3. These figures are probably conservative in light of the uncertainty involved in predicting the amount of future plant modifications.
(6) Decrease in Heat Loss The elimination of pipe whip restraints associated with arbitrary breaks will
, preclude the requirement for cutback insulation or special insulating assemblies l near the closefitting restraints. This will reduce the heat loss to the sur-l rounding environment, especially inside containment, and will result in improved operational efficiency.
3.6.2.2 Staff Evaluation of the Bases for the Elimination of Arbitrary Breaks The technical bases for the elimination of the AIB criteria as discussed in the preceding section of this report provided many arguments supporting the licensee's conclusion that the current SRP guidelines on this subject should be changed. [
However, it is not apparent that a unilateral position by the utility (conclud- t ing an unconditional deletion of the AIB criteria) can be justified without a clear understanding of the safety implications that may result for the various classes of high-energy piping systems involved. In this section, the staff discusses the bases behind the current AIB criteria from an ASME Code design standpoint and put into perspective the uncertainty factors on which the need to postulate AIBs should be evaluated.
The ASME Code design requirements for Class 1 piping systems differ from those for Class 2 and 3 piping systems; there are, however, other design considerations that are common to Class 1, 2, and 3 systems. These other design considerations
[viz. , (1) intergranular stress corrosion cracking, (2) water / steam hammer, and (3) thermal fatigue] can affect the safety of the systems in which AIDS are eliminated. Therefore, while evaluating the acceptability of the licensee's proposed deviation from SRP Section 3.6.2, the staff examined the significance of the above three additional design considerations for the specific Millstone 3 piping systems proposed by the licensee for elimination of AIBs.
(1) ASME Code Class 1 Piping Systems In accordance with BTP MEB 3-1 [ paragraph B.1.c.(1)] breaks in ASME Code Clar '
piping should be postulated at the following locations in each piping Jrd br:
run:
(a) at terminal ends; (b) at intermediate locations where the maximum stress range as calculated by Eq. (10) and either Eq. (12) or (13) of ASME Code NB-3650 exceeds 2.4 Sm I (c) at intermediate locations where the cumulative usage factor exceeds 0.1; (d) if two intermediate locations cannot be determined by (b) and (c) above, two highest stress locations based on Eq. (10) should be selected.
The AIB criteria are stated in (d) abnve. It should be noted that the request for alternative criteria does not propose to deviate from the criteria in (a),
(b), and (c) above. Pipe breaks will continue to be postulated at terminal ends irrespective of the piping stresses.
Millstone 3 SSER S 3-4
Pipe breaks are to be postulated at intermediate locations where the maximum stress range as calculated by Eq. (10) and either (12) or (13) exceeds 2.4 The stress evaluation in Eq. (10) represents a check of the primary plus S*.
secondary stress intensity range from ranges of pressure, moments, thermal gradients, and combinations thereof. Equation (12) is intended to prevent for-mation of plastic hinges in the piping system caused only by moments resulting from thermal expansion and thermal anchor movements. Equation (13) represents a limitation for primary plus secondary membrane plus bending stress intensity excluding thermal bending and thermal expansion stresses; this limitation is intended to ensure that the K e -factor (strain concentration factor) is conserva-tive. The K -factor was developed to compensate for absence clastic shakedown when primary *plus secondary stresses exceed S,.
With respect to piping stresses, the pipe break criteria were not intended to imply that breaks will occur when the piping stress exceeded 2.4 5, (80% of the primary plus secondary stress limit). The staff believes, however, that if a pipe break were to occur (on one of those rare occasions), it is more likely to occur at a piping location where there is the least margin to the ultimate tensile strength.
Similarly, from a fatigue strength standpoint, the staf f believes that a pipe break is more likely to occur where the piping is expected to experience large cyclic loadings. Although the staff concurs with the industry belief that a cumulative usage factor of 0.1 is a relatively low limit, the uncertainties involved in the design considerations with respect to the actual cyclic load-ings experienced by the piping tend to be greater than the uncertainties involved in the design considerations used for the evaluation of primary and secondary stresses in piping systems. The staff finds that the conservative fatigue considerations in the current SRP guidelines provide an appropriate margin of safety against uncertainties for those locations where fatigue fail-ures are likely to occur (e.g., at local welded attachments).
(2) ASME Code Class 2 and 3 Piping Systems In accordance with BTP MEB 3-1 [ paragraph B.1.c(2)], breaks in ASME Code Class 2 and 3 piping should be postulated at the following locations:
(a) at terminal ends (b) at intermediate locations selected by one of the following criteria:
(i) at each pipe fitting, welded attachment, and valve (ii) at each location where the stresses exceed 0.8 (1.2 S h
- SA ), but at not less than two separated locations chosen on the basis of highest stress.
In its proposal the licensee has not proposed changing criterion (a) above.
Postulation of pipe breaks at terminal ends will not be eliminated in the pro-posed SRP deviation for Class 2 and 3 piping systems.
The AIB criterion is stated in criterion (b)(fi) above where breaks are to be postulated at intermediate locations where the stresses exceed 0.8 (1.2 Sh+S) A Millstone 3 SSER 5 3-5
but "at not less than two separated locations chosen on the basis of highest stress." The stress limit provided in the above pipe break criterion represents the stress associated with 80% of the combined primary and secondary stress limit. Thus, a break is required to be postulated where the maximum stress range as calculated by the sum of Eq. (9) and (10) of NC/ND-3652 of the ASME Code,Section III, exceeds 80% of the combined primary and secondary stress limit, when one considers those loads and conditions for which level A and 3 level 8 stress levels have been specified in the system's design soecification (i.e., sustained loads, occasional loads, and thermal expansion), including an operating basis earthquake (OBE) event. However, the Class 2 and 3 pipe break criteria do not have a provision for the postulation of pipe breaks based on a fatigue limit since an explicit fatigue evaluation is not required in the ASME Code for these classes of construction because of favorable service experience and lower levels of operating cyclic stresses.
For those Class 2 and 3 piping systems which experience a large number of stress cycles (e.g., main steam and feedwater systems), the ASME Code has provisions which are intended to address these types of loads. The rules governing con-siderations for welded attachments in ASME Class 2 and 3 piping which do pre-ciude fatigue failure are partially given in paragraph NC/ND-3645 of the ASME Code. The Code states:
External and internal attachments to piping shall be designed so as not to cause flattening of the pipe, excessive localized bending stresses, or harmful thermal gradients in the pipe wall. It is im-portant that such attachments be designed to minimize stress concen-trations in applications where the number of stress cycles, due either to pressure or thermal effect, is relatively large for the expected life of the equipment.
Code rules governing the fatigue effects associated with general bending stresses caused by thermal expansion are addressed in NC/ND-3611.2(e) and are generally incorporated into the piping stress analyses in the form of an allowable stress reduction factor.
Thus it can be concluded that when the piping designers have appropriately con-sidered the fatigue effects for Class 2 and 3 piping systems in accordance with NC/ND-3645, the likelihood of a fatigue failure in Class 2 and 3 piping caused by unanticipated cyclic loadings can be significantly reduced.
(3) Additional Design Considerations In its presentation to the Advisory Committee on Reactor Safeguards (ACRS) on June 9,1983, and in an October 5,1983, meeting between a group of pressurized- i water reactor (PWR) near-term operating license utilities and the NRC staff, the staf f indicated that the elimination of AIBs was not to apply to piping systems in which stress corrosion cracking, large unanticipated dynamic loads such as steam or water hammer, or thermal fatigue in fluid mixing situations could be expected to occur. In addition, the elimination of AIBs was to have no effect on the requirement to environmentally qualify safety-related equipment and in fact this requirement was to be clarified to assure positive qualification requirements.
Millstone 3 SSER 5 3-6
(a) Intercranular Stress Corrosion Cracking (IGSCC)
As discussed in the licensee's letter of January 14, 1986, the IGSCC potential is likely to be reduced if the following factors are controlled: high residual tensile stresses, susceptible piping material and a corrosive environment. Al-though any stainless or carbon steel piping will exhibit some degree of residual stress and material susceptibility, the licensee has demonstrated (January 14, 1986, letter) that Millstone 3 minimizes the potential for IGSCC by utilizing piping material with low susceptibility to stress corrosion and by preventing the existence of a corrosive environment. The likelihood of stress corrosion cracking in stainless steel increases with carbon content. Consequently, only the lower carbon content stainless steel (304, 304L, 316, 316L) have been used for the primary systems at Millstone 3. For the secondary piping systems, carbon steel has been used for the piping, fittings, and valve bodies forming the pressure boundaries. The piping is flushed with demineralized water sub-ject to limits on total dissolved solids, conductivity, chlorides, fluorides, and pH. Water chemistry for preoperational testing is controlled by written specifications.
During plant operation, primary- and secondary-side water chemistry are moni-tored. Containment concentrations are maintained below the thresholds known to be conducive to stress corrosion cracking. The water chemistry control standards are included in operating procedures for the systems where arbitrary breaks are being eliminated.
On the basis of the above information, the staff has concluded that adequate protection against IGSCC has been provided by the licensee.
(b) Water / Steam Hammer Because of the susceptibility of main feedwater (MFW) systems to water hammer, the licensee has incorporated several water hammer prevention / minimization features into the design of the MFW piping at Millstone 3. As discussed in the January 14, 1986, letter and ir SER Section 10.4.7, the potential for water hammer caused by rapid condensation of a steam bubble in the steam generator (SG) feedring has been reduced by installing J-tubes in the feedring to prevent drainage of water during low steam generator water level. Also, the feedwater (FW) connections to the steam generators utilize a downward-turned 90-degree elbow that does not present a horizontal pipe run immediately upstream of the FW nozzles. This configuration is intended to prevent formation of steam pockets during steam generator low water level conditions and to minimize the volume of water external to the SG which could pocket a steam bubble. This piping arrangement follows Westinghouse design guidelines.
On the basis of the above information and the inclusions in SER Section 10.4.7, the staff concludes that the design features and operating procedures described above will minimize the potential for water hammer occurrence in the main FW piping system.
(c) Thermal Fatigue The staff has concluded that the Millstone 3 systems for which A!85 are to be eliminated are not susceptible to thermal fatigue and mixing for the following reasons:
Millstone 3 SSER 5 3-7
(i) The fatigee analysis performed by the licensee for Class 1 piping systems shows that all of the Class 1 AIB locations involve cumtlative usage factors below the AIB pos'ulation limit of 0.1 (licensee's letter of Janu-ary 14, 1986). For Class i and 3 piping components, fatigue failure pro-tection is assured by the ASME Code-allowable stress range checks and a stress range reduction factor for thermal expansion stress. The mandatory breaks are postulated at 80% of the Code-allowable stresses, even after eliminating the AIBs identified in the January 14, 1986, letter).
(ii) The plant systems are designed to minimize thermal cycling and thermal shock. The FW system is arranged so that during normal operation, mixing (backflow from the steam generator) of hot water in the steam generator with the lower temperature FW is localized in the area of the FW inlet nozzle. The FW piping geometry has a downward elbow directly off the steam generator nozzle which forms a loop seal geometry to minimize mix-ing. The inverted J-tubes located on the FW sparger further reduce the mixing effect. Feedwater flow surges are precluded by control valve de-sign. The auxiliary FW ties into the vertical main FW piping just up-stream of the loop seal geometry. The location of the auxiliary FW con-f nection minimizes the length of piping subjected to the lower temperature suxillary FW. Thermal sleeves are provided in the branch connection to the main FW piping and in the steam generator nozzle to minimize the thermal transient effects of auxiliary FW initiation.
(4) Class 1 Pipina Systems Evaluation For Class 1 piping, a considerable amount of quality assurance in design, anal-yses, fabrication, installation, examination, testing, and documentation is provided which ensures that the safety concerns associated with the uncertain-ties discussed above are significantly reduced. On the basis of the staff eval-uation cf the design considerations given to Class 1 piping, the stress and fatigue limits provided in the BTP MEB 3-1 break criteria, and the relatively small degree of uncertainty in unanticipated loadings, the staff finds that the need to postulate arbitrary intermediate pipe breaks in ASME Code Class 1 piping in which large unanticipated dynamic loads, stress corrosion cracking, and ther-mal fatigue such as in mixing situations are not present and in which all equip-ment has been environmentally qualified is not compensated for by an increased level of safety. In addition, systems may actually perform more reliably for the life of the plant if the SRP criterion to postulate arbitrary intermediate breaks for ASME Code Class 1 piping is eliminated. The staff has concluded that the above-described requirements are present for those ASME Code Class 1 piping systems identified in the licensee's submittal of January 14, 1986.
(5) Class 2 and 3 Pipina Systems Evaluation On the basis of the staff evaluation of the design considerations given to Class 2 and 3 piping, the stress limits provided in the SRP break criterion, and the relatively small degree of uncertainty in unanticipated loadings, the staff finds that dispensing with arbitrary intermediate pipe breaks is justified for Class 2 and 3 piping in which stress corrosion cracking, large unanticipated dynamic loads, or thermal fatigue in fluid mixing situations are not expected to occur provided (a) the piping designers have appropriately considered the effects of local welded attachments per NC/ND-3645 and (b) all safety related Millstone 3 SSER S 3-8
equipment in the vicinity of Class 2 and 3 piping systems has been environ-mentally qualified for the nondynamic effects of a nonmechanistic pipe break with the greatest consequences on the equipment. The staff has concluded that the above-described requirements are present for those ASME Code Class 2 and 3 piping systems identified in the licensee's letter dated January 14, 1986.
