ML17255A483

Safety Evaluation Report Related to a Full-Term Operating License for the R.E. Ginna Nuclear Power Plant.Docket No. 50-244.(Rochester Gas and Electric Company)

ML17255A483
Person / Time
Site:	Ginna
Issue date:	10/31/1983
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0944, NUREG-944, NUDOCS 8310310109
Download: ML17255A483 (166)

Text

NUR EG-0944 Safety Evaluation Report related to the full-term operating license for R. E. Ginna Nuclear Power Plant Docket No. 50-244 Rochester Gas and Electric Corporation U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1S83 c>

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ABSTRACT The Safety Evaluation Report for the full-term operating license application filed by Rochester Gas and Electric Corporation for the R. E. Ginna Nuclear Power Plant has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is located in Wayne County, Rochester, New York. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can continue to be operated without endangering the health and safety of the public.

Ginna SER

TABLE OF CONTENTS

~Pa e ABSTRACT . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ill ACRONYMS AND INITIALISMS ix 1 INTRODUCTION AND DISCUSSION .

1.1 Introduction . 1-1 1.2 Description of Plant 1-4 1.3 Summary of Operating History and Experience . 1-5 2 SITE CHARACTERISTICS . 2-1 2.1 Geography and Demography .. 2-1 2.2 Nearby Industrial, Transportation, and Military Facilities ..... 2-1 2.3 Meteorology ... 2-3 2.4 Hydrologic Engineering 2" 3 2.5 Geology and Seismology 2-5 3, DESIGN CRITERIA - STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS .... 3-1 3.1 Classification of Structures, Systems, and Components . 3-1 3.2 Wind and Tornado Loadings . 3-1 3.3 Flood Level Design .............. 3-3 3.4 Missile Protection .. 3-4 3.4.1 Tornado Missiles ...... 3-4 3.4.2 Turbine Missiles ..................... 3-5 3.4. 3 Internally Generated Missiles 3-5 3.4.4 Si te Proximi ty Mi ss i 1 es .............,... 3-6 3.5 Effects of Pipe Break on Structures, Systems, and Components ... 3-6 3.6 Seismic Design Considerations 3-8 3.7 Design of Seismic Category I Structures ...... ~, ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3-10 3.7.1 Design Codes ... 3-10 3.7 ' Containment Structure ... 3-11 3.8 Flooding of Safety-Related Equipment 3-12

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4 REACTOR ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ I ~ ~ I ~ ~ 4-1 4.1 Fuel System Design . 4-1

,4.2 Operation With Less Than All Loops in Ser'vice 4-1 4.3 Loose-Parts Monitoring 4-2 4.4 Reactor Materials ............,. 4-3 4.5 Reactivity Control Systems Protection'Against Single Failure ... 4-3 4.6 Operating Problems .

Ginna SER

TABLE OF CONTENTS (Continued)

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4. 6. 1 Fuel Failures

4. 6. 2 Fuel Densification, Cladding'ollapse, and Fuel Rod Bowing .

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5. 1 Summary Description 5-1 5.2 Integrity of Reactor Coolant Pressure Boundary 5-1 5.3 Reactor Vessel 5-2 5.4 Component and Subsystem Design 5-3

5. 4. 1 Reactor Coolant Pumps 5" 3 5.4.2 Steam Generators 5-5 5.4.3 Residual Heat Removal System 5-8 5.4.4 Pressurizer Power-Operated Relief Valves 5-9 6 ENGINEERED SAFETY FEATURES

6. 1 Engineered Safety Features Materials 6-1 6.2 Containment Systems 6-2

6. 3 Emergency Core Cooling System .. 6-3 7 INSTRUMENTATION AND CONTROLS . 7-1

7. 1 Reactor Protection System . 7-1 7.2 Engineered Safety Features Actuation System (ESFAS)......... ~ ~ ~ 7-3 7.3 Systems Required for Safe Shutdown .. 7-5 7.4 Other Systems. Required for Safety ~ ~ ~ 7-5 7.4.1 Overpressure Protection During'Low Temperature Operation 7-5 7.4.2 .'ngineered Safety Feature Switchover From Injection to Recirculation Mode 7-6 7.4.3 Accumulator, Isolation Valves Power and Control System Design . 7-6 7.4.4 Frequency Decay 7-7 7.4.5 Electric Type Valve Operators 7-7 8 ELECTRIC POWER SYSTEMS 8-1

8. 1 Potential Equipment Failures Associated With Degraded

,Grid Voltage . 8-1 8.2 Adequacy of Station Electric Distribution System Voltages .. ~ ~ 8-1

8. 3 Onsite Emergency Power Systems Diesel Generator ...;-...... 8-2 8.4 Station Battery Capacity Test Requirements 8-2 8.5 DC Power System Bus Voltage Monitoring and Annunciation .... ~ ~ ~ 8-3 8.6 Electrical Penetrations of Reactor Containment 8" 3 Ginna SER vi

TABLE OF CONTENTS (Continued)

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9 AUXILIARY SYSTEMS .. 9-1

9. 1 Fuel Stor age .. 9-1 9.2 Water Systems . 9-1 9.3 Boron Addition System 9-3 9.4 Chemical and Volume Control System . 9-4 9.5 Ventilation Systems 9-5 9.6 Fire Protection .

10 STEAM AND POWER CONVERSION SYSTEM 10-1

10. 1 Steam and Feedwater System 10-1 10.2 Turbine Disc Cracks 10-2 10.3 Secondary Water Chemistry . 10-2

10. 4 Auxiliary and Standby Auxiliary Feedwater Systems ... 10-3 RADIOACTIVE WASTE SYSTEM 11.1 Radioactive Liquid Effluent .. ll-l 11.2 Radioactive Gaseous Effluent . ll-l 11-1 11.3 Radiological Effluent Technical Specifications 12 RADIATION PROTECTION .. 12-1 13 CONDUCT OF OPERATIONS . 13-1

13. 1 Organizational Structure 13-1 13.2 Emergency Preparedness Evaluation . 13-2 13.3 Physical Security Plan 13 2 II 14 INITIAL TEST PROGRAM 14-1 15 ACCIDENT ANALYSIS 15-1 16 TECHNICAL SPECIFICATIONS . 16-1 17 QUALITY ASSURANCE 17-1 18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS,... 18-1 19 COMMON DEFENSE AND SECURITY. 19-1 20 FINANCIAL QUALIFICATIONS . 20-1 21 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS '.. 21-1 22 CONCLUSIONS. 22-1 APPENDIX A REFERENCES APPENDIX B THREE MILE ISLAND - LESSONS LEARNED REQUIREMENTS APPENDIX C UNRESOLVED SAFETY ISSUES LIST OF FIGURES Schematic Diagram of Westinghouse-Designed Pressurized-Water Reactor of Rochester Gas and Electric Corporation's Ginna Nuclear Power Plant .