(6) Piping Systems Not Included in Proposal For those piping systems, or portions thereof, which are not included in the licensee's submittal of January 14, 1986, the staff requires that the existing guidelines in BTP ME8 3-1 of the SRP (NUREG-0800), Revision 1, be met. However, should other piping lines which are not specifically identified in the licensee's submittal subsequently qualify for the conditions described above, the imple-mentation of the proposed elimination of the AIB criteria may be used provided those additional piping lines are appropriately identified to the staff.
3.6.2.3 Conclusion The licensee has proposed a deviation from the current guidelines of the SRP by requesting relief from postulating arbitrary intermediate pipe breaks in high-energy piping systems which are not susceptible to intergranular stress corrosion cracking, steam or water hammer ef fects, and thermal fatigue in fluid mixing. The SRP guideline which requires that two intermediate breaks be postu-lated even when the piping stress is low resulted from the need to ensure that equipment qualified for the environmental consequences of a postulated pipe break was provided over a greater portion of the high-energy piping run.
This proposal is based, in part, on the condition that all equipment in the spaces traversed by the fluid system lines, for which AIBs are being eliminated, is qualified for the environmental (nondynamic) conditions that would result from a nonmechanistic break with the greatest consequences on surrounding equip-ment. In addition, the licensee has committed to perform preoperational test-ing of all the systems identified in the letter of January 14, 1986, and also monitor those systems for vibration during preoperational and startup testing.
The staf f has evaluated the technical bases for the proposed deviation with respect to satisfying the requirements of GDC 4. Furthermore, the staff has considered the potential problems identified in NUREG/CR-2136 which could impact overall plant reliability when excessive pipe whip restraints are installed.
On the basis of its review, the staff finds that when those piping system condi-tions as stated above are met, there is a sufficient basis for concluding that an adequate level of safety exists to accept the proposed deviation.
Thus, on the basis that the piping systems have satisfied the above conditions, the staf f concludes that the pipe rupture postulation and the associated effects are adequately considered in the design of Millstone 3 and, therefore, the devi-ation from the Standard Review Plan is acceptable.
- 3. 9 Mechanical Systems and Components 3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment In Section 3.9.2 of SSER 4, the staff identified a confirmatory item regarding the implementation of the seismic interaction program. The licensee had Millstone 3 SSER 5 3-9
previously submitted for staff review its seismic interaction program for addressing potential interactions between seismic Category I and non-seismic Category I piping and equipment. The staff evaluation provided in SSER 4 found the program to be acceptable. However, the implementation of the program had not been completed at that time. Thus, the staff considered the issue to be acceptable contingent upon a satisfactory implementation of the progra'n, the acceptability of the results, and the completion of any corrective actions which may be required as a results of its implementation.
In a letter from J. F. Opeka to H. Denton dated January 14, 1986, the licensee provided the results of its seismic interaction program. The results of the program found that the aeismic interactions which could occur between seismic Category I and non-seismic Cateogry I piping and equipment will not adversely affect the ability of seismic Category I piping and equipment to perform their safety functions. The calculated stresses induced in the seismic Category I piping and equipment resulting from the seismic interactions were all found to be within the acceptance criteria previously accepted by the staff in SSER 4, except as noted below.
As a confirmatory measure, the staff performed an audit of the seismic inter-action program results at the Sargent & Lundy offices on January 9, 1986. The audit focused on the basis for concluding that the piping analyses performed are bounding for the evaluation of non seismic Category I piping systems. Fur-thermore, the staff reviewed the program results to evaluate the appropriate-ness of overall conclusions drawn from those piping subsystems where high stresses and large lateral deflections were found, in order to determine the need for generic recommendations or additional corrective actions.
During the audit, the staff reviewed the resolution of high piping stresses and large lateral deflections as noted in several piping subsystems.. The staff found that the use of damping values per ASME Code Case N-411 in conjunc-tion with refined analytical techniques resulted in acceptable piping stresses.
Except for two cases, the piping stresses were found to meet the ASME Code Ser-vice Level 0 allowable values as stated in the 1983 Edition (including Winter 1984 Addenda).
The two cases in which Service Level 0 allowables were exceeded occurred in a chilled water subsystem (AX-107X) and in a boron recovery subsystem (SL-7A).
Further evaluation by the licensee of the AX-107X overstress condition deter-mined that the formation of a plastic hinge would occur in a branch line pipe elbow. However, the maximum angular rotation at the hinge was calculated to be 2.2 degrees and would not lead to total collapse of the piping subsystem.
Furthermore, the high stress was primarily caused by a main header displacement and not by the branch pipe inertia acceleration. If this secondary displace-ment component of the pipe stress were removed from the stress evaluation, the calculated primary stress would meet ASME Code-allowable values.
The second condition of overstress occurred in subsystem SL-7A which consisted of mainly 1-inch- and 2-inch-diameter piping. Two overstress conditions were found at socket welded tees. However, when the actual size of the fillet welds was used in the calculations, the recalculated stresses were found to be within Level 0-allowable values. The staff finds that the additional evaluations per-formed by the ifcensee adequately address the overstress conditions and are acceptable.
Millstone 3 SSER 5 3-10
1 1
Excessive piping lateral deflections were identified as a generic concern by 2
the licensee because of the potential for the piping to slip off certain types i of supports. Thus, the licensee initiated a walkdown of all affected piping i systems and implemented modifications to those supports to preclude the poten-tial of pipe slipoff. All modifications to these supports will be completed before 5% power is exceeded. The staff finds the actions taken by the licensee I
to address large lateral pipe deflections to be acceptable. !
The staff also reviewed the methodology used to evaluate the effect of upward
! loads on supports designed for downward loads only and found the approach i adequate. The staff also noted that the seismic accelerations in the vertical 4 direction, in general, tend to be less than the dead weight of the piping.
For the evaluation of pipe supports, the staff reviewed the qualification of l certain standard component supports in which the vendor-recommended allowable values were exceeded. Fcr U-bolts and U-straps, the licensee performed anal- !
- yses to qualify the U-bolts and U-straps to higher loads. The analysis results ,
i were also compared with recent test results available from the vendor for '
! further substantiation and were found to be in agreement. For concrete expan- l
} sion anchor bolts, the licensee had previously conducted extensive testing to ,
j qualify the anchor bolts for the specific bolt sizes and concrete strengths used at Millstone 3. For the non-seismic Category I piping systems, the li-censee used faulted (seismic Level D) allowables for the anchor bolts, which ;
resulted in a minimum factor of safety of 2.95. Although the safety factor k is less than the vendor recommended design factor of safety of 4.0, the staff found the lesser factor of safety which could result in a slight slippage of ,
the anchor bolt (but not complete pullout failure) to be acceptable for the .
faulted condition of non-seismic Category I piping systems where structural integrity, not piping functionality, is the primary consideration, j On the basis of the staff's review of the results of the seismic interaction i
program, the staff concludes that the bounding analyses of the non-seismic '
! Category I piping systems and the supplemental walkdown measures, including I
corrective actions taken by the licensee to resolve potentially unacceptable r seismic /non-seismic interactions, provide a reasonable basis to conclude that 3 the non-seismic Category I piping systems will maintain their structural integ- '
rity and will not adversely affect the ability of sel,mic Category I piping and equipment to perform their safety functions. Thus, the staff finds the imple-mentation of the seismic interaction program to be adequate and the results to I
! be acceptable. All modifications to preclude potentially unacceptable swing /
j sway interactions will be completed before 5% power is exceeded. The piping stress reports have yet to be completed to include reconciliation of these sup-port modifications. However, the staff believes the impact of these modifica-i tions will not likely change the acceptability of the program results nor the I conclusions reached by the staff. The licensee should inform the staff when
! the final reconciliation is completed. If further modificaticas are required, I the staff will pursue this matter at that time. Thus, License Condition 2.C(4) j of the Millstone 3 Operating License, NPF-44, has been adequately satisfied. ,
j 3.11 Environmental Qualification of Electrical Equipment Important to Safety j and safety-Related Mechanical Equipment t
! License Condition 2.C(3) of the Millstone 3 Low Power License stated "By Decem-I ber 27, 1985, the licensee shall submit a revised compartment analysis using the [
I i j !
! Millstone 3 SSER 5 3-11
mass and energy release data provided by the Westinghouse Owners Group Program."
By letters dated December 20, 1985, and January 7 and 14,1986, the licensee provided the requested information.
Tc calculate pressure and temperature transients following a main steamline break (MSLB), the licensee used mass and energy data from Westinghouse Topical Report WCAP-10961-P to account for the ef fect of superheated steam release due to steam generator tube uncovery. By a letter from J. F. Opeka (Northeast Utilities) to V. S. Noonan (NRC), dated Jant.ary 17, 1986, the licensee provided for staff review one copy of WCAP-10961-P ent tied " Steam Line Break Mass / Energy '
Release for Equipment Environmental Qualificacion Outside Containment." In the interim, Westinghouse is preparing a formal submittal of the topical report in
- accordance with the requirements of 10 CFR 2.790.
1 The staff finds that Millstone 3 Model F steam generators and its main steamline i break protection system are included in the studies of the Westinghouse Topical i Report. The report includes six categories of Westinghouse Owners Group plants '
in which Hillstone 3 is categorized as Category 1. In each category, a spectrum of breaks in different sizes, power levels, break locations, and auxiliary feed-water models was studied in detail. The report is still under review; however, 1 the review has progressed sufficiently for the staff to conclude there is rea-
- sonable assurance that the concerns regarding Millstone 3 superheated steam blowdown have been resolved pending the generic approval of the topical report.
I In addition, the licensee stated that all Millstone 3 equipment which is required ,
[ to function to mitigate the consequences of a main steamline break accident is i qualified to function at the maximum compartment temperature of 325"F at steam-j line isolation. The licensee also stated that the equipment will remain in its 4 safe position regardless of the fact that it will be exposed to temperatures i above the qualification temperature. The staff reviewed all the information l provided by the licensee and found it acceptable. Accordingly, the staff con-j cludes that the aforementioned license condition has been satisfied. ,
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M111 stone 3 SSER S 3-12 1
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4 REACTOR 4.4 Thermal-Hydraulic Design l
- 4.4.4 Operating Abnormalities I
) 4.4.4.2 Crud Deposition and Flow Measurement Uncertainty I
In Section 4.4.4.2 of SSER 3 for Millstone 3, the staff reviewed the licensee's July 15, 1985, analysis for flow measurement uncertainty. It was concluded that l enough detail was not presented in the analysis to enable the staf f to find the ;
)' flow uncertainty values acceptable. Also, contrary to past experience, the flow measurement uncertainty for both four- and three-loop operation was presented as >
being the same.
A revised flow measurement analysis was presented (letter from J. F. Opeka ;
I (NNECO) to 8. J. Youngblood (NRC), dated September 26, 1985). The revised flow '
l measurement uncertainty analysis included both four-loop and three-loop opera- l l tion as well as elbow tap error. The total flow uncertainities with two indi- !
l cators per loop were given as: 2.4% (four loops in operation) and 2.76% (three
! loops in operation). These values did not include the 0.1% additional penalty to account for venturf fouling. The licensee committed to add inspection ports '
upstream and downst' ream of the venturis during the first refueling outage.
1 Before the start of each cycle (before performing the calorimetric measurement for flow), the venturf meters will be verified to be clear (by performing a visual inspection (borascope, photography, etc.)] and will be cleaned when necessary. However, if the venturi meters are not inspected, the licensee '
committed to add an additional 0.1% to the total RCS flow measurement
! uncertainty.
1, ,
i The staff reviewed the Millstone 3 plant-specific flow measurement uncertainty !
j analysis and compared the results with a generic Westinghouse flow measurement j analysis. The Hillstone 3 values were found to be conservative, but a number <
of details were still missing that were required for the analysis; however, '
l because of the conservative results presented in this analysis, the staff ac- '
i cepted the flow measurement uncertainities of 2.4% (four-loop operation) and ;
2.76% (three-loop operatien), subject to receiving a more detailed analysis for "
t confirmation. These values were specified in the Technical Specifications and
] were also used for the minimum allowed reactor coolant flow Technical j Specification, i The Technical Specifications also state the conditions for which the 0.1% ven-turf fouling penalty would be eliminated if the venturf meters were determined ,
to be cleaned. The licensee subsequently presented the detailed flow measure- '
ment uncertainty analysis for confirmation in a letter from J. F. Opeka (NNECO) i to V. S. Noonan (NRC), dated January 7,1986. This analysis provided sufficient detail. However, the flow measurement uncertainities were changed from the pre-vlous analysis in a less conservative direction. The new values are given as: i 4
2.0% (four-loop operation) and 2.3% (three-loop operation). !