1-6 Ginna SER v) 1

TABLE OF CONTENTS (Continued)

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5. 1'inna Reactor Coolant System Diagram .... 5-11 5.2 Ginna Steam Generator 5-12 10.1 Flow Diagram of Main Steam and Reheat System 10-5 10.2 Flow Diagram of Condensate System . 10-6 10.3 Flow Diagram of Main Feedwater Sytem 10-7 4

'J.l. 1 Liquid Radwaste Treatment Systems, Effluent Paths, and Controls for Ginna Nuclear Power Plant 11-2 11.2 Gaseous Radwaste Treatment Systems, Effluent Paths, and Controls for Ginna Nuclear Power Plant 11-3

13. 1 Rochester Gas and Electric Corporation Management Organization Chart 13-4 13.2 R. E. Ginna Nuclear Powe'r Plant Organization Chart 13-5 13.3 R. E. Ginna Nuclear Power Plant Administrative and Health Physics and Chemistry Sections 13-6 13.4 R. E. Ginna Nuclear Power Plant Maintenance and Nuclear Assurance Sections 13-7 13.5 R. E. Ginna Nuclear Power Plant Operations and Technical Sections 13-8 LIST OF TABLES 5.1 Materials of Construction of the Reactor Vessel 5-13 5.'2 Reactor Vessel Design Data 5-13 5.3 Comparison of Blowdown Chemistry .. ........... .............. 5-14 Ginna SER viii

ACRONYMS AND INITIALISMS ACRS Advisiory Committee on Reactor Safeguards AEC Atomic Energy Commission, U.S.

AEOD Office of Analysis and Evaluation of Operational Data AFW auxiliary feewater ALARA as low as is reasonably achievable ANSI American National Standards Institute ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials ATWS transient(s) without scram 'nticipated AVT all-volatile treatment B&W Babcock 8 Wilcox BTP branch technical position BWR boiling-water reactor CCW component cooling water CE Combustion Engineering CFR cfs cubic feet per second CRD control rod drive CVCS chemical and volume control system DBA design-basis accident DG diesel generator DNB departure from nucleate boiling DOR Division of Operating Reactors ECC emergency core cooling ECCS emergency core cooling system ECT eddy current testing EFPY effective full-power year EILC electrical instrumentation and control ENC Exxon Nuclear Corporation EPIA emergency preparedness implementation appraisal Eg environmental qualification ESF engineered safety feature(s)

ESFAS engineered safety features actuation system ESSA Environmental Science S'ervices Administration FES final environmental statement N FR Federal Receister FRC Franklin Research Center FTOL full-term operating license GDC general design criterion(a) gpm gallon(s) per minute HELB high energy line break HEPA high-efficiency particulate air HPSI high-pressure safety injection ID inner diameter IE Office of Inspection and Enforcement IPSAR integrated plant safety assessment report Ginna SER ix

ISI inservice inspection LLL Lawrence Livermore Laboratory LOCA loss-of-coolant accident LWR light-water reactor MOX mixed oxide mph mile(s) per hour MSL mean sea level MSLB main steam line break MWt megawatts thermal mybp million years before present NEPA National Environmental Policy Act NFPA National Fire Portection Association NPSH net positive suction head NRC Nuclear Regulatory Commission, U.S.

NSSS nuclear steam supply system OBE operating-basis earthquake ODCM Offsite Dose Calculation'anual OPS overpressure protection system ORNL Oak Ridge National Laboratory PCT peak cladding temperature PMF probable maximum flood POL provisional operating license PORC Plant Operations Review Committee PORV power-operated relief valve ppb part(s) per billion ppm part(s) per million PRA probabi listic risk assessment psi pound(s) per square inch psia pounds(s) per square inch absolute psl g pounds(s) per square inch gage PTS pressurized thermal shock PWR pressurized-water reactor RCC rod cluster control RCP reactor coolant pump RCS reactor coolant system RETS Radiological Effluent Technical Specifi cation RG regulatory guide .

RG8(E Rochester Gas and Electric Corporation RHR residual heat removal RPS reactor protection system NDT nil-ductility transition reference temp erature SALP Systematic Assessment of Licensee Perfo rmance SEP Systematic Evaluation Program SER safety evaluation report SFPCS spend fuel pool cooling system SGTR steam generator tube rupture SIS safety injection system SQUG Seismic Qualification Utility Group SPF standard project flood SRP Standard Review Plan SSE safe shutdown earthquake SWS service water system TAP task action plan Ginna SER

TER technical evaluation report

, TMI Three Mile Island TMI-2 Three Mile Island Unit 2 UPI upper plenum injection USGS U.S. Geological Survey USI unresolved safety issue W Westinghouse WOG Westinghouse Owners Group Ginna SER

1 INTRODUCTION AND DISCUSSION C

l. 1 Introduction This report is a Safety Evaluation Report (SER) on the application for a full-term operating license (FTOL) for the R. E. Ginna Nuclear Power Plant (Ginna or the facility), based on an application filed by Rochester Gas and Electric Corporation (RG8E or the licensee). This report was prepared by the U.S.

Nuclear Regulatory Commission (the staff) and summarizes the results of the staff's review of the proposed conversion from a provisional operating license (POL) to an FTOL:

From 1959 to 1971 the Atomic Energy Commission issued POLs to 15 power reactors for periods up to 18 months as an intermediate stage before issuing an FTOL.

. The purpose of the POL was to provide an interim period of routine operation during which the licensee and staff could assess plant operating parameters and performance against predicted values and resolve generic concerns identified during the licensing process.

cation for conversion to a

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POLs have been held longer than 18 months because each POL licensee submitted a timely application for renewal under Part 2. 109 full-term license.

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RG&E filed an application to convert POL No. DPR-18 for Ginna to an FTOL in a letter dated August 15, 1972. The facility received its POL on September 19, 1969, achieved initial criticality on November 9, 1969, and commenced electric power generat'ion on December 2, 1969.

In 1975, because of a large backlog of unresolved generic issues that were relevant to the operation of the POL plants, the staff stopped its review of the POL conversions and set out to establish the appropriate scope of review needed to support the full-term conversion.

r In 1977 the NRC staff recommended to the Commission that POL facilities be included in Phase II of the Systematic Evaluation Program (SEP) because much of the review necessary for conversion of the POLs was similar to the scope of the review proposed for the SEP. That recommendation was adopted, and the major portion of the technical input supporting this SER comes from the SEP topic evaluations and the SEP Integrated Plant Safety Assessment Report (IPSAR) for Ginna (NUREG-0821).

The SEP was conceived in recognition that because of the evolutionary nature of licensing requirements and advances in technology, better documentation was needed to substantiate the staff's opinion that currently operating plants are acceptably safe. The objectives established for the SEP were listed on page 3 of SECY-76-545 as:

"The POL review is documented in a Safety Evaluation forwarded by letter dated June 24, 1969.

Ginna SER

(1) The Systematic Evaluation Program must assess the safety adequacy of the design and operation of currently licensed nuclear power plants.

(2) The program should establish documentation which shows how each operating plant reviewed compares with current criteria on significant safety issues, and should provide a rationale for acceptable departures from these criteria.

(3) The program should provide the capabil,ity to make integrated and balanced decisions with respect'o any required back-fitting.

(4) The program should be structured for early identification and resolution of any significant deficiencies.

(5) The program should efficiently use available resources and minimize requirements for additional resources by NRC or industry.

Thus, the review provides (1),an assessment of the significance of differences between current technical positions on safety issues and those that existed when a particular plant was licensed, (2) a basis for deciding how these differ-ences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. To document the results of the SEP review for Ginna, the staff has issued NUREG-0821. NUREG-0821 was initially published in draft format in May 1982 and was issued in final form after. Commission review in December 1982. Some followup requirements for additional analysis by the licensee that may result in the need for facility modification or other correc-tive action were identified in the Final IPSAR. These requirements have been reviewed as operating. reactor licensing actions and are addressed in Supple-ment 1 to the IPSAR.

The major portion of the technical input supporting the staff SER has been provided by the IPSAR and SEP topic evaluations. The remainder of this SER will address other operating license issues not covered under the SEP. The SER includes consideration of major plant modifications that have occurred since the POL was issued, major substantive regulations adopted since the POL was issued, requirements stemming from the accident at Three Mile Island Unit 2 (TMI-2), and unresolved safety issues (USIs). USIs are issues consideredon a generic basis after the staff has made the initial determination that the safety significance of the issue does not"prohibit continued operation or required licensing actions while the longer term generic review is under way.

The format of the SER follows the general, format of SERs currently issued for new operating, licenses, but for many of the major headings, particularly those covered in the SEP, the SER briefly summarizes the findings of the Final IPSAR, and its supplements, or the SEP topic SERs. Similarly, when SERs have been issued on other topics, such as compliance with Appendix I, this SER briefly summarizes the previous SER and assesses whether the earlier findings are still val id.

Ginna SER 1-2

Appendix A contains a list of references cited in this report." For TMI Action Plan items, Appendix B identifies the status and plant-specific implementation of each TMI Action Plan item. For USIs, Appendix C not only discusses the status of the USIs but also satisfies the guidelines provided by the Atomic Safety and Licensing Appeal Board in the River Bend case (ALAB-444, 6 NRC 760 (1977)).

Because of the number. and volume of the SEP topic SERs, the topic SERs were not printed as an appendix to the IPSAR. Rather, they were identified in Appendix E of the IPSAR and Appendix A of Supplement 1. Sets of all topic SERs were forwarded to and are available for public inspection at the NRC Public Document Room, 1717 H St., N.W., Washington, D. C. 20555 and the Rochester Public Library, 115 South Avenue, Rochester, New York 14604. Copies of all topic SERs were provided to the licensee and plant service list at the time they were issued. The IPSAR and complete sets of the SEP topic SERs will accompany this SER when it is issued to the Atomic Safety and Licensing Board and other involved parties, as part of their consideration of the full-term license conversion proceeding.

The staff plans to issue a supplement to this SER when the Advisory Committee on Reactor Safeguards'eport to the Commission is available, as discussed in Section 18. The supplement will append a copy of the Committee's report, will address comments made by the Committee, and will describe steps taken by the NRC staff to resolve any issues raised as a result of the Committee's review.

There are a number of ongoing licensing actions for Ginna that are currently under staff review as noted in this SER. The staff has determined that these items do not require resolution before the issuance of an FTOL and should not delay the POL to FTOL conversion process. All of these items will be addressed as routine operating reactor licensing actions after the FTOL is issued.

In accordance with the provisions of the National Environmental Policy Act (NEPA) of 1969, the staff prepared the Draft and Final Environmental Statements that set forth the considerations related to the proposed POL to FTOL conver-sion. The Draft Environmental Statement was issued in April 1973, and the Final Environmental Statement (FES) in December 1973. Because the FES was issued a number of years ago, the staff performed an Environmental Evaluation to determine if an FES supplement was necessary. The Environmental Evaluation issued on June 17, 1983 (letter from F. Miraglia, NRC), concluded that an FES supplement is not necessary.

The NRC Project Manager assigned to the FTOL review for Ginna is Mr. George Dick. Mr. Dick may be contacted by calling (301) 492-7215 or writing:

Mr. George Dick U.S. Nuclear Regulatory Commission Division of Licensing Washington, D.C. 20555 "Avai'lability of all material cited is given on the inside front cover of this report.

Ginna SER 1-3

1. 2 Descri ti on of Plant The Ginna plant, located in Wayne County, near Rochester, New York, is a pressurized-water reactor designed by Westinghouse. The licensee is Rochester Gas and Electric 'Corporation (RG8E). RG8E filed the application for a con-struction permit and operating license in October 1965 (letter'ated November 1, 1965). The construction permit was issued by letter dated April 25, 1966. The initial submittal of the Final Safety Analysis Report was filed on, January 18,.

1968 (letter dated January 24, 1968) and the 'initial provisional operating license was issued on September 19, 1969 (letter dated September 25, 1969). By letter dated August 15, 1972, the licensee applied for a full-term operating license. The licensed thermal-power rating currently is 1,520, megawatts thermal (MWt). The Ginna primary coolant system configuration consists'of two hot, legs, two U-tube generators, 'a pressurizer, and two cold legs with a reactor coolant pump in each cold leg. The primary coolant system schematic is shown in Figure 1. 1.

The secondary system consists basically of the 'tu'rbine/generator, the condenser, and the feedwater system. Saturated steam is supplied to the turbine from the steam generator headers, where the steam expands through the high-pressure tur-bine and then flows through reheaters and intercept valves to two double-flow, low-pressure turbines, all in tandem. Five stages of extraction are provided from the low-pressure turbine for the feedwater heaters. The feedwater heaters for the lowest three stages are located in the condenser neck. The secondary system schematic also is shown in Figure 1. 1.

Condensate is taken from the condenser hotwell through the condensate booster pumps and full-flow demineralizers to the suction of the condensate pumps, through the hydrogen coolers, air ejectors, gland steam condenser, and low-pressure heaters to the suction of the feedwater pumps. The feedwater pumps then send feedwater through the high-pressure heaters to the steam generators.

Each main steam line has four American Society of Mechanical Engineers (ASME)

Code-approved safety valves that provide pressure relief for the steam genera-tors. There is also one power-operated relief valve on each line for long-term plant cooldown by atmospheric steam discharge if the condenser steam dump is not available. Each steam line is equipped with a fast-closing isolation valve and a nonreturn check valve. The isolation and nonreturn valves are located outside containment. Each feedwater line also is equipped with a nonreturn check and an isolation valve.

The reactor containment structure is a reinforced concrete, vertical right cylinder with a flat base and a hemispherical dome. A welded steel liner is attached to the inside face of the concrete shell to ensure a high degree of leaktightness. The thickness of the liner in the cylinder and dome is 3/8 in.

and in the base it is 1/4 in. The cylindrical reinforced concrete walls are 3 ft, 6 in. thick, and the concrete hemispherical dome is 2 ft, 6 in. thick.

The concrete base slab is 2 ft thick with an additional 2-ft-thick concrete fill over the bottom liner plate. The containment structure is 99 ft high to the spring line of the dome and has an inside diameter of 105 ft. The contain-ment vessel provides a minimum free volume of approximately 972,000 fthm. Access is provided during operation by means of two personnel air locks designed with an interlocked single-door-opening feature that is leak testable at containment Ginna SER

design pressure between doors. The open and closed status of each door is indicated in the control room.