I 1 l l
Millstone 3 SSER 5 41 l
i
The staff has not completed its review of the January 7, 1986, submittal.
Therefore, until this review is complete, the licensee will be required to use the flow uncertainty values and resulting Technical Specifications based on the September 26, 1985, submittal. These values are: 2.4% (for four-loop operation) ano 2.76% (for three-loop operation). These numbers do not include a possible venturf fouling uncertainty as discussed previously in this section.
4.4.8 Instrumentation for Detection of Inadequate Core Cooling 4.4.8.1 Clarification of Requirements A clarification of requirements for inadequate core cooling instrumentation (ICCI) was provided in Item II.F.2 of NUREG-0737, " Clarification of TMI Action Plan Requirements." In November 1982, the Conmission determined that an in-strumentation system for detecting inadequate core cooling (ICC) consisting of upgraded subcooling margin monitors (SMMs), core exit thermocouples (CETs),
and a reactor coolant inventory tracking system (ITS) is required for the operating pressurized water reactor facilities.
4.4.8.2 Inadequate Core Cooling Instrumentation System Design The Millstone 3 licensee, Northeast Nuclear Energy Co. (NNECO), has provided information in the revised FSAR Section 4.4.6.5 ( Amendtrent 9) and in a letter dated June 14, 1984, from W. G. Counsil (NNECO) to 8. J. Youngblood (NRC) in response to staf f concerns regarding ICCI.
The Millstone 3 ICC monitoring system consists of three instrumentation sub-systems: (1) the subcooling margin monitor (SMM), (2) the core exit thermo-couples (CETs), and (3) the reactor coolant inventory monitoring system. The Millstone 3 ICC system has been designed as Class IE with redundant trains, each containing standalone processing electronics and displays to monitor, alarm, and trend ICC. The monitoring system was tested and calibrated before fuel load. Millstone 3 is provided with two ICC information display systems.
Redundant Class IE cabinets are provided outside the control room and display SMM, CET, and coolant level information. In the control room, the primary ICC information is provided through the safety parameter display system (SPOS).
All the ICC information transmitted to the SPOS has been provided with optical isolation. The SPOS displays are designed to incorporate accepted human fac-tors principles so that the displayed ICC information can be readily under-stood by plar$t operators during normal and abnormal plant conditions.
4.4.8.3 Subcooling Margin Monitor The SMM system uses reactor coolant system (RCS) temperatures and pressures to calculate subcooling (to 300 F) and superheat (to 45"F) either in terms of tem-perature or pressure. The calculation is based on the most conservative values of the temperature and pressure input. The calculation of the subcooling/
superheat is performed by the SPOS using input from the T hot and Teold '*SI5t' ance temperature devices (RTDs), CET, the unheated junction temperatures of the heated junction thermocouple (HJTC), and the RCS pressure. Signal vall-dation techniques are utilized to ensure the quality of the input variables.
Saturation /superheat trouble alarms are provided on the main control board from the ICC cabinet.
Millstone 3 SSER 5 4-2
4.4.8.4 Core Exit Thermocouples The core exit thermocouple (CET) monitoring system consists of two redundant independent trains that monitor the 50 chromel-alumel CETs. All CETs are pro-vided with the required cold junction temperature compensation which consists of an RTO providing a signal to the ICC processor in the Class IE cabinet and O display. The CET temperature range is from 200*F to 2300 F. All CETs are dis-played on a digital panel meter selectable from a switch panel. All CETs are uniformly dispersed in the core, therefore, satisfying the requirement of a min-imum of four CETs per quadrant. A CET high alarm is provided in the main con-trol boards from the ICC cabinet.
4.4.8.5 Reactor Vessel Level Instrumentation System The heated junction thermocouple system monitors coolant inventory in the reg ion above the core. Redundant strings of heated junction thermocouples are arranged in the reactor vessel head area to provide an indication on conditions at eight distinct levels. The sys' tem includes two channels, each consisting of a string of eight equidistant sensors. The system indicates percent of level in the plenum and the head areas. One of the ways of displaying the ICC information is provided through the SPOS. However, the SPDS display is not Class 1E. Dur-ing the staff's audit of the SPDS on July 29, 1985, the licensee provided a
" Design Availability Calculation" for the SPOS estimating the availability at
- 99. 54T,. The plant-specific ICC procedures will be based on Combustion Engineer-ing reports CEN-185, CEN-152, and the CE0G 1etter to D. M. Crutchfield dated June 1, 1982, and will be incorporated into the Westinghouse emergency response guidelines.
The generic Combustion Engineering topical report on inadequate core cooling instrumentation using heated junction tnere.ocouples for reactor vessel level measurement has been reviewed by the staff dnd was found acceptable. The
! evaluation was published in NUREG/CR-2627, March 1982.
4.4.6 6 Evaluation and Conclusions The staff has reviewed the licensee's submittal dated June 14, 1984 [(W. G.
Counsil (NNECO) to 8. J. Youngblood (NRC)), the revised FSAR Section 4.4.6.5 (Amendment 9), and the licensee's submittal dated January 20, 1986 (J. F. Opeka (NNECO) to V 5. Noonan (NRC)].
The staff concludes that the ICC system design is acceptable for an operating license and to satisfy the requirements of NUREG-0737, Item II.F.2. The ICC '
primary and backup display instrumentation is acceptable with the SPDS being non-Class 1E but with an estimated availability of over 99%.
l Millstone 3 SSER 5 4-3
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1 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS i 5.2 Integrity of Reactor Coolant Pressure Boundary i
5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing i This evaluation supplements conclusions in this section of the SER (NUREG-1031) !
4 which addressed the definition of examination requirements and evaluation of i
- compliance with 10 CFR 50.55a(g). In previous supplements to NUREG-1031, Sec- {
j tion 5.2.4, the staff reported that the Preservice Inspection Program for the L j systems and components within the reactor coolant pressure boundary is consis- !
tent with the applicable regulation and Code requirements and the review is !
a confirmatory issue contingent upon the licensee completing the required ex- l aminations and identifying all impractical preservice inspection requirements [
with a supporting technical justification, j i In a letter dated December 23, 1985, the licensee submitted relief requests from .
! ASME Code Section XI requirements which the licensee has determined to be imprac-tical to perform at Hillstone 3 and provided supporting information pursuant to l l l
) 10 CFR 50.55a(a)(3). The staff evaluated the ASME Code-required examinations j
! that the Ifcensee determined to be impractical and the staff found that the r j licensee has demonstrated that compliance with the requirements would result in !
j hardships or unusual difficulties without a compensating increase in the level of quality and safety. On the basis of review of the licensee's submittals and granting of relief from the preservice examination requirements, the staff con-l cludes that the Preservice Inspection Program for systems and components within the reactor coolant pressure boundary at Millstone 3 is acceptable and in com- i pliance with 10 CFR 50.55a(g)(3). The detailed evaluation supporting this con- l 3
clusion is provided in Appendix N to this report. -
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I I i I i 6 ENGINEERED SAFETY FEATURES l
l 6.2 Containment Systems !
! 6.2.1 Containment Functional Design I i
j In the Safety Evaluation Report for Millstone 3, the staff evaluated the ade- l quacy of the containment structure functional design (SER Section 6.2.1.1).
In SSER 2, the staff evaluated the adequacy of the pressurizer cubicle design )
(SSER 2, Section 6.2.1.2). After SSER 2 was issued, the licensee submitted i several amendments to the FSAR that alter the analyses on which the staff's i above reviews are based. A brief description of the key changes made in each i amendment, and the impact of the changes on the licensee's licensing basis ;
calculations, is provided below. I 6.2.1.1 Containment Structure In FSAR Amendment 15, the licensee (1) revised the mass and energy release rate j
, data for the limiting pipe breaks for containment depressurization (generally !
j lower release rates for the double-ended pump suction guillotine (DEPSG), and j 4 the 0.6 DEPSG), (2) modified certain containment recirculation cooler parameters 1 (service heat water(UA) transfer flowfor to each each cooler cooler changed changed from from 65,000 gpmto 3.86 X 106 to3.79 6,230X gpm; 10 overall j 8tu/hr/*F), and (3) changed the initial containment pressure for depressuriza- :
tion analysis from 9.76 psia to 9.81 psia. In FSAR Amendment 17, the Ifcensee r (1) shifted the recirculation spray pump starting time from 220 seconds to 670 ;
seconds (2) took credit for approximately 260,000 square feet of additional ;
steel heat sink, and (3) made additional modifications to the mass and energy '
4 release rate data for the limiting pipe breaks for containment depressurization. !
/ The licensee reported that as a result of these changes, the peak containment l t pressure decreased from 39.4 psig to 36.09 psig (because of additional heat ;
j sinks), the maximum calculated time to reestablish a subatmospheric condition L
! decreased from 3350 seconds to 2560 seconds (largely from revisions in the mass, j j and energy release rate data), and the maximum subatmospheric peak pressure j increased from -0.13 psig to -0.07 psig. ,
The staff has performed revised confirmatory calculations to reflect the above i
{ changes. The staff's calculations result in a peak containment pressure (for .
j the hot-leg double-ended rupture) of 36.8 psig using the CONTEMPT-LT/28 computer '
code, and a maximum depressurization time (for the 0.6 DEPSG) of 2720 seconds using the CONTEMPT-4, M006 computer code. The staff's analyses approximately 1 confirm the applicant's calculations. The staff has also performed a revised [
minimum containment pressure analysis using the CONTEMPT-LT/28 computer code :
to determine the impact of the additional heat sink area reported by the 11- (
consee. The CONTEMPT results remain in good agreement with those originally i i provided by the Itcensee for the entire transient. On the basis of these cal- ;
i culations, the staff concludes that the ifcensee has satisfactorily demonstrated l
! the adequacy of the containment functional design, and the minimum containment ;
l pressure analysis. l Millstone 3 SSER 5 6-1 i
l 6.2.1.2 Subcompartment Analyses In FSAR Amendment 16, the licensee reported results of analyses for two addf-tional pipe break locations in the lower pressurizer cubicle subcompartment, and revised the subcompartment design pressure from 20.5 psid to 27.3 psid. As a result of the change in pipe break location, the limiting differential pres-sure for the lower pressurizer cubicle (surge line double-ended rupture) in-creased from 20.31 psid to 24.15 psid. This peak calculated differential pres-sure is below the revised design value for the subcompartment.
The staff has performed a confirmatory analysis of the postulated pipe break using the COMPARE-MODIA computer code. The staff's calculation also produces a peak differential pressure of 24.1 psid. On this basis, the staff finds the licensee's subcompartment analyses acceptable.
6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures License Condition 2.C(9) of the Millstone 3 Low-Power License states that "NNECO shall justify the applicability of the Westinghouse LOFTRAN methodology to the Model F steam generator as contained in Topical Report WCAP-8822-P-52." By letter from J. F. Opeka (NNECO) to V. S. Noonan (NRC) dated January 13, 1986, the licensee provided the requested justification. The ifcensee states that the primary factors which influence containment response to superheat conditions are independent of steam generator type, and consequently, the Westinghouse LOFTRAN methodology is applicable to Millstone 3. The staff has reviewed the IIcensee's justification and concurs with the conclusion that the LOFTRAN meth-odology is applicable to Millstone 3 with the Model F steam generator. There-fore, the staff concludes that its concerns expressed in License Condition 2.C(9) are resolved.
6.6 Intervice Inspection of Class 2 and 3 Components 6.6.3 Evaluation of Compliance With 10 CFR 50.55a(g)
This evaluation supplements conclusions in Section 6.6.3 of the SER (NUREG-1031) which addressed the definition of examination requirements and the evaluation of compliance with 10 CFR 50.55a(g). In previous SER supplements, the staff reported in Section 6.6 that the Preservice Inspection Program for Class 2 and 3 systems and components is consistent with the applicable regulation and Code requirements, and the review is a confirmatory issue contingent upon the licensee completing the required examinations and identifying all impractical preservice inspection requirements with a supporting technical justification.
In a letter dated December 23, 1985, the Itcensee submitted relief requests from ASME Code Section XI requirements which the ifcensee determined to be impractical to perform at Millstone 3 and provided supporting information pursuant to 10 CFR 50.55a(a)(3). The staf f evaluated the ASME Code-required examinations that the licensee determined to be impractical and the staff found that the licensee has demonstrated that compliance with the requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety. On the basis of a review of the ifcensee's submittals and the granting of relief from the preservice examination requirements, the Millstone 3 SSER 5 6-2 l
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! staff concludes that the Preservice Inspection Program for systems and compo-nents within the reactor coolant pressure boundary at Millstone 3 is acceptable and in compliance with 10 CFR 50.55a(g)(3). The detailed evaluation supporting i
! this conclusion is provided in Appendix N to this report. l i
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Millstone 3 SSER 5 6-3
9 AUXILIARY SYSTEMS 9.5 Other Augillary Systems 9.5.1 Fire Protection
- 9. 5.1. 5 Fire Detection and Suppression Ca. bon Oloxide Suppression Ssstem In SSER 4 the staff stated that the license would be conditioned to require the licensee to demonstrate the operability of the gaseous (CO2 ) fire suppression system before initial criticality. By letter from J. F. Opeka (NNECO) to V. S. Noonan (NRC) dated December 3, 1985, the licensee submitted the results of the acceptance tests on the carbon dioxide systems. These results confirmed that the design concentration of carbon dioxide gas was achieved and maintained at all test probes. On this basis, the staff concludes that the carbon dioxide fire suppression systems conform with Section C.6.d of BTP CME 8 9.5-1 and are, therefore, acceptable.
9.5.3 Lighting System In SSER 3, the staff incorrectly stated that the licensee committed, in its letter of July 18, 1985, to increase the de lighting systems illumination level to a minimum of 10 foot-candles in certain areas of the purple switchgear room, control room, orange switchgear room, and diesel generator room. The licensee instead committed to provide a maintained average of 10 foot-candles in those areas. The staff finds this acceptable.
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l Millstone 3 SSER 5 9-1
13 CONDUCT OF OPERATIONS 13.1 Organizational Structure of Applicant 13.1.2 Plant Staff As stated in Low-Power License Condition 2.C(14), training of the shift advisers and the shift crew by the licensee shall be completed and approved by the NRC before 5% power is exceeded.
This safety evaluation presents the staff's assessment of the licensee's provi-sions for hot participation experience on shift at Millstone 3 during low power testing and power operations.
Shift Advisor Qualifications Three shift advisor candidates were selected. All are employees or previous employees of Northeast Utilities (NU), having served at other NU nuclear plants.
Each has had extensive previous nuclear experience, fully meeting the minimum experience levels for shift advisors set forth in Generic Letter 84-16. Each shift advisor meets the medical requirements for licensed operators. The staff concluded that the shift advisor candidates were acceptable.
Shift Advisor Duties and Responsibilities The licensee provided a copy of Station Operating Procedure OP-3262, Operations Shift Advisor, which sets the duties, responsibilities, and qualifications of the shift advisor. The staff reviewed this procedure and concluded that it acceptably describes the duties and responsibilities of the shift advisors.
However, during a telephone conversation with representatives of the NU train-ing department and licensing staff on November 6, 1985, the staff pointed out that Section 6.1 of the procedure should be expanded to include a requirement that shift advisors participate in the licensed operator requalification train-ing program so that they can be cognizant of changes in procedures design and license conditions. In addition, the shift advisors should participate in shift training, including training on the Millstone 3 simulator. On November 8, 1985, the staff was informed by the licensee that Procedure OP-3262 would be expanded as noted above. This commitment was confirmed by a letter from the licensee dated November 12, 1985. The staff concluded that the licensee's de-scription of shift advisor duties and responsibilities was acceptable.
Shift Advisor Training Program The licensee described the initial training program to be presented to the shift advisor candidates to qualify them to serve as shift advisors at Millstone 3.
The staff's initial review of this program indicated that it was generally ac-ceptable. The staff did identify, however, a number of areas in which some revision to the program was in order. The staff discussed these matters with a -
representative of the licensee's training department during a telephone con-versation on November 6, 1985. On November 8, 1985, the staff was informed by Millstone 3 SSER 5 13-1
the licensee.that these revisions to the training program would be made; this commitment subsequently was confirmed by letter dated November 12, 1985. On this basis, the staff concluded that the shift advisor training program was
- acceptable.
Systems Training W staff reviewed the training objectives described in the training program for the equipment and systems described, and considered these objectives suit-able for shift advisor training. However, staff review of the systems portion of the training program concluded that the systems training did not include training on a number of systems that are vital to normal plant operation and others of which the shift advisors should be knowledgeable to provide advice to the operating shifts. These include:
reactor coolant pumps and motors, including lubrication, cooling, and monitoring equipment residual heat removal component cooling reactor protection non-nuclear instrumentation, including incore temperature monitoring l
containment, including normal and emergency cooling reheat steam and feedwater heating main turbine, generator, and condenser, including auxiliary systems ac and de vital power supply service water systems for normal and emergency operations area and process radiation monitors fire protection systems In the letter of November 12, 1985, the licensee committed to include these systems in the training program. The staff concluded that the proposed systems training was acceptable.
Technical Specification Training NU proposed to train shift advisors on the technical specifications that apply to Millstone 3 during operations b modes 1 and 2. However, once assigned, the I
services of shift advisors are required whenever the unit is in operating modes i 1-4. Therefore, the staff required that the training be broadened to include training in technical specifications applicable to operating modes 1-4. Further-more, the staff felt that the objectives of the technical specification training should include the bases for the technical specifications. The licensee com-mitted orally to broaden this training to include technical specifications Millstone 3 SSER 5 13-2
applicable to operation in modes 1-4. This commitment was confirmed in the November 12, 1985, letter, and the staff thus concluded that the technical specification training was acceptable.
Procedure Training The staff questioned whether too much emphasis would be given to abnormal and emergency operation procedures in lieu of concentrating on procedures for normal plant operation. In discussions with the NU staff, NU clarified that the intent of this segment is to concentrate on procedures which are germane to the shift advisor duties for normal operations, while providing a suitable background for abnormal and emergency conditions that may occur. This was acceptable. The NRC staff also questioned the apparent lack of training in administrative procedures which are germane to shift operations. The NU representatives clarified this issue by reference to a September 20, 1985, letter which describes the similar-ity of administrative procedures at NU nuclear plants. By virtue of the exten-sive previous experience of each of the shift advisor candidates at one or more of the other NU nuclear plants, NU considered that special training in adminis-trative procedures was not required. The NRC staff agreed.
Simulator Training The staff review did not disclose any provisions in the training program for the shift advisor candidates to practice problem solving involving use of their knowledge of the procedures, systems, and technical specifications. The staff recommended that problem solving be introduced into both the classroom and simu-lator portions of the program and be included in the final certification examination.
The simulator training for shift advisors consisted of observation of reactor and plant startups, power escalation of at least 25% above 20% power, and normal plant and reactor shutdown. In the letter of November 12, 1985, the licensee confirmed plans to incorporate problem solving specific to Millstone 3 through the use of instrument-failure-induced transients in the simulator training pro-gram. The staff concluded that the planned simulator training was appropriate for shift advisors.
General The staff believed that knowledge of the previous checkout and test experience at Millstone 3 and an awareness of the overall startup and test program planned for Millstone 3 would enable the shift advisors to perform their duties better.
The staff recommended that a brief overview of the past experience and future plans be incorporated into the training program.
In the November 12, 1985, letter, the licensee stated that the shift advisors would be briefed on the past experience at the plant and that they would be brieferJ on the upcoming startup and test programs at the briefings conducted for their respective shifts. The staff concluded that this was acceptable.
On December 10 and 11, 1985, the staff reviewed the written examination for the shift advisors and witnessed simulator exercises conducted with the shift advisors at the Millstone 3 plant-referenced simulator.
l Millstone 3 SSER 5 13-3
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The staff reviewed the. written examination which was administered to the shift I
advisor candidates on December 4, 1985; the answer key was also reviewed. In addition, the staff reviewed grading of the examination by the Millstone 3
. training staff. The examination consisted of 21 questions requiring 39 re-sponses, and covered the areas of Millstone 3 systems, procedures, and techni-cal specifications. The questions were developed from the objectives contained l in lesson plans for the shift advisor training program and are considered to be l at the appropriate level for the candidates.
Review of the grading of the written examinations was conducted using criteria contained in NUREG-1021, " Operator Licensing Examination Standards." The grading
, was acceptable.
In a letter of January 10, 1986, the licensee provided copies of the written examination, answer key, and grades of the advisors. All candidates passed the written examination.
, On December 11, 1985, the staff observed the shift advisors during exercises at d
the Millstone 3 simulator. The exercises were conducted over a 5-hour period
- and consisted of a normal plant shutdown from 60% power with a number of mal-4 functions. The plant shutdown was performed by an operating crew consisting of .
personnel who were preparing for operator licensing examinations. Each shift !
advisor was paired with a licensed or certified instructor / evaluator who also acted as an intermediary with the crew. The evaluations were conducted by use of simulator checklists and were submitted in the letter of January 10, 1986.
1 During the exercise period, the shift advisors were evaluated in the areas of I
plant systems, use of procedures, and technical specifications. The shift ad-
! visors also observed the planning and progress of the plant shutdown by the l operating crew. During the shutdown and after malfunctions, the shift advisors i often questioned the crew's method of resolving the malfunctions and its in-
,! fluence in continuing the planned shutdown. The staff found the setting of the
! evaluations was appropriate and that it presented ample opportunities for the
! shift advisors to act and be evaluated in the role of advisor. The staff also
- found that the evaluation process was conducted in a manner that met the guid-i ance for simulator examinations, as applicable for shift advisors, contained
! in NUREG-1021.
1 In the letter of January 10, 1986, the licensee provided a simulator training
- performance summary for each of the advisors. The summary consisted of an j evaluation of each advisor's i
ability to follow normal plant operations using plant procedures ;
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! - knowledge of Millstone 3 systems, control instrumentation, and reference i material I
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ability to evaluate abnormal and emergency conditions, including problem solving and establishing priorities t
] Although conclusions contained in the summaries are positive, they also include j recommendations for additional training or familiarization. The staff concludes that the evaluations were conducted in an acceptable manner and represent a i thorough assessment of the shift advisors.
Millstone 3 SSER 5 13-4
F Operating Shift Crew Training l In a November 1, 1985, letter, the licensee provided an outline of the planned training of operating shif t crews to ensure they understand the role of shift advisors. The staff reviewed the program and determined the training was appro-priate. During a telephone conversation with the licensee's staff on January 17, 1986, the staff requested confirmation that crew training had been completed.
The licensee plans to inform the staff by letter of the completion of operating crew training.
Conclusion I
! In a Janury 10, 1986, letter confirming the certification of shift advisors, the
! licensee also provided an update of the activities of the shift advisors which includes: additional simulator training with their assigned crews; participa-tion in the Millstone 3 Licensed Operator Requalification Program; and continu-ing involvement in the daily activities of their operating shifts. Subject to receipt of a letter from the licensee confirming completion of operating crew training, the staff concludes that the shift advisor training program has been i completed in an acceptable manner and that the shift advisors are properly j trained and qualified to perform advisory duties.
13.5 Station Administr'ative Procedures 13.5.2 Operating and Maintenance Procedures 13.5.2.3 Reanalysis of Transients and Accidents; Developm'ent of Emergency Operating Procedures i
Section 13.5.2.3 of SSER 4 discussed three open issues related to the Millstone 3 Emergency Response Guidelines (ERGS). These items are (1) guidance for use of the reactor coolant system (RCS) loop isolation (stop) valves during recovery from a steam generator tube rupture (SGTR) event (2) correcting degraded core cooling guidelines (E0P-35, FR-C.2).
(3) including in the Millstone 3 Emergency Operating Procedures (E0Ps) reactor vessel level monitoring system (RVLMS) setpoints corresponding to 50% steam-
] water mixture as provided in the Westinghouse Generic ERGS using the reactor vessel liquid inventory system (RVLIS) a
! In a letter from J. F. Opeka (NNECO) to V. S. Noonan (NRC) dated January 14, 1986, the licensee addressed these items. The staff's evaluation of each i follows:
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(1) Use of Loop Stop Valves The licensee provided for the use of loop isolation valves after a steam gen-
] erator tube rupture (SGTR). Among the conditions required for stop valve use are achievement of stabilization of transient effects of the SGTR event, avail-l ability of offsite power, and adequate subcooling and level.
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The staff finds j the guidance acceptable for immediate implementation based on the limitations j I
j Millstone 3 SSER 5 13-5 i
l for its use assuring an acceptably low likelihood for negative effects and the i potential benefits of loop isolation for some beyond-design-basis scenarios.
I However, the staff recommends that (a) the guidance be retitled to clarify the limitations of its use (by whom, how, and when) and (b) copies of the guidance be appended to, but not be made a part of the E0Ps to facilitate communications
, between operators and technical staff when considering loop isolation.
In the longer term, consistent with resolution of N-1 operation considerations, i the staff requires reassessment of this guidance, with commensuratt modifica-tions including explicit identification of loop isolation considerations as discussed in SSER 4 and identification of situations where loop isolation might be beneficial (relative to other options).
l (2) Degraded Core Cooling Guideline (E0P-35, FR-C.2)
In its submittal, the licensee has stated that the level check (for exiting !
FR-C.2) would be restored to steps 5 and 7a to effect consistency between the approved generic ERGS and the Millstone 3 ERGS. By doing so, the typographical 1 errors noted in the licensee's submittal would be corrected. This action is consistent with the staff's requirement and is acceptable.