1.3 Summar of 0 eratin Histor and Ex erience The Ginna plant received a provisional operating license in September 19, 1969, and began commercial operation in July 1970. In March 1972, the licensee increased power from 1,300 MMt to 1,520 MWt. The licensee has made some 'major modifications to the Ginna plant. In 1974, the revetment was upgraded. In 1975, in response to concerns raised on pipe break outside containment, a stand-by auxiliary feed system and housing structure was constructed. In 1974, to improve steam generator reliability, the licensee started all-volatile treatment for secondary coolant. Section 1.4 of NUREG-0821 gives a more detailed summary of the operating history and experience at Ginna.

Ginna SER 1-5

gl Wo Qr Hi Strength Concrete Dome and Walls 4 Carbon Steel Rate Liner To PRT ATMOS ATMOS PORV PORV PORV PORV Safety Safety Manual Manual Va ves Block Block Block Valve Valve Valve MSIV Turbine. Generator Set MSIV Spray vsrvss Control Rod Driv Mechani snl s Lv',

Reactor Standby Aux. Vessel Standby Auxiliary Feed Pump Feed Pump Coogng Water Low Head I .v Intake. Tunnel Safety Injection Stop . Governor Circulating Steam Steam Letdown Valves Valves Water Pumps Generator Generator I Motor Driven Cyt High Head High Head Aux. Feed Pump Safety Injection Safety Injection Reactor Reactor Discharge Canal Main To Lake Motor Driven 5 Coolant Corriant Pump Pump s~ Feedwater Aux. Feed Pump Reactor Core Reactor Turbine Drive """ Coo4nt Water C3 Aux. Feed Pump Concrete Pad Main st Bedrock Feedwater P D ECCS (Emergency Core Cooling System) Chemical and Volume 3 High Pressure Safety Injection Pumps Control System 2 Low Pressure Safety fnjection Pumps Condenser 2 Passive Tanks for Low Pressure Injection RB Coobng Water

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Figure l. 1 Schematic diagram of Westinghouse-designed pressurized-water reactor of Rochester Gas and Electric Corporation's Ginna Nuclear 'Power Plant

2 SITE CHARACTERISTICS 2.1 Geo ra h and Demo ra h The staff reviewed RG8E's'xclusion area authority and control and the popula-tion distribution under SEP Topics II-1.A and II-l.B (NUREG-0821). The staff concluded that the site exclusion area is completely within site boundaries.

No public highways or railroads traverse the exclusion area. The exclusion area does not extend over the waters of Lake Ontario. However, the staff has concluded that the documented agreements RG8E has with the U.S. Coast Guard for the control of water traffic in the event of a plant emergency meet the intent of the criteria in 10 CFR 100 and are acceptable. RG8E has committed to adding a map of the new exclusion area boundary to the Technical Specifications.

As discussed under SEP Topic II-1. B, the land surrounding the site is primarily of an agrarian nature 'and sparsely populated. There are no substantial popula-tion centers, industrial complexes, transportation arterials, or parks or other recreational facilities within a 3-mi radius of the Ginna site. The city of Rochester is .the largest population center within a 50-mi radius of the site (241,539 people, with 701,745 residing in the metropolitan area). The nearest community with a population of 1,000 or more is the town of Ontario, with its center located about 3-1/2 mi from the site. The staff concluded that the site conforms to current licensing criteria for population distribution.

2.2 Nearb Industrial Trans ortation and Hilitar Facilities The staff reviewed the potential hazards to safety-related structures, systems, and components resulting from nearby transportation, institutional, industrial, and military facilities under SEP Topic II-1. C.

There is little industrial activity in the vicinity of the Ginna plant. Wayne County, where Ginna is located, is primarily a rural area. Industrial activity is most heavily concentrated in the town of Webster, about 6 mi from the site, and consists primarily of light manufacturing (Xerox copiers). No industrial development is expected to'occur in the vicinity of the Ginna site.

The nearest transportation routes to the plant are Lake Road and U.S. Route 104, which pass about 1,700 ft and 3-1/2 mi, respectively, from the plant at their closest points of approach. The highway separation distances at Ginna exceed the minimum distance criteria given in Regulatory Guide 1.91, Revision 1, and, therefore, provide reasonable assurance that transportation accidents resulting in explosions of truck-size shipments of hazardous materials will not have an adverse effect on the safe operation of the plant. Any large quantities of, hazardous material would be shipped via U.S. Route 104 which is sufficiently distant (3-1/2 mi from the plant site) not to be of concern. Highway accidents on Lake Road involving certain hazardous chemicals theoretically could result in exCeeding toxicity limits in the plant control room assuming the most con-servative set of spill parameters and atmospheric dispersion conditions.

However, because of the highway separation distances and because shipments of Ginna SER 2-1

hazardous chemicals are infrequent past the plant (shipment would normally be along U.S. Route 104), there is reasonable assurance that the likelihood of a hazardous chemical spill affecting the operation of the plant is low. This matter has been evaluated separately under NUREG-0737, Item III.0.3.,4, "Control Room Habitability" (letter dated April ll, 1983).

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The railroad nearest to the plant is the Ontario Midland Railroad about 3-1/2 mi to the south. Comparing this distance'ith the guidance provided in Regulatory Guide 1.91, it is apparent that potential railroad accidents involv-ing haza'rdous materials are not considered to be a credible risk to the safe operation of the plant.

The nearest large pipelines to the,.plant are a 12-in. gas line located about 6 mi southwest of the plant and a 16-in gas line located about 10,mi south of the plant. These pipelines are far enough away to ensure that pipeline acci-dents will riot affect the safety of the nuclear plant. The gas line service to the Ginna house heating boiler and the boiler controls were reviewed and compared with National Fire Protection Association (NFPA) 85, as required in the staff's Fire Protection SER, forwarded by letter dated February 14, 1979 (Item 3. 1.46), and were found acceptable in Supplement 2 to the Fire Protection SER, issued by letter dated February 6, 1981. On the basis of the resolution of all gas line items during the Fire Protection review, the staff concludes that the existence of the gas line on the plant site does not present a safety hazard.

There are no large commerical harbors along the southern shore of Lake Ontario near the plant. Some freight is shipped through Rochester harbor about 20 mi to the west. Major shipping lanes in the lake are located, well off shore, at least 23 mi or more from the plant. The possibility of damage to the service water intake structure was also considered. As discussed in SEP Topic II-1.C, shipping on Lake Ontario is not considered to be a hazard to the plant.

The closest airport to the plant is the Williamson Flying Club Airport, a small, privately owne'd, general aviation facility located approximately 10 mi east-southeast of the plant.

The small number of operations at this airport is substantially less than the criteria in Section III.3 of Section 3.5. 1.5 of the Standard Review Plan (SRP)

(NUREG-0800) and, therefore, is .not considered a potential hazard.

Monroe County Airport, in Rochester, New York, about 25 mi southwest of the plant, is the nearest airport with scheduled commercial air service. The staff has reviewed the probabilities for an airline crash from the low-altitude Federal airways in the vicinity of Ginna. The calculated probabilities are 5. 1 x 10- for airway V2 and 1.4 x 10- for airway V2N. Because both probabilities are less than the 1 x 10-~ acceptance criteria, the staff concludes that the probability of a commercial air traffic crash at Ginna is acceptable.