(3) RVLMS Setpoints With Pumps j Running--In its submittal, the licensee stated that readings using RVLMS are l
not influenced by the operating status of RCS pumps as are those using the RVLIS in the approved referenced ERGS. The licensee stated that in the RVLIS the deleted steps compensated for pump status and that the same functional require-ment of monitoring core covery is accomplished by the 19% level indi, cation using RVLMS. The licensee indicated that the RVLMS level criterion would be included in the Millstone 3 E0Ps. The staff finds this acceptable, since the licensee states that the function of the level criterion will be satisfied using the 19%
RVLMS indication.
i Conclusions The staff finds that the licensee has resolved the concerns identified in SSER 4
) regarding the Millstone 3 ERGS, and therefore Millstone 3 Low-Power License i Conditions 2.C(15)(a), (b), and (c) have been satisfied. The staff's longer term requirement for improvements to loop isolation guidance for N-1 loop opera-tion must be resolved before the staff approves N-1 loop operation as presented
, in Low-Power License Condition 2.C(6). -
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Millstone 3 SSER 5 13-6
15 ACCIDENT ANALYSES 15.3 Decrease in Reactor Coolant Flow Rate 15.3.3/15.3.4 Reactor Coolant Pump Rotor Seizure and Shaft Break l
I (1) Locked Rotor Accident The licensee has provided an analysis of the postulated locked rotor accident assuming that the atmospheric dump valve on one of the four steam generators fails open. The analysis assumes that the steam generator with the failed-open valve can be isolated within 30 minutes and, as a result of the departure from nucleate boiling, 3% of the fuel rods experience cladding failure. In addition, the licensee assumed (a) a steam dump of 159,000 lb from the affected steam gen-erator before isolation, (b) a steam release from the remaining three steam gen-erators, of 384,000 lb during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the accident and 1,363,000 lb for the remaining 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the accident, (c) a 1 gpm leak from the primary side to the secondary side, and (d) an iodine partition coefficient of 100 for the unaffected steam generators and a partition coefficient of 1 for the af-fected steam generator.
The staff reviewed the licensee's analysis and then performed an independent dose assessment based on the parameters listed in Table 15.1. The staff analy-sis considered two cases: (a) all the leakage from the primary to the secondary side event goes into the affected steam generator and (b) all of the primary to secondary leakage goes into the unaffected steam generators. These two cases give an upper and lower bound for the locked rotor accident. The results of the staff analysis are presented in Table 15.2.
The staff concludes that the distances to the exclusion area and to the low-population zone (LPZ) boundaries for Millstone 3 are sufficient to provide reasonable assurance that the calculated radiological consequences of a poten-tial locked rotor accident, consistent with the failure of an atmospheric dump
- valve and loss of offsite power, would not exceed 10 CFR 100.11 dose guidelines.
15.6 Decrease in Reactor Coolant Inventory 15.6.3 Steam Generator Tube Rupture By letter dated October 25, 1985, the licensee proposed to use the results of the Westinghouse Owners GroLp (WOG) generic program for steam generator tube rupture (SGTR) to resolve the Millstone 3 SGTR licensing issue for both four-loop and three-loop operation. s The subgroup of WOG submitted WCAP-10698 in December 1984, and Supplement 1 to WCAP-10698 in May 1985, to support the resolution of the licensing issues asso-ciated with an SGTR accident. The subgroup also plans to submit an evaluation of the consequences of steam generator overfill resulting from an SGTR.
Millstone 3 SSER 5 15-1
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3 The licensee has indicated that the results of the generic program can be used to demonstrate the margin available for steam generator overfill for both four- !
loop and three-loop operations. On the basis of the comparison of the prelimi-
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nary estimates of the time to overfill for the two plant types presented in ,
WCAP-10698 and a comparison of the other factors that may affect the margin to overfill, the licensee stated that the evaluation would demonstrate increased
! steam generator overfill margin for four-loop operation. The overfill margin i for three-loop operation may be slightly less than for four-loop operation as a result of the differences in initial plant conditions and recovery times associated with the reduced power level for three-loop operation. The staff concludes that the proposed methodology to evaluate an N-1-loop SGTR event is acceptable pending the results of the review.of WCAP-10698.
Supplement 1 to WCAP-10698 has been reviewed by the staff (Mueller, October 22, 1985). The staff concluded that the dose analysis methodology used in the evalua-tion is acceptable, with the exception of the iodine transport models which i were not provided by the licensee.
j WCAP-10698 is currently being reviewed by the staff, and the staff will report
! its findings when the review is completed.
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Table 15.1 Locked rotor accident assumptions Parameter Value Power 3636 MWt Primary to secondary side leak rate 1 gpm Unaffected steam generator 0-2-hour steam dump 384,000 lb 2-8-hour steam dump 1,363,000 lb Iodine partition coefficient 100 Affected steam generator 0-30-min steam dump 159,000 lb Iodine partition coefficient 1 0-2-hour EAB meteorology 5.3 x 10 4 sec/m3 0-8-hour LPZ meteorology 2.7 x 10.s sec/m3 l
l Table 15.2 Locked rotor accident dose calculations EAB LPZ Case Thyroid (whole body) Thyroid (whole body) 1: 0-2 hr 48 0.2 2.5 0 0-8 hr Total 48 0. 2 2.5 0 2: 0-2 hr 10 0.2 0. 5 0 2-8 hr - -
6.4 0 Total 10 0.2 6.9 0 l
Millstone 3 SSER 5 15-3 l
17 QUALITY ASSURANCE 17.6 Independent Design Verification 17.6.4 Staff Assessment 17.6.4.3 Hazard Analysis Program In SSER 4, the staff stated that Stone & Webster Engineering Corporation (SWEC) had reviewed the applicant's " Hazards Review Program Summary," NERM-69, Revi-sion 0 (September 17, 1985) and determined that when the program is success-fully implemented, it will correct conditions identified in the action items and audit observations discussed in SSER 4. The summary report identified the cri-teria and procedures to be followed to document the status of the hazards pro-gram, including high-energy line breaks (HELBs) and moderate-energy line breaks (MELBs) (pipe whip, jet impingement, spray wetting, and flooding) and internal missile postulation.
Since the hazards program had not been completed at the time the Low-Power License was issued, License Condition 2.C(18) of the low-Power License required the licensee to complete its hazards program and provide a schedule for making any required modifications before exceeding 5% power.
In a letter from J. F. Opeka (NNECO) to V. S. Noonan (NRC) dated January 23, 1986, the licensee stated that the hazards program was complete and the results showed that modifications were necessary.
On the basis of this information, the staff concludes that Low-Power License Condition 2.C(18) has been satisfied.
Millstone 3 SSER 5 17-1
18 HUMAN FACTORS ENGINEERING 18.2 Safety Parameter Display System In Supplement 1 to NUREG-0737, the staff identified five critical safety func-tions that would provide an overview of the safety status of the plant. These are (1) reactivity control (2) reactor core cooling and heat removal from the primary system (3) reactor coolant system integrity (4) radioactivity control (5) containment conditions The specific parameter or variables selected to be displayed to represent the critical safety functions are reviewed by the staff for adequacy and basis.
The licensee has chosen different critical safety functions to be consistent with the Westinghouse Owners Group Emergency Response Guidelines and provide displays of the critical safety function status trees used by the guidelines.
These are (1) subcriticality (2) core cooling (3) heat sink (4) integrity (5) containment (6) reactor coolant system inventory (7) radiation release Upon review of these critical safety functions and the specific parameters selected to be displayed to represent the critical safety functions, the staff concluded that four additional parameters were required to adequately present an overview of the safety status of the plant. The staff stated in SSER 4 that:
"The status of the following variables may not be available on the SPOS proposed for Millstone 3:
RHR flow Containment isolation Containment hydrogen concentration Hot leg temperature "During the RHR and emergency core cooling system (ECCS) modes of cooling when steam generators are not available, the viability of the heat removal system should be monitored. The applicant has cited variables used to monitor the ' core cooling safety function' to address ' heat removal safety function' monitoring. In drawing a distinction between ' core cooling' and Millstone 3 SSER 5 18-1
' heat removal' safety functions, a more direct variable (s) should be con-sidered to monitor the heat removal safety function. RHR flow may be suf-ficient to address both of the above cooling modes (RHR and ECCS); i.owever, the applicant should further evaluate the ' heat removal' critical safety function and discuss the SPDS variable (s) that will assess its status during the RHR and ECCS modes of cooling.
" Containment isolation is an important parameter for use in making a rapid assessment of ' Containment Conditions. ' In particular, a determination that known process pathways through containment have been secured provides a significant additional assurance of containment integrity. Containment hydrogen concentration is a key parameter used in the emergency guidelines to monitor combustible gas control and to indicate a compromise of the
' Containment Conditions' safety function.
" Hot leg temperature is a key indicator used in the ERGS (ES-0.1, Attach-ment A, Generic Instrumentation, Revision 1, Page 3) to determine the viability of natural circulation as a mode of heat removal. NUREG-0737, Supplement 1 gives 'RCS Temperature' as a proposed variable, but does not specify hot leg temperature.
"Opeka (May 1985) [ letter from J. F. Opeka (NNECO) to B. J. Youngblood (NRC), dated May 24, 1985] indicates that, although RHR flow, containment hydrogen concentration, and hot leg temperature can be displayed at the SPDS console through various plant computer monitoring systems, the above four variables are not considered necessary to the SPDS by the applicant, and as such are not included in the Millstone SPDS.
"The above variables do, for given scenarios, provide unique inputs to determinations of status for their respective CSFs [ critical safety func-tions], which have not been discussed by the applicant as being satisfied by other variables in the proposed Millstone SPDS list. The applicant should address these variables and their functions by: (1) addirig these variables to the Millstone SPDS, (2) providing alternative added variables along with justifications that these alternates accomplish the same safety functions for all scenarios, (3) providing justification that variables cur-rently on the Millstone SPDS do in fact accomplish the same safety functions for all scenarios, or (4) identifying that these variables are in fact available from the SPDS variables." [(5) The licensee can also provide justification acceptable to the staff for not including these variables.]
The licensee has provided the following responses to the SER. The staff's position regarding these responses is aise included.
RHR Flow For Millstone 3, the RHR system is not used for containment recirculation; therefore, it is not an appropriate variable.
The staff agrees with the licensee's response. However, the licensee should propose an alternate variable for monitoring heat r'moval during this ECCS mode.
Millstone 3 SSER 5 18-2
Containment Isolation The licensee's position is that containment isolation status is not a symptom-oriented variable but is rather a system status, and that the adequacy of con-tainment isolation can be monitored by measuring radiation inside and outside containment. In addition, Millstone 3 operators can easily observe cortainment isolation status by looking at the engineered safety feature (ESF) status panel directly across from the primary SPDS station.
The staff's response is that Supplement 1 to NUREG-0737 . J for the SPDS to display sufficient variables to monitor several critical safety functions that reflect plant safety status including containment conditions. Supplement 1 does not specify that these variables should be symptom-oriented variables. Nor does Supplement 1 forbid the use of system status variables. .In the staf f's judgment, an adequate assessment of containment conditions must include information re-garding the isolation status of containment, in addition to radiation measures inside and outside of containment, because of the function of containment; that is, because the function of containment is to provide a barrier, the status of known process pathways through the containment must be known to assess contain-ment conditions and their possible effect on plant safety status.
The staff would find the current containment isolation (CI) display (on the ESF panel) acceptable on these conditions: (1) it should be defined as part of the SPDS system, (2) a commitment should be made to always retain the relative phys-ical location of the primary SPDS and the CI display or to otherwise ensure that the CI display always remains easily viewable from the primary SPDS station, and (3) a commitment should be made to confirm that the CI display is pattern recog-nizable and that pattern-recognition aspects of the display will be retained in the future if addition or deletion of display tiles is necessary. l Containment Hydrogen Concentration This variable is not monitored until recombiners are turned on in containment and, therefore, does not lend itself to input to one of the critical safety functions on the primary SPDS display.
The staff's position is that containment hydrogen concentration should be added to the containment critical safety function using available sensors when they are in operation (turned on) or any new sensors that may be required in resolu-tion of other regulatory issues. It would also be acceptable to add a separate indicator on the primary SPDS display alerting the operator to various condi-tions of hydrogen concentration in containment.
Hot-Leg Temperature This parameter has been added to the inadequate core cooling (ICC) display.
However, it does not feed into any of the critical safety functions (as defined at Millstone 3) that make up the primary SPDS displays.
The staff's position is that this is not satisfactory and that there should be an indicator on the primary SPDS display alerting the operator to abnormal hot-leg temperature readings.
Millstone 3 SSER 5 18-3
Summary Unresolved items identified in SSER 4, have not yet been resolved and therefore will remain as a condition in the full power license.
Millstone 3 SSER 5 18-4
APPENDIX A CONTINUATION OF CHRONOLOGY OF THE NRC STAFF RADIOLOGICAL REVIEW 0F THE MILLSTONE NUCLEAR POWER STATION UNIT NO. 3 November 12, 1985 Letter from licensee concerning training p ogram for shift advisors.
November 25, 1985 Letter to licensee issuing NPF-44 for 5% power and test-ing for Millstone Unit 3. Transmittal included Facility Operating License N.NPF-44, Appendix A and 8 (Technical Specifications and Environmental Protection Plan, respec-tively), Amendment No. 17 to Indemnity Agreement No. B-39, and Assessment of the Effects of License Duration on Matters Discussed in the FES.