Air Force Restricted Area 4-5203 is located about 8 mi north of the plant site, and slow-speed, low-altitude military training route (SR-826) passes about 6 mi'est of the plant. Acceptance Criterion II.2 of SRP Section 3.5. 1.6 states that, for military air space, a minimum distance of 5 mi is adequate for low-level training routes, except those associated'ith unusual activities, Ginna SER 2-2

such as practice bombing. Because no unusual activities such as practice bombing take place, the staff concludes that this criterion is met.

Extreme meteorological conditions and severe weather phenomena at the Ginna site were examined in SEP Topic II-2.A to determine if safety-related struc-tures, systems, and components were designed to function under all severe weather conditions. The licensee has consolidated the effects of, adverse weather into one overall structural review that is di,scussed in IPSAR Supple-ment 1. Therefore, the staff has found that the design to withstand severe weather is acceptable.

For SEP Topic II-2.C the staff performed a review to determine the appropriate onsite and near-site atmospheric transport and diffusion characteristics. In particular, the short-term relative ground-level air concentrations (X/Q) were determined for estimating offsite exposures resulting from postulated accidents.

The staff concluded that the X/Q values presented in .the SEP Topic II-2.C SER are appropriate for estimating exposures from postulated accidents and should be used in all accident calculations.

2.4 H drolo ic En ineerin Lake Ontario, on which the site is located, is about 190 mi long, 50 mi wide, and a maximum of 780 ft deep, and covers about 7,500 mP. The average lake level, based on over 100 years of record, is 246 ft mean sea level (MSL). The highest instantaneous still water lake level was 250. 2 ft NSL. Lake Ontario seldom freezes over, but ice occurs in'he winter, usually along the southern and northern shores and in the northeastern end of the lake.

The surface of the land on the southern shore of Lake Ontario, at the site and east and west of it, is either flat or gently rolling. It slopes upward to the south from an elevation of about 255 ft MSL near the .edge of the lake to 440 ft MSL at Ridge Road, 3.5 mi south of the lake.

There are no perennial streams on the site, but Deer Creek, an intermittent stream with a drainage area of about 13.3 mi~, enters the site from the west, passes south of the plant,'and empties into the lake near the northeastern corner of the site. Because no original Deer Creek design-crasis flooding elevation was identified for use in the SEP review of hydrologic hazards at the Ginna plant site, RG8E and the NRC performed a flood evaluation as part of the SEP.

The main plant ar'ea and buildings are at grade elevation 270.0 ft NSL; the north side of the turbine building and the screenhouse are at elevation 253.5 ft MSL. The plant grade entrances to the auxiliary building are at elevation 271 ft MSL. The lowest limiting elevation of safety-related equip-ment in the subbasement within the a'uxiliary building is 221.5 ft MSL.

The plant is protected from surges and wind-driven waves by a revetment with a top elevation of 261.0 ft NSL. All facilities necessary .to shut down and to maintain safe shutdown are flood protected to a maximum stillwater level of, 254.25 ft MSL. The screenhouse floor is at elevation 253.5 ft NSL, and the 0.75-ft curbs provide additional protection from potential exterior flooding.

Ginna SER 2-3

Additional flood protection is available in the screenhouse for the diesel gen-erator buses, which are set 16 in. above the floor, and the service water pump motors, which are .set 24 in. above the floor. The diesel generators, which are located in the north side of the turbine building, are flood protected by steel curbs projecting 18 in. above elevation 253.5 ft MSL. The switchgear is located at elevation 253 ft MSL and is protected by a 15-in. dam to elevation 254.25 ft MSL.

In the review of SEP Topics II-3.B, II-3.8. 1, and II-3.C, the staff determined that the current design-basis flood on Deer Creek (275 ft MSL) would inundate safety-related equipment in the auxiliary building, turbine building, and screenhouse.

Under Topic .II-3.B. 1, it was determined that procedures and facilities were not available to cope with a design-basis flood of Deer Creek.

Equipment that may be affected if a flood of Deer Creek is assumed includes service water pumps, emergency core cooling system (ECCS) pumps, charging pumps, and 480-V electrical buses 17 and 18. The ability to effect a'afe shutdown under such conditions would be seriously hampered.

The staff estimated a maximum flood level of 275.4 ft MSL (at the 271.0-ft MSL grade level) based on a probable maximum flood (PMF) flow of 38,700 cfs. This maximum flood flow would result in a"water depth of 'about 12.0 ft (elevation 265 ft MSL) in the vicinity of the screenhouse and lower level of the turbine building. Furthermore, it has been determined that flooding would not be a concern if the flow in Deer Creek did not exceed approximately 12,000 cfs. The discharge capability of Deer Creek was also evaluated against maximum rainfall and resulting runoff that has occurred historically in the region. Of the eight gaged basins reviewed, the largest recorded normalized peak discharge flow was 284 cfs/mi~, which is approximately one-third of the capacity of Deer Creek to convey water without overflowing the banks. Small-gaged drainage basins with relatively short records do not yield consistent results when subjected to frequency analyses. However, such analyses indicate recurrence intervals of several hundred of years for these historic floods. The staff concludes that the return period for a flood on Deer Creek using the largest normalized peak discharge flow recorded for eight regional basins would be of the same order of magnitude. This is not adequate for the staff to conclude that there is no potential for Deer Creek to flood.

Because of this flooding potential, a standard project flood was estimated for the Deer Creek Basin using standard project rainfall from the U.S. Army Corps of Engineers "Standard Project Flood Determination Procedure," EM 1110-2-1411, as revised March 1965. The standard 'project flood peak discharge was estimated to be approximately 15,000 cfs. The staff concluded that the licensee should at a minimum provide protection (e. g., watertight "doors and barriers) from flooding of De'er Creek to" the levels produced by the standard project flood plus 1 ft and justify on a cost-benefit basis why protection should not be pro-vided for higher levels associated with a probable maximum flood.

In a letter dated May 20, 1983, the licensee proposed to provide protection to an elevation of 273.8 ft MSL, which the licensee states is equivalent to a dis-charge of 26,000 cfs or 80K of his PMF."'his was determined to be a level for I

Ginna SER 2-4

which protection should reasonably "be provided because several other modifica-tions resulting from the structural upgrade program can be used to assist in attaining additional flood protection.

To accomplish this level of protection, the licensee has determined that, a num-ber of modifications will be required. As discussed in IPSAR Supplement 1, the staff finds that the proposed modifications are acceptable provided the licensee develops emergency procedures for the installation of flood protection devices.

2.5 Geolo and Seismolo The geology and seismology of the Ginna site were first reviewed in 1965 and 1966 by the U.S. Atomic Energy Commission (AEC) and its advisors, the U.S.

Geological Survey (USGS) and the Environmental Science Services Administration (ESSA) of the Coast and Geodetic Survey. In its report the USGS concluded that a relationship between seismicity and mapped faults had not been demonstrated within the area. The ESSA'concluded after its review of the regional seismicity that the plant should be designed for a moderately strong earthquake having an acceleration of approximately 0.20g without loss of function of components important to safety.

As a part of the SEP, a re-review of the seismological hazard at the Ginna site was conducted through the Site Specific Spectra Program for the SEP facilities in the eastern United States. The current recommendation for the seismic input ground motion was transmitted to the SEP owners in a letter from D. M. Crutchfield dated June 8, 1981.

The geology of the site was reassessed by the AEC staff when RG&E applied for a full-term license in August-1972. During the reassessment, RG&E reported the discovery of faults adjacent to Ginna during investigations for an alternate site for its Sterling Power Project. The staff reviewed the available data concerning these faults and concluded that they were not capable within the meaning of Appendix A, 10 CFR 100.

Other new information obtained since the construction permit review indicated the existence of relatively high residual stresses in bedrock in the Lake Ontario region (FitzPatrick Final Safety Analysis Report (Power Authority State of New York, 1972); Sbar and Sykes, 1973; and Dames and Moore, 1978).

The staff has reviewed the available data and concludes that if such stresses were present at the Ginna site, they were most likely relieved during excava-tion and construction and do not represent a problem to the plant. In response to staff questions, RG&E confirmed that during the life of the plant there have been no occurrences such as cracked walls or foundations that can be attributed to high stresses in bedrock.

During the SEP geological review of the 'Ginna site, the staff reviewed the fol-lowing materials: *th'e Ginna Safety Analysis Report (letter dated Junuary 24, 1968) and SER (letter dated June 24, 1969); 'Dames and Moore geological and geophysical investigations at the Ginna site, 1974; Dames and Moore geologic investigations at the Nine Mile Point Unit 2, 1978; aerial photographs; topo-graphic maps; and selected documents'from the open literature.

The following paragraphs present a brief description of the regional and site geology.

Ginna SER 2-5

The site is located on the southern shore of Lake Ontario in the eastern por-tion of the Erie-Ontario Lowlands Physiographic Province (Fenneman, 1938). The regional topography is of low relief and rises gradually from an elevation of

+250 MSL at the lake to +500 ft MSL at the Portage Escarpment, which is the northern boundary of the Appalachian Plateau Province to the south. A beach ridge 10 to 25 ft high parallels the shoreline of Lake Ontario 4 mi to the south'. North of the ridge is the lake plain of former glacial Lake Iroquois.

The site lies on this plain.

The southern margin of Lake Ontario is characterized by many promontories which seem to reflect prominent joint directions in bedrock. The site is located near one such promontory called Smokey Point. Major joint directions are N75'o 85 E and N10 E to 30 W. Erosional bluffs along the lake range from 15 to

,30 ft high. Smokey Point is located at the eastern end of a 5-mi-long ridge, the crest of which is about +310 ft. Relief in the site area is low, with ft elevations ranging from +350 to +300 ft. The site is underlain by 20 to 60 of, glacial deposits and approximately 2,700 ft of Paleozoic (570 million years before present (mybp) to 225 mybp) sedimentary rocks over crystalline basement.

The uppermost Paleozoic unit is sandstone of Upper Ordovician (455 to 430 mybp) gueenstone Formation.

The glacial deposits include at least two till horizons. The lower unit over-lies bedrock and varies in thickness from 6 to 25 ft. This unit consists of grayish red, calcareous, silty clay. The unit is poorly sorted and contains numerous striated and faceted pebbles, cobbles, and boulders.

unit is at or near the ground surface and ranges from 7 to 30 The upper ft till in thickness.

This unit is composed of relatively uniform olive gray to yellow brown silty, sandy clay, with large boulders several feet in diameter. Between the two ti 11 horizons is a zone of lakebed deposits consisting of gray, very plastic clay.

RG8E has determined by regional correlation that the lower till unit is associ-ated with the Woodfordian glacial advance, a substage of the Wisconsinan Stage, which took place about 22,000 years ago. The lakebed deposit is believed to have been deposited in the bed of Lake Iroquois. The upper till is related to a minor glacial readvancement that occurred about 12,000 years ago.. The staff has examined the evidence and agrees with the licensee's interpretation.

On the basis of the staff's review of SEP Topics II-4, II-4.A, II-4.B, and II-4.C, the staff concludes that the information. used for developing site-specific spectra is adequate and that the local geologic and,seismologic phenomena wi 11 not affect the plant.

I By letter dated November 18, 1982, the USGS clarified its previous recommenda-tions to the NRC regarding the reoccurrence of the 1886 Charleston-type earth-quake. The staff is studying this matter, and any requirements resulting from the study wi 11 be addressed as a separate licensing action. The staff also is performing a study of U.S. Geological Survey's Open File Report 82-1033 entitled "Probabilistic Estimates of Maximum Acceleration and Velocity in Rock in the Contiguous United States." The acceleration levels in this study are arrived at in a different (i. e., solely probabi listic) manner than those developed for individual nuclear power plants. Any changes in the staff's position will be reported separately.

Ginna SER

In the review of SEP Topic II-4.D, the staff performed an analysis for the two slopes on site. The staff determined that the slope west of the plant would not affect any safety-related structures. The staff also has determined that failure of the slope east of the plant may damage the all-volatile treatment but that backfitting was not recommended.

From the information provided by the licensee for SEP Topic II-4.F, the staff made the following findings:

The containment, 'the auxiliary, and the intermediate buildings are founded on the bedrock of the gueenstone Formation.

(2) The control and diesel generator buildings are founded on lean concrete placed over the bedrock.

(3) The buried service wate'r pipelines are founded on granular fill compacted deter-to at least 95%%u'f maximum density at optimum moisture content as mined in accordance with modified American Association of State Highway Officials procedure.

On the basis of the above considerations, the staff concluded that the settle-ment of foundations and buried equiment is not a safety concern at the Ginna plant.

Ginna SER 2-7

II t'I

3 DESIGN CRITERIA - STRUCTURES, 'COMPONENTS, EQUIPMENT, AND SYSTEMS

3. 1 Classification of Structures S stems and Com onents 10 CFR 50 (General Design Criterion (GDC) 1), as implemented by Regulatory Guide 1.26, requires that structures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards com-mensurate with the importance of safety functions to be performed. The codes used for the design, fabrication, erection, and testing of the Ginna plant were compared with current codes in the staff's review of SEP Topic III-1.

The development of the current edition of the ASME Boiler and Pressure Vessel Code has been a process evolving from earlier ASME Code, American National Standards Institute, and other standards, and manufacturer's requirements. In general, the materials of construction used in earlier designs are similar to those used in current designs. For most factors, older designs provide compa-rable levels of safety.

The initial review of this SEP topic identified several systems and components for which the licensee was unable to provide information to justify a conclusion that the quality standards imposed during plant construction meet quality stand-ards required for new facilities. The staff did not identify any inadequate ,

components with the exception of radiographic and fatigue analysis requirements for the regenerative heat exchanger and excess letdown heat exchanger. However, the licensee has stated that although the heat exchangers were considered Class C components in the Ginna Final Safety Analysis Report (FSAR), the equip-ment specifications for these items actually specified the regenerative heat exchanger and the tube side of the excess letdown heat exchange as Class A vessels. This information along with the information provided in response to the IPSAR has resolved all of the open issues from the original SEP review.