December 3, 1985 Letter to Westinghouse withholding from public disclosure
" Justification for the Use of the W-3 Correlation for the Millstone Unit 3 Steamline Break Analysis," CAW-85-71.
December 3, 1985 Letter from licensee concerning Fire Protection Audit--C0 2 Systems Test Results.
December 6, 1985 Letter to licensee transmitting two copies of Supplement No. 4 to the SER (SSER 4).
December 17, 1985 l Letter to licensee transmitting 20 bound copies of SSER 4.
December 17, 1985 Letter to licensee transmitting FSAR Amendment 17.
December 18, 1985 Letter to licensee concerning 10 CFR 50.54(f) on station blackout.
December 20, 1985 Letter from licensee concerning evaluation of environ-mental effects of main steamline break outside containment. l December 23, 1985 Letter from licensee concerning reactor coolant pump rotor seizure event.
December 23, 1985 Letter from licensee submitting request for relief from pre-service inspection.
December 23, 1985 Letter to licensee transmitting corrected page for Millstone Unit 3 Technical Specifications originally issued November 25, 1985.
December 24, 1985 Letter to licensee concerning Environmental Effects of Main Steamline Break and Seismic Interaction Program.
Millstone 3 SSER 5 1 Appendix A
1 l
l l
l December 26, 1985 Letter from licensee responding to information request regarding station blackout.
January 7, 1986 Letter from licensee concerning evaluation of environ-mental effects of main steamline break outside contain-ment.
January 10, 1986 Letter from licensee concerning training evaluation and certification of shift advisors.
January 13, 1986 Letter from licensee concerning applicability of Westinghouse LOFTRAN methodology to Model F steam generator.
January l',1986 Letter to licensee requesting information concerning station blackout for Millstone 3.
January 14, 1986 Letter from licensee concerning Seismic Interaction Pro-gram at Millstone 3.
January 14, 1986 Letter from licensee concerning evaluation of environmental effects of main steamline break outside containment.
January 14, 1986 Letter from licensee concerning Procedures Generation Package.
January 17, 1986 Letter from licensee concerning evaluation of environ-mental effects of main steamline break outside contain-ment.
Janua ry 20, 1986 Letter from licensee concerning installation, testing, and calibration of inadequate core cooling system.
January 23, 1986 Letter from licensee concerning Hazards Review Program.
l Millstone 3 SSER 5 2 Appendix A
i APPENDIX 8 REFERENCES Combustion Engineering Owners Group (CE0G), letter to D. M. Crutchfield (NRC),
June 1, 1982.
Muller, D. R. (NRC), Memorandum to D. M. Crutchfield, October 22, 1985.
Northeast Nuclear Energy Company (NNECO), NERM-69 (Rev. 0), " Hazards Review Program Summary," September 17, 1985.
U.S. Nuclear Regulatory Commission (NRC), Generic Letter 84-16, " Adequacy of On-Shift Operating Experience for Near-Term Operating License Applicants,"
June 27, 1984.
-- , NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980; Supplement 1, January 1983.
-- , NUREG-0800 (Rev.1), " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," July 1981.
-- , NUREG-1021 (Rev. 1), " Operator Licensing Examination Standards," October 1, 1984.
-- , NUREG/CR-2136, "Effect of Postulated Event Devices on Normal Operation of Piping Systems in Nuclear Power Plants," May 1981.
-- , NUREG/CR-2627, " Inadequate Core Cooling Instrumentation Using Heated Junction Thermocouples for Reactor Vessel Level Measurement," March 1982.
Westinghouse Corp., WCAP-8822-P-52, " Mass and Energy Releases Following a Steamline Rupture," September 1976.
-- , WCAP-10698, " Steam Generator Tube Rupture Analysis Methodology To Deter-mine the Margin to Steam Generator Overfill," December 1984.
-- , WCAP-10698 (Supplement 1), " Evaluation of Offsite Radiation Doses for a Steam Generator Tube Rupture Accident," May 24, 1985.
-- , WCAP-10961-P, "Steamline Break Mass / Energy Releases for Equipment Environ-mental Qualification Outside Containment," October 1985.
Millstone 3 SSER S 1 Appendix B
APPENDIX D ABBREVIATIONS ACRS Advisory Committee on Reactor Safeguards AIB arbitrary intermediate break ASME American Society of Mechanical Engineers BTP Branch Technical Position CET core exit thermocouple (
CFR Code of Federal Regulations l
CI containment isolation CRV Code required volume CSF critical safety function DEPSG double ended pump suction guillotine EAB exclusion area boundary ECCS emergency core cooling system E0P emergency operating procedure ERG emergency response guideline ,
ESF engineered safety feature FSAR Final Safety Analysis Report FW feedwater GDC General Design Criterion (a)
HELB high-energy line break HJTC heated junction thermocouple ICC inadequate core cooling ICCI inadequate core cooling instrumentation IGSCC intergranular stress corrosion cracking IR inner radius ISI inservice inspection ITS inventory tracking system LPZ low population zone MELB moderate-energy line break MFW main feedwater MSLB main steamline break NNECO Northeast Nuclear Energy Company NRC U.S. Nuclear Regulatory Commission NU Northeast Utilities OBE operating basis earthquake Millstone 3 SSER 5 1 Appendix D
PSI preservice inspection l PWR pressurized-water reactor QA quality assurance 1
RCS reactor coolant system RHR residual heat removal RTD resistance temperature device RVLIS reactor vessel liquid inventory system RVLMS reactor vessel level monitoring system SER Safety Evaluation Report SG steam generator SGTR steam generator tube rupture l
5MM subcooling margin monitor SPDS safety parameter display system SRP Standard Review Plan SSER supplement to Sa'fety Evaluation Report SWEC Stone & Webster Engineering Corporation WOG Westinghouse Owners Group WRV weld required volume I
1 l
l Millstone 3 SSER 5 2 Appendix D
APPENDIX F NRC STAFF CONTRIBUTORS This supplemental safety evaluation report is a product of the NRC staff. The NRC staf f members listed below were principal contributors to this report.
Name Title
- Branch
- H. Balukjian Nuclear Engineer Reactor Systems (PWR-A) i L. Bell Nuclear Engineer Reactor Systems H. Brammer Senior Mechanical Engineer Engineering (PWR-A)
J. Buzy Senior Reactor Engineer Maintenance and Training R. Giardina Mechanical Engineer Plant Systems (PWR-A)
R. Goel Mechanical Engineer Plant Systems (PWR-A)
C. Li Mechanical Engineer Plant Systems (PWR-A)
R. Lobel Section Leader Reactor Systems (PWR-A)
F. Orr Reactor Systems Engineer Facilities Operation (PWR-A)
R. Palla Mechanical Engineer Regulatnry Improvements l D. Terao Mechanical Engineer Engineering (PWR-B) l H. Walker Mechanical Engineer Electrical Instrumentation, I and Control Systems (PWR-A) l J. Wilson Section Leader Reactor Systems (PWR-A) '
- Reflects reorganization since Supplement 4 was issued.
Millstone 3 SSER 5 1 Appendix F
APPENDIX N PRESERVICE INSPECTION RELIEF REQUEST EVALUATION INTRODUCTION For nuclear power facilities whose construction permit was issued on or af ter July 1,1974,10 CFR 50.55a(g)(3) specifies that components shall meet the preservice exa.nination requirements set forth in editions and addenda of Sec-tion X: of the ASME Boiler and Pressure Vessel Code applied to the construction of the particular component. The provisions of 10 CFR 50.55a(g)(3) also state that components (including supports) may meet the requirements set forth in subsequent editions and addenda of this Code which are incorporated by refer-ence in 10 CFR 50.55a(b) subject to the limitatius and modifications listed therein.
In a submittal dated December 23, 1985, the licensee requested relief from ASME Code Section XI requirements which the licensee has determined to be impractical and provided supporting information pursuant to 10 CFR 50.55a(a)(3). Therefore, the staff evaluation consisted of reviewing the licensee's submittal to the requirements of the applicable Code edition and addenda and determining if relief from the Code requirements was justified.
TECHNICAL REVIEW CONSIDERATIONS The construction permit for Millstone 3 was issued on August 9, 1974. In accordance with 10 CFR 50.55a(g)(3), components (including supports) which are classified as ASME Code Class 1 and 2 have been designed and provided with access to enable the performance of required preservice examinations.
Verification of as-built structural integrity of the primary pressure boundary is not dependent on the Section XI preservice examination. The applicable construction codes to which the primary pressure boundary was fabricated con-tain examination and testing requirements which by themselves provide the neces-sary assurance that the pressure boundary components are capable of performing safely under all operating conditions reviewed in the FSAR and described in the plant design specification. As a part of these examinations, all of the pres-sure boundary full penetration welds were volumetrically examined (radiographed) and the system was subjected to hydrostatic pressure tests.
The intent of a preservice examination is to establish a baseline reference before the initial operation of the facility. The results of subsequent inser-vice examination can then be compared with the original condition to determine whether changes have occurred. If the inservice inspection results show no change from the original condition, no action is required. Should no baseline data be available, all flaws must be treated as new flaws and evaluated accord-ingly.Section XI of the ASME Code contains acceptance standards that may be used as the basis for evaluating the acceptability of such flaws.
Millrtune 3 SSER 5 1 Appendix N
Other benefits of the preservice examination include providing redundant or alternative volumetric examination of the primary pressure boundary using a test method different from that employed during the component fabrication.
Succe'ssful performance of the preservice examination also demonstrates that the welds so examined are capable of being effectively inspected during the subsequent inservice examinations using a similar test method.
In the case of Millstone 3, a large portion of the preservice examination required by the ASME Code was performed. Failure to perform a 100% preservice examination of the welds identified below will not significantly affect the assurance of the initial structural integrity.
In some instances where the required preservice examinations were not performed to the full extent specified by the applicable ASME Code, the staff may require that these examinations or supplemental examinations be conducted as part of the inservice inspection program. The performance of supplemental examinations, such as surface examinations, in areas where volumetric examination is difficult will be more meaningful after a period of operation. Acceptable preoperational integrity has already been established by similar ASME Code Section III fabrica-tion examinations.
Several of the preservice inspection relief requests involve limitations to the examination of the required volume of a specific weld. The inservice inspection (ISI) program is based on the examination of a representative sample of welds to detect generic service-induced degradation. In the event that the welds identified in the preservice inspection (PSI) relief requests are required to I
be examined again, the possibility of augmented inservice inspection will be evaluated during review of the licensee's initial 10 year ISI program. An augmented program may include increasing the extent and/or frequency of examination of accessible welds.
EVALUATION OF RELIEF REQUESTS The licensee requested relief from specific preservice inspection requirements in a submittal dated December 23, 1985. On the basis of the information sub-mitted by the licensee and the staff's review of the design, geometry, and materials of construction of the components, certain preservice inspection requirements of the ASME Boiler and Pressure Vessel Code,Section XI have been determined to be impractical to perform. The licensee has demonstrated that either (1) the proposed alternative would provide an acceptable level of quality and safety or (2) compliance with the specified requirements of this section would result in hardships or unusual difficulties without a compensating in- l crease in the level of quality and safety. Therefore, pursuant to 10 CFR l 50.55a(a)(3), conclusions that these preservice requirements are impractical l are justified as follows. Citations of the Code refer to the ASME Code, Sec-tion XI, 1980 Edition including Addenda through Winter 1980.
(A) Relief Request PR-1, Examination Categories B-B and 8-D, Pressure-Retaining Welds in the Reactor Vessel For Millstone 3 a volumetric (ultrasonic) examination of essentially 100% of the weld length shall be conducted for the following items in accordance with the ASME Code, Article IWB-2500:
Millstone 3 SSER 5 2 Appendix N
Item Description 8.1.12 Longitudinal shell welds B.1. 21 Head circumferential welds 8.1.22 Lower head meridional welds 8.1.30 Shell-to-flange weld 8.3.90 Nozzle-to-shell welds B.3.100 Nozzle inner radius areas Pursuant to 10 CFR 50.55(a)(g)5(iii), relief is requested from performing the preservice volumetric examination of the inaccessible portions of the subject vessel welds.
Because geometric configuration and permanent obstructions affected the matrix of welds (see December 23, 1985, submittal), a 100% volumetric examination was not possible. The PSI limitations and the specific relief as they apply to i each weld, are noted in the submittal.
The licensee offered the following support for relief / alternative examinations proposea:
(1) The subject welds received both volumetric examination by radiography and surface examinations during fabrication, in accordance with ASME Code Sec-tion III requirements which provide adequate assurance of the structural integrity of the welds.
1 (2) A preservice hydrostatic test was conducted successfully on the class /
pressure boundary of which these welds are a part thereof (IWB-2500-1).
(3) Inservice system leakage tests will be performed per Category 8-P, !
IW8-2500-1.
Staff Evaluation This relief request is acceptable based on the following considerations:
(1) All of the reactor pressure vessel welds passed volumetric examinations during fabrication in accordance with the rules of ASME Code Section III for Class 1 components.
(2) All of the identified welds will be subject to a system pressure test in accordance with Section XI Class 1 requirements.
(3) Accessible portions of the above-listed welds received a preservice volu-metric examination in accordance with the ASME Code Section XI.
(4) The limited Section XI ultrasonic examination, the radiography performed during fabrication, and the hydrostatic test provide an acceptable level of preservice structural integrity.
Millstone 3 SSER 5 3 Appendix N
(B) Relief Request PR-2, Examination Category B-A, Pressure-Retaining Weld in Reactor Vessel Closure Head For Millstone 3, a volumetric examination of essentially 100% of the weld length shall be conducted for tha following items in accordance with the ASME Code, Article IWB-2500:
Item Description B1.21 Circumferential head weld B1.22 Meridional head weld 81.40 Head-to-flange weld Pursuant to 10 CFR 50.55(a)(g)5(iii), relief is requested from performing the preservice volumetric examination of the inaccessible portions of the subject vessel welds.
Geometric configuration and permanent obstructions limited the volumetric examination of the following three welds. Examination data sheets and limita-tion sketches provided in the December 23, 1985, submittal depict the affected areas. Relief is therefore requested from complying with the 100% WRV (weld required volume) coverage of these welds.
(1) Weld No. 101-101, Head-to-Flange Weld Access to this weld is limited to essentially one side only because of the forged flange configuration. Additional limitations from the top side of the weld are due to permanently attached head lifting lugs. Required volume not examinable, s38%.
(2) Weld No. 103-101, Circumferential Head Weld Permanently attached head lifting lugs prevented volumetric examination of N7% of the WRV.
(3) Weld No. 101-1040, Meridional Head Weld A 2.7-in.-diameter repair area (surface concavity) on the weld centerline prohibited sufficient coverage of the WRV in that area. Required volume not examinable, s2%.
The licensee offered the following support for relief / alternative examinations:
(1) The subject welds received both volumetric examination by radiography and surface examinations during fabrication in accordance with ASME Code Sec-tion III requirements which provide adequate assurance of the structural integrity of the welds.
(2) A preservice hydrostatic test was conducted successfully on the Class 1 pressure boundary of which these welds are a part (IWB-2500-1).
(3) In-service system leakage tests will be performed per Category B-P, IWB-2500-1.
Millstone 3 SSER 5 4 Appendix N
l l
Staff Evaluation This relief request is acceptable based on the following considerations:
(1) All of the reactor pressure vessel welds passed volumetric examinations during fabrication in accordance with the rules of ASME Code Section III for Class 1 components.
(2) All of the identified welds will be subject to a system pressure test in accordance with Section XI Class 1 requirements.
(3) Accessible portions of the above-listed welds received a preservice volumetric examination in accordance with the ASME Code Section XI.
(4) The limited Section XI ultrasonic examination, the radiography performed during fabrication, and the hydrostatic test provide an acceptable level of preservice structural integrity.
(C) Relief Request PR-3, Examination Category B-8, Pressure-Retaining Welds in Steam Generators For Hillstone 3, a volumetric examination of essentially 100% of the weld length shall be conducted for the following items in accordance with the ASME Code, Article IWB-2500:
Item Description 82.11 Circumferential shell-to-head welds (pressuri'er) z 82.12 Longitudinal shell welds (pressurizer)
B2.40 Tubesheet-to-head welds (steam generators)
Pursuant to 10 CFR 50.55(a)(g)S(iii), relief is requested from performing the preservice volumetric examination of the inaccessible portions of the subject vessel welds.
Geometric configuration and permanent obstructions limited the volumetric examination of the following listed welds. Examination data sheets and limita-tion sketches depict the affected areas. Relief is therefore requested on complying with the 100% WRV coverage of the welds.
Pressurizer (1) 03-007-SW-J , Shell- to-Upper-Head Weld Permanently installed insulation support ring obstructed part of the required scanning area.
Required volume not examinable, ~9%.
(2) 03-007-SW-F, Shell-to-Lower-head Weld Permanent obstructions (alignment target pads and instrumentation lines) and the geometric configuration--weld transition between plate thickness variations prohibited complete coverage.
Millstone 3 SSER 5 5 Appendix N
Required volume not examinable, s30L (3), 03-007-SW-A, Longitudinal Seam Weld Permanently installed insulation support ring obstructed part of the required scanning area.
Required volume not examinable, s5L i
Steam Generators 03-003-SW-Z S/G A 03-004-SW-Z S/G B 03-005-SW-Z S/G C 03-006-SW-Z S/G D Permanent obstructions- permanent I-beam support columns for each generator restricted scans as shown in the December 23, 1985, submittal.
I Required volume not examinable, s30L The licensee offered the following support for relief / alternative examinations:
(1) The subject welds received both volumetric examination by radiography and surface examinations during fabrication in accordance with ASME Code Sec-tion III requirements which provide adequate assurance of the structural integrity of the welds.
(2) A preservice hydrostatic test was conducted successfully on the Class I i pressure boundary of which these welds are part thereof (IWB-2500-1).
(3) Inservice system leakage tests will be performed per Category B-P, IWB-2500-1.
Staff Evaluation This relief is acceptable for PSI based on the following considerations:
(1) The subject welds received both volumetric (radiography) and surface examinations during fabrication in accordance with ASME Code Section III requirements.
(2) The required volumes not examinable of welds 03-007-SW-J and 03-007-SW-A represent a relatively small percentage of the total.
(3) The examinations performed demonstrate an acceptable level of preservice structural integrity.
(D) Relief Request PR-4, Examination Categories B-L-2 and B-M-2, Internal Surfaces of Pump Casings and Valve Bodies For Millstone 3, a visual examination (VT-3) of the internal surfaces shall be conducted for the following items in accordance with the ASME Code, Article j IWB-2500: i Millstone 3 SSER 5 6 Appendix N
Item Description B12.40 Valve bodies exceeding 4 in. (nominal pipe size)
B12.20 Pump casings Note: Examinations are limited to:
One valve within each group of valves that are of the same design, manu-facturing method, and are performing similar functions in the system.
One pump per group performing similar functions.
Pursuant to 10 CFR 50.55(a)(g)S(iii), relief is requested from performing a preservice visual examination (VT-3).
The licensee offered the following support for relief / alternative examinations:
For the reactor coolant pump casings and 35 valve bodies (see December 23, 1985, submittal) in the reactor coolant, pressurizer, safety injection, and residual heat removal systems, relief is requested from disassembly of an operable valve or pump for performing a preservice visual examination (VT-3).
The requirement to disassemble an operable valve or pump for the sole purpose of performing a visual examination (VT-3) of the internal pressure-retaining boundary is impractical and not commensurate with the increased safety achieved by this inspection. Class 1 valves and pumps are installed in their respective systems, and many have completed functional testing. To dicassemble these items would provide a very small potential for increasing plant safety margins with a very disproportionate impact on expenditures of plant manpower and resources.
The manufacturer's test data will be used in lieu of a preservice visual exami-nation (VT-3). This includes documentation of examinations performed during l fabrication and installation of the subject valves. The examinations performed may include volumetric, surface, and visual examinations, as required by ASME l Code Section II, " Material Specifications for Ferrous and Nonferrous Material,"
and ASME Code Section III, " Construction and Installation Requirements."
Class 1 valves and pumps are subjected to numerous types of nondestructive
~
testing and a rigorous quality assurance program during all stages of fabrica-tion, storage, and installation. These valves and pumps have been found acceptable by the manufacturer, the ASME Authorized Nuclear Inspector, and Northeast Utilities' Quality Assurance group. During maintenance of Class 1 pumps, a visual examination (VT-3) will be performed.
Staff Evaluation The purpose of the VT-3 inspection of the internal surfaces of pumps and valves is to determine if severe degradation of the materials is occurring or has occurred over a period of time. It is unlikely that the valves and pumps in Millstone 3 have experienced material degradation, considering the materials from which the components were fabricated and the inservice performance of similar components in other facilities. To disassemble the valves and pumps at Millstone 3 SSER S 7 Appendix N
this time solely to perform the required Section XI preservice visual examina-tion of the internal surfaces is impractical. The staff has determined that the nondestructive examinations and functional tests performed to date signiff-cantly exceed the requirements of the Section XI visual examination and, therefore, these examinations and tests are an acceptable alternative to the Code requirement.
(E) Relief Request PR-5, Examination Categories 8-D and C-8, Nozzle Inner Radius Sections of Steam Generators and Pressurizer For Millstone 3, a volumetric (ultrasonic) examination of nozzle inner radius (IR) sections for the stear. generator and pressurizer shall be conducted in accordance with the ASME Code, Articles IW8-2500 and IWC-2500, for Code Classes 1 and 2, respectively.
Pursuant to 10 CFR 50.55(a)(g)S(iii), relief is requested from performing the preservice volumetric examination of the nozzle inner radius sections of the steam generator and pressurizer as listed below:
Class Component Nozzle (IR) 1 Pressurizer Surge 03-007-SW-S(IR)
Spray 03-007-SW-E(IR) 1 Pressurizer Relief 03-007-SW-D(IR) 1 Pressurizer Safety 03-007-SW-A(IR) 1 Pressurizer Safety 03-007-SW-8(IR) 1 Pressurizer Safety 03-007-SW-C(IR) 1 Steam generator Primary inlet 03-003-IR Primary inlet 03-004-IR Primary inlet 03-005-IR Primary inlet 03-006-IR 1 Steam generator Primary outlet 03-003-IR Primary outlet 03-004-IR Primary outlet 03-005-IR Primary outlet 03-006-IR 2 Steam generator Feedwater 03-053-SW-R-IR Feedwater 03-054-SW-R-IR Feedwater 03-055-SW-R-IR Feedwater 03-056-SW-R-IR 2 Steam generator Steam outlet 03-053-SW-T-IR Steam outlet 03-054-SW-T-IR Steam outlet 03-055-SW-T-IR Steam outlet 03-056-SW-T-IR 8 Appendix N Millstone 3 SSER 5
Reasons for the licensee's request are:
(1) Currently available ultrasonic examination techniques do not provide results which yield a meaningful baseline for comparison with subsequent examinations. This is a result of complex geometrical configuration, long metal paths, limited accessible scan areas, and cladding on some of the nozzles. (Refer to the drawings of the applicable nozzles depicting the complex geometries in the December 23, 1985, submittal.)
(2) The areas involved have received extensive surface examination during fabrication and reveal no indications of cracking. In addition, the general area has been interrogated with ultrasonics during PSI nozzle-to-l vessel weld examinations.
(3) Cracking in the inner radius areas is generally attrit,uted to thermal cycling when it has occurred. Hence, for PSI, there is little or no safety impact as a result of not performing these examinations.
The licensee proposed the following alternative examinations:
(1) Before operation, the licensee will assemble a file of as-built informa-tion for each of the steam generator and pressurizer nozzles to support application of new examination techniques riuring ISI. This information will include factors which could impact a sound beam directed at this area from the OD.
(2) Follow industry progress with technique development and work toward the performance of a meaningful ultrasonic examination during ISI.
Staff Evaluation This relief request is acceptable for PSI based on the following considerations:
(1) All pressure-retaining components were hydrostatically tested to the re-quirements of ASME Code Section III before plant <tartup.
(2) The staff review of the design configurathn of the nozzle inner radius has concluded that the Code-required v'O "atric examination is impractical.
The staff has determined that perfont'rc ne ASME Section III hydrostatic test along with the surface exam' # 'in i an acceptable alternative.
(3) The staff will continue to evaluate the development of new or improved procedures and will require that these procedures be made part of the ISI examination requirements.
(F) Relief Request PR-7, Examination Category C-8, Nozzle-to-Shell Welds and Nozzle Inside Radius Sections of the Residual Heat Removal Heat Exchangers For Millstone 3, volumetric and surface examination of the nozzle-to-shell weld (C2.21) and volumetric examination of the nozzle inside radius section (C2.22) for the residual heat removal heat exchangers shall be conducted in accordance with the ASME Code, Article IWC-2500.
Millstone 3 SSER 5 9 Appendix N
i Pursuant to 10 CFR 50.55(a)(g)5(fii), relief is requested from performing the preservice volumetric examination of the nozzle inner radius sections and sur-face and volumetric examinations on the nozzle-to-shell welds of the residual heat removal heat exchangers.
The licensee gave the following reasons for requesting relief:
(1) The residual heat removal heat exchanger nozzle-to-shell welds and nozzle inside radius areas are totally inaccessible to RT, UT, and surface tech-niques because permanently welded reinforcement plates have been placed over the welds. (See Figure IWC-2500-4(c) from the 1983 Edition of Sec-
, tion XI which accurately depicts the construction.)
(2) The 1983 Edition of Section XI has recognized and taken action to resolve the need for relief as this is a generic problem with many heat exchanger designs. Since the NRC has accepted the 1983 Edition for use, the licensee is utilizing this to support its request for relief.
- The licensee proposed the following alternative examinations
(1) The subject welds re_eived both volumetric examination by radiography and surface examinations during fabrication in accordance with ASME Code Sec-tion III requirements which provide adequate assurance of the structural integrity of the welds.