3.2 Wind and Tornado Loadin s 10 CFR 50 (GDC 2), as implemented by SRP Sections 3.3. 1 and 3.3.2 and Regula-tory Guides 1.76 and 1. 117, requires that the plant be designed to withstand the effects of natural phenomena such as wind and tornadoes.

The original design of the Ginna structural systems did not consider tornado effects. Therefore, as discussed in SEP Topic III-2, the existing design and construction of structures important to safety do not meet current licensing criteria regarding the ability of safety-related structures to resist tornado winds of 250 mph and differential pressures of 1.5 psi. In general, the limit-ing structural items identified by the staff are (1) unreinforced masonry walls, (2) certain steel members of the auxiliary building having long unbraced lengths, and (3) the siding system.

The siding system and masonry walls may. not meet the strength requirements for localized straight wind forces.' summary o'f the structures and their limiting elements is given in the IPSAR.

Ginna SER 3" 1

In addition to this topic, there are a number of SEP topics related to the structural design of the Ginna facility that required further analysis. RG8E has consolidated Topics II-2.A, "Severe Weather Phenomena," III-2, "Wind and Tornado Loadings," III-4.A, "Tornado Missiles," and III-7.B, "Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Design Criteria," into one overall structural review (letter dated April 29, 1982). This effort has been coordinated with other SEP topics such as III-6, "Seismic Design Consid-erations," III-5.A, "Effects'of Pipe Break on Structures, Systems, and Compo-nents Inside Containment," III-5.8, "Pipe Break Outside Containment," and IX-6, "Fire Protection."

The structural upgrade program is being performed in two phases. The purposes of the first phase are to define tornado windspeeds for which structures wi 11 be upgraded, demonstrate that there will exist a safe shutdown path completely protected from tornado missiles, and define general acceptance criteria for overall structural analyses. The second phase wi 11 consist of performing.

detailed design analyses to meet .the parameters accepted in the first phase and to identify necessary structural modifications. The results of the first phase were submitted by the licensee in letters dated April 22, 1983, April 28, 1983(a), May 19, 1983, and May 27, 1983. The staff's review of these submittals is described in an evaluation forwarded by letter dated August 22, 1983.

The evaluation concluded that the overall proposals by the licensee were acceptable.'he licensee proposed to upgrade the plant to withstand a 132-mph tornado to (1) achieve a safe hot shutdown with one completely protected train of equipment and the ability to proceed to cold shutdown; (2) protect the reactor coolant pressure boundary, the main steam and feedwater lines, the spent fuel assemblies; and (3) prevent accidents that wi 11 result in releases .

greater than the guidelines in 10 CFR 100.

The staff's evaluation further concluded that the methods proposed by the licensee to address the design code changes were adequate. The licensee demonstrated that adequate margins already exist at the facility, proposed modifications to achieve adequate margins (e. g., diesel generator building shear walls), or committed to review the concerns and modify as necessary during phase 2 of the structural upgrade program.

Although the staff found the proposed resolutions acceptable, it requires the licensee to do the following as described in its SER forwarded by letter dated August 22, 1983:

(1) Address items classified as Category 3 in the SER during phase 2 of the structural upgrade program.,

tl Institute

'N (2) procedures. requiring that cr'anes be unloaded during a tornado watch.

(3) Obtain staff approval of "secondary level" acceptance criteria once they have been determined in phase 2.

,(4) Pubmit an analysis to demonstrate the integrity of steam and feedwater lines given a failure of nearby block walls.

Ginna SER 3-2

(5) Submit, upon completion, the damage analysis study 'currently being con-ducted to determine the effects of a utility pole entering the spent fuel pool.

The acceptance criteria established for the licensee's two-phased structural upgrade program correspond to events with a probability, of 10-s per reactor year (i.e., 132-mph windspeed) and encompass issues raised for wind loads (Sec-tion 3. 2), tornado missiles (Section 3. 4. 1), the effects of block wall failure on the main steam and feedwater piping (Section 3.5), and'portions of the seis-mic review (Section 3.6). Moreover, there are margins beyond the proposed level of protection inherent in the analysis techriiques to be used to define needed plant modifications. On this basis, the staff has concluded that the structural upgrade program approach is acceptable, and additional protection, would not be cost effective.

3.3 Flood Level Desi n The general plant grade at Ginna is about 270 ft, with the exception of the area between Lake Ontario and the turbine building where. the grade level is at elevation 253 ft. Because the plant is protected from the lake by breakwater with a top elevation of 261 ft and because of the elevation of the plant, flooding was not considered a problem, and the plant structures were, therefore, not designed for the dynamic effects of the flooding. Moreover, the licensee stated that the probable maximum flood considered originally in the design of Ginna was based on 250.0 ft and later (1973) revised to a level of 253.3 ft.

This flood level was basically caused by Lake Ontario water. No other source of water was'onsidered- to produce water higher" than this design level (253.3 ft).

The staff's SER on SEP Topics II-3.A, II-3.B, and II-3.C indicates that the "probable maximum flood in Deer Creek could flood the site,to a depth of about 12.0 ft on the north side grade at 253.3 ft MSL and about 5.4 ft on'he. south side grade at 270 ft MSL. The licensee stated that the seismic Category I structures, systems, and equipment were not designed to withstand flooding (the licensee only postulated flooding by Lake Ontario and not by Deer Creek) and that this amount of flooding would be unacceptable from a systems'iewpoint.

As discussed in Section 2.4, the staff concluded that the data were not ade-quate to conclude that there is no potential for Deer Creek to flood; Because of this flooding potential, the staff estimated a standard project flood for the Deer Creek Basin. The staff concluded that the licensee, as a minimum, should provide protection (e.g., watertight doors and barriers) from the flood-ing of Deer Creek to the levels produced 'by the standard project flood plus 1 ft. Additionally, the licensee was to justify on a cost-benefit basis why protection should not be provided for higher levels associated with a probable maximum flood.

The licensee has proposed to provide flood protection to an elevat'ion of 273.8 ft MSL. The staff found this level of protection acceptable provided the licensee develops emergency procedures for the installation of'he flood protection devices.

Ginna SER 3-3

Effects of Groundwater On the basis of SEP Topics II-3.A, II-3.8, and II-3.C, the staff. concluded that the design-basis groundwater level to be used in determining the effects of high water on structures (SEP Topic III-3.A) should be ground level. The licensee assumed a groundwater level of 250 ft MSL, which is approximately 20'ft below grade at the upper portions of the plant. However, because of inadequate historical data, the staff has not been able to substantiate levels less than ground level. Rather than evaluate the effects of groundwater at grade level, RGKE has proposed to implement a groundwater monitoring program to verify the original design-basis groundwater level of 250 ft MSL. The staff finds this acceptable.

Inservice Ins ection of Mater Control Structures As discussed in SEP Topic III-3.C, RG&E identified water control structures and components that required surveillance in accordance with 10 CFR 50 (GDC 1) as implemented by Regulatory Guide 1. 127. The structures and components identified by the licensee were the intake structure and tunnel, discharge canal, trash rack and traveling screens of the service water system, and the revetment to prevent flooding from Lake Ontario. The staff concurred with the licensee's selection of the above structures and components, but indicated that the Deer Creek Basin should also be included in the inspection program and that the wooded area downstream of the Visitors Center should be cleaned out to initially establish adequate water conveyance.