(2) A preservice hydrostatic test was conducted successfully on the Class 2 '
pressure boundary of which these welds are a part thereof (IWC-2500-1).
4 i (3) Inservice system leakage tests will be performed per Category C.-H, IWC-2500-1.
1 l Staff Evaluation l This relief request is acceptable for PSI based on the following considerations:
4
, (1) Although the 1983 Edition of Section XI has not been accepted for use by i the NRC at this time as stated by the licensee, the staff is in agreement c with the examination requirements of the 1983 Edition for welds covered by l reinforcement plates.
(2) The staff's review of the design configuration of the nozzle inner radius has concluded that the Code-required volumetric examination is impractical.
The staff has determined that the ASME Code Section III examinations demonstrate an acceptable level of preservice structural integrity.
l (3) The subject weld area received radiographic examination and a hydrostatic test during fabrication in accordance with ASME Code Section III require-ments. An ultrasonic examination has been performed on the nozzle-to-vessel welds per ASME Code Section XI requirements.
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i (G) Relief Request PR-8, Examination Category B-D, Nozzle-to-Vessel Welds in j Steam Generators and Pressurizer For Millstone 3, volumetric (ultrasonic) examination of 100% of the full pene-tration nozzle-to-vessel welds for the steam generator and pressurizer shall be conducted in accordance with the ASME Code, Article IWB-2500 for Code Class 1.
i Pursuant to 10 CFR 50.55(a)(g)S(fii), relief is requested from performing the i preservice volumetric examination on the inaccessible portions of the nozzle-to-shell welds of the steam generator and pressurizer.
! The licensee requested relief because:
(1) Geometric configuration of the nozzle-to-shell welds listed below and their close proximity to one another limits the volume that can be examined.
I (2) Scanning is limited to one side only with a -V technique. Restriction on j axial scan is due to the close proximity of the welds to each other.
\ Pressurizer 03-007-SW-A 03-007-SW-C 03-007-SW-E (spray r.ozzle) j 03-007-SW-8 03-007-SW-0 03-007-SW-S (surge nozzle)
Required volume not examinable, s80%.
Steam Generators Coverage is from both sides of weld with -V technique. Restriction on axial l scan is due to the steam generator supports integral extensions.
1
- 03-003-SW-V inlet 03-004-SW-II outlet 03-006-SW-V inlet 1 03-003-SW-II outlet 03-005-SW-V inlet 03-006-SW-II outlet l 03-004-SW-V inlet 03-005-SW-II outlet l Required volume not examinable, s10L I The licensee supported its request for relief / alternative examinations as follows
i i (1) The subject welds received both volumetric examination by radiography and surface examinations during fabrication in acccrdance with ASME Code Sec-
! tion III requirements which provide adequate assurance of the structural l integrity of the welds.
l l
(2) A preservice hydrostatic test was conducted successfully on the Class 1 pressure boundary of which these welds are a part thereof (IWB-2500-1).
(3) Inservice system leakage tests will be performed per Category B-P, IWB-2500-1.
\
Millstone 3 SSER 5 11 Appendix N
Staff Evaluation This relief request is acceptable for PSI based on the following considerations:
(1) All pressure-retaining components were hydrostatically tested to the re-quirements of ASME Code Section III before plant startup.
(2) The staff's review of the design configuration of the nozzle inner radius has concluded that the Code-required volumetric examination is impractical.
The staff has determined that performing the ASME Code Section III hydro-static test along with the surface examination is an acceptable alternative.
(3) The staff will continue to evaluate the development of new or improved procedures and will require that these procedures be made part of the ISI examination requirements.
(H) Relief Request PR-10, Examination Category B-J, Pressure-Retaining Welds in Main Coolant Piping System For Millstone 3, volumetric (ultrasonic) and surface examination of essentially 100% of tN length of each weld in the main coolant piping system shall be conducted in accordance with the ASME Code, Article IWB-2500.
Pursuant to 10 CFR 50.55(a)(g)5(iii), relief is requested from performing the preservice volumetric examination of the inaccessible portions of the welds in Table 1 of the December 23, 1985, submittal for the following reasons:
(1) Geometric configuration, permanent obstructions, and structural interfer-ences prohibit 100% volumetric exam coverage of the Code-required examina-tion volume. Relief is therefore requested from performing preservice examination on the inaccessible portions of the volume required as noted in Table 1.
(2) It should be noted that the Westinghouse-developed UT technology for CCSS piping was utilized in performing all examinations on the main coolant piping. Refer to J. F. Opeka's letter to B. J. Youngblood, 811576, dated May 7, 1985, for NUSCO's response to the staff question (SER Question 250.12) which addresses NRC concerns.
1 (3) A0 longitudinal beam examination was conducted on all CCSS welds to map ID geometry contours. This was done in addition to Section XI requirements to aid in the performance and evaluation of angle beam examination results.
The licensee proposed the following alternative examinations:
(1) The subject welds received both volumetric examination by radiography and surface examinations during fabrication in accordance with ASME Code Sec-tion III requirements. Having met these requirements, adequate assurance of the structural integrity of the subject welds is provided.
(2) A preservice hydrostatic test was conducted successfully on the Class 1 pressure boundary, of which these welds are a part thereof, per IWB-2500-1 requirements.
Millstone 3 SSER 5 12 Appendix N
(3) Inservice system leakage tests will be performed per Category B-P, IWB-2500-1, as well as surface and volumetric exams as required by Sec-tion XI, " Selection Criteria." Any advances in UT technology will be evaluated to determine its application for achieving maximum volume coverage and results.
Staff Evaluation The staff has reviewed the December 23, 1985, submittal including the tables identifying the welds for which relief is being requested. These tables list the weld number, configuration (pipe-to elbow, pipe-to-nozzle, pipe-to pump casing, etc.), material type, and the licensee's estimate of the percentage of the Code-required examination that was completed.
Complete examinations meeting the requirements of ASME Code Section XI were performed on welds of similar configurations which utilized the same weld techniques, procedures, and materials. The inspected welds will be subject to the same operating and environmental conditions as the partially inspected welas or the uninspected weld. Therefore, the acceptable preservice examina-tion results of the inspectable welds provide reasonable assurance by sampling
} of the structural integrity of the subject welds.
On this basis, the staff has concluded that the limited Section XI volumetric examination and the Section III fabrication examinations provide an acceptance level of preservice structural integrity and that compliance with the specific requirements of Section XI would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
(I) Relief Request PR-11, Examination Category B-J, Pressure-Retaining Welds in Piping For Millstone 3, volumetric (ultrasonic) and surface examination of essentially 100% of the length of each Code Class 1 piping weld > 4-in, nominal pipe size shall be conducted in accordance with the ASME Code,Section XI, Article IWB-2500.
~
Pursuant to 10 CFR 50.55(a)(g)S(fii), relief is requested from performing the preservice volumetric examination of the inaccessible portions of the welds listed in Table 1 of the December 23, 1985, submittal.
Permanent structural interferences prohibit 100% volumetric examination coverage of the Code-required volume (CRV). Relief is therefore requested from perform-ing preservice examinations on the inaccessible portions of the volume required as noted in Table 1 of the December 23, 1985, submittal.
j The licensee proposed the following alternative examinations:
(1) The subject welds received both volumetric examination by radiography and
, surface examinations during fabrication in accordance with ASME Code Sec-l tion III requirements. Having met these requirements, adequate assurance l of the structural integrity of the subject welds is provided.
1 l
j Millstone 3 SSER 5 13 Appendix N
(2) A preservice hydrostatic test was conducted successfully on the Class 1 preservice boundary, of which welds are a part thereof, per IWB-2500-1 requirements.
(3) Inservice system leakage tests will be performed per Category B-P, IWB-2500-1, as well as surface and volumetric exams as required by Sec-tion XI selection criteria. Any advances in UT technology will be evaluated to determine its application for achieving maximum volume coverage and results.
Staff Evaluation The staff has reviewed the December 23, 1985, submittal including the welds, configurations, limitations, and percent coverage of the required examinations.
The examinations performed during fabrication to Section III requirements, the preservice hydrostatic test, and the percentage of each weld examined during preservice examination provide adequate bases for acceptance of the welds' structural integrity.
(J) Relief Request PR-12, Examination Category C-F, Pressure-Retaining Welds in Piping For Millstone 3, volumetric (ultrasonic) and surface examination of essentially 100% of the length of each weld requiring examination (> 4-in. nominal pipe size, > l-in. s thickness) shall be conducted in accordance with the ASME Code, Article IWC-2000.
Pursuant to 10 CFR 50.55(a)(g)5(iii), relief is requested from performing the preservice volumetric and/or surface examinations on the inaccessible portions of the welds listed in Table 1 of the December 23, 1985, submittal.
Geometric configuration, permanent obstructions, and/or structural interfer-ences prohibit 100% examination coverage of the Code-required volume or area.
Relief is therefore requested from performing preservice examinations on the inaccessible portions.
I The licensee proposed the following alternative examinations:
1 (1) The subject welds received volumetric examination by radiography during fabrication in accordance with ASME Code Section III requirements. Having met these requirements, adequate assurance of the structural integrity of the subject welds is provided.
(2) A preservice hydrostatic test was conducted successfully on the Class 2 preservice boundary, of which welds are a part thereof, per IWC-2500-1 requirements.
(3) Inservice system leakage tests will be performed per Category C-H, IWC-2500-1 as well as surface and volumetric exams as required by Section XI selection criteria. Any advances in UT technology will be evaluated to determine its aoplication for achieving maximum volume coverage and results.
Millstone 3 SSER 5 14 Appendix N
Staff Evaluation This relief request is acceptable for PSI based on the following considerations:
(1) The subject welds received volumetric examination by radiography during fabrication in accordanco with ASME Code Section III requirements.
(2) The welds weie ubjected to a preservice hydrostatic test.
(3) The percentage of each of the required volumes of the welds to be examined by UT is relatively high, ranging from 74% to 98L CONCLUSIONS On the basis of the foregoing, pursuant to 10 CFR 50.55a(a)(3), the staff has determined that certain Section XI required preservice examinations are im-practical. The licensee has demonstrated that either (1) the proposed alterna-tives would provide an acceptable level of quality and safety or (2) compliance with the requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
The staff's technical evaluation has not identified any practical method by which the existing Millstone Nuclear Power Station, Unit 3 can meet all the specific preservice inspection requirements of Section XI of the ASME Code.
Requiring exact compliance with all Section XI-required examinations would delay startup of the plant in order to redesign a significant number of plant systems, obtain sufficient replacement components, install the new components, and repeat the preservice examination of these components. Even after the redesign efforts, complete compliance with the preservice examination require-ments probably could not be achieved. However, the as-built structural inte-grity of the existing primary pressure boundary has already been established by the construction code fabrication examinations.
On the basis of the staff's review and evaluation, it is concluded that the public interest is not served by imposing certain provisions of Section XI of the ASME Code that have been determined to be impractical. Pursuant to 10 CFR 50.55a(a)(3), relief is allowed from these requirements which are impractical to implement.
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l Millstone 3 SSER 5 15 Appendix N
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3 TITtt A%0 SU8Taf tt 4 **tCiestN T S ALCt 55#0% Nove6 M SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF MILLSTONE NUCLEAR POWER STATION, UNIT N0. 3 , o u . .. ,0 , co ,a , . o MONTM VEAM JANUARY 1986 i
6 AU T MOHei , Def t HEP T ISSU E D MON T M f viAR 1986 JAfARY 9 P' st CT e re$a WQMa 9% t Nuvete S Pt af OMuiNG CMGANi2 A fl0% Navt AN0 MarteNG A ME SS f,em t # Coder Division of Pressurized Wat r Reactor ;
Licensino-A ,o ,,~ ~uo. a Office of Nuclear Reactor Re_ lation U. S. Nuclear Regulatory Comm 'sion Washington, D. C. 20555 o seussoa.w o u Aw.Y,c, s.v. .so ...u% .m.ss ,,- .e,c. u. > V,e o, ai,oa r Same As 8. above Technical 120 ,$ N tOD CQv { N g Q (,sertges.og d.res, November 1985 - January 1986 13 $uPPtt Vt NT A4 v NOT E S Docket No. 50-423 14 A65f M AC T ,200 -ords e' ess The Safety Evaluation Report issued in u st 1984 provided the results of the NRC Staff review of Northeast Nuclear ners Company's application for a license to operate the Millstone Nuclear Pow Stat n, Unit No. 3. Supplement No. I to that report, issued in March 1985 update. the inf .ation contained in the Safety Evaluation Report and addressed th ACRS Repo t issued on September 10, 1984.
Supplement Nos. 2 and 3 dated Sep mber 1985 a d November 1985, respectively, updated the information contained in the afety Evaluat on Report and Supplement No.1 and addressed prior unresolved item Supplement N 4, issued in November 1985 addressed the items concerning the issua e of a low power icense (5%). The low power license, NPF-44 was issued on November 5, 1985.
This Supplement No. 5 addres es the items relating o the issuance of a full power license (100%).
The facility is located i Waterford Township, New London County, Connecticut.
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