The licensee has stated that the inspection of Deer Creek is not necessary because the. above factors would have no significant effect on the flood levels produced by the flooding used in the analysis. However, the licensee has agreed to establish the site flood protection level consistent with the present physical condition of Deer Creek. Any effects of shrubbery and natural debris, as well as of manmade obstructions such as culver ts and bridges, wi 11 be con-sidered in, determining the channel capacity at the different plant-stage levels. The staff, concurred with this approach in the IPSAR and backfitting was not recommended.

RG8E has implemented a revised inspection program that is essentially in con-formance with Regulatory Guide 1. 127. The staff, therefore, concludes that the surveillance of water control structures is acceptable.

3.4 Missile Protection

3. 4. 1 Tornado Missiles SEP Topic III-4.A evaluated Ginna's protection against objects and debris blown before tornado winds to determine (1) the integrity of the reactor coolant pressure boundary (2) the capability to shut the reactor down and maintain it in a safe shutdown condition (3) the capability to prevent accidents that could result in offsite exposures in excess of the dose guidelines of 10 CFR 100.

Ginna SER 3-4

The design-basis missiles used for this evaluation were a steel rod, 1 in. in diameter and 3 ft long, weighing 8 lb and traveling at 219 ft/sec. and a 35-ft-long wooden utility pole, weighing 1,490 lb and traveling at 146 ft/sec. The staff's evaluation indicated that major portions of the safety-related systems are inadequately protected from these'missiles.

The licensee has committed to evaluate and upgrade as necessary the following safety-related equipment as part of the overall structural upgr'ading described in Section 3.2 of this SER:

(1) refueling water storage tank (potential for flooding safety-related equip-ment and loss of primary water supply)

(2) electrical buses 17 and 18 (3) service water system (4) diesel generators and their fuel supply (5) relay room (6) main steam and feedwater piping between isolation valves and the contain-ment penetrations of the spent fuel pool (the top surface of the spent fuel pool (7)

'sinternals open, and, therefore, the internals are exposed)

The staff finds this to be acceptable.

3.4.2 Turbine Missiles In SEP Topic III-4.B and the generic review of turbine disc cracks, the staff evaluated all safety-related systems, structures, and components for either adequate protection from turbine missiles by means of structural barriers or acceptably low probability of damage from turbine missiles.

The findings of the generic review completed in August 1981 (letter dated August 28, 1981) concluded that an inspection schedule based on an approach developed by Westinghouse for their turbine provides an acceptably high degree of assurance that discs will be inspected before cracks can grow to one-half the size that could cause disc failure at speeds up to design speed. In a letter dated September 16, 1981, RG8E committed to the Westinghouse inspection program, and, therefore, the staff concludes that the probability of damage from turbine missiles at Ginna is acceptably low.

3. 4. 3 Internally Generated Missiles Missiles that are generated internally to the reactor facility (inside or out-side containment) may cause damage to structures, systems, and components that are necessary for the safe shutdown of the reactor facility or accident mitiga-tion to the structures, systems, and components whose failure could result in a significant release of radioactivity. The potential sources of such missiles are valve bonnets; hardware retaining bolts; relief valve parts; instrument Ginna SER 3-5

wells; pressure-containing equipment, such as accumulators and high-pressure bottles; high-speed rotating machinery; and rotating segments (e.g., impellers and fan blades). Under SEP Topic III-4.C, the staff reviewed the systems and components needed to perform safety functions and concluded that the design of protection from internally generated missiles met the intent of the criteria.

3. 4. 4 Site Proximity Missiles The potential for hazardous activities in the vicinity of the Ginna plant has been addressed in SEP Topic II-1.C, "Potential Hazards Due to Industrial, Trans-portation, Institutional, and Military Facilities. " As indicated therein, there is little industrial activity near the plant. The distances to the nearest land transportation routes are such (about 1,700 ft to the nearest highway and 3-1/2 mi to the nearest railroad) that the risks associated with potential missiles from transportation accidents on these routes are within the SRP Sec-tion 2. 2. 3 guidelines. Similarly, the nearest large gas pipelines are about 6 mi from the plant and do not pose a missile threat to the plant. Major Lake Ontario shipping routes are also sufficiently far away (about 23 mi) that they do not present a credible missile hazard from lake traffic. There are no mi li-tary facilities or activities near the plant that would create a missile hazard.

The review of SEP Topic II-1.C also evaluated the potential for aircraft becom-ing a missile hazard, both in connection with the operation of the Williamson Flying Club Airport, which is about 10 mi east>southeast of the plant, and the commercial air traffic in and out of Rochester via Federal airways V2N and V2, which are 2-1/2 and 10 mi from the plant site.

As evaluated in Topic II-1.C, it was determined that, because the Williamson Flying Club Airport expected a maximum of only 5,000 operations per year, and is about 10 mi from the site, the criteria of SRP Section 3.5.1.6 were met and there is no need to determine the probability of an airplane crash into the plant. Further, the hazard to the plant from commercial aircraft use of air-ways V2 and V2N was shown to be only 5. 1 x 10-8 and 1.4 x 10-8 per year, respectively. No danger to the plant from commercial airline traffic is thus expected.

3.5 Effects of Pi e Break on Structures S stems and Com onents The staff has evaluated the effects of pipe breaks to ensure that they would not cause the loss of needed functions of safety-related structures, systems, and components and to ensure that the plant can be safely shut down in the event of such breaks. Pipe breaks inside containment were evaluated under SEP Topic III-5.A, and pipe breaks outside containment were evaluated under SEP Topic III-5.B.

The staff reviewed the layout drawings, analyses, and other information pro-vided by the licensee. In addition, the staff toured representative locations in the Ginna containment on June 1-2, 1981, to observe the pipe configurations and proximity to safety-related equipment. As a result of this review, the licensee has performed additional analyses and has proposed rerouting of cer-tain instrumentation circuits away from high-energy lines. A fracture mechan-ics analysis has shown that the present leakage detection systems are adequate to detect primary system leaks before any crack grows to a substantial size.

Ginna SER 3-6

Therefore, the staff concludes that the licensee. has satisfactorily addressed the pipe whip and jet impingement effects of high-energy-line breaks inside containment and has demonstrated an adequate level of protection.

During its review of pipe breaks outside containment, the staff determined that (1) high- and moderate-energy-line breaks in the screenhouse could damage the power supplies to all service water pumps, (2) main steam line and feedline breaks, in the.. turbine and intermediate buildings could damage various safety-related equipment, and (3) moderate-energy-line. breaks in the, mechanical equip-ment room could result in the flooding of the battery rooms. The licensee re-sponded to these concerns in a letter dated March 22, 1983, and the resulting staff evaluation was issued on April 12, 1983. The resolution of, these concerns is described below.

The licensee has described methods that can be used in attaining and maintain-ing safe shutdown following the loss of all service water pumps. The staff concluded that such methods are acceptab)e provided that operating procedures for use of the alter nate sources of auxiliary feedwater are developed. The licensee has committed to develop such procedures in conjunction with imple-mentation of operating procedures for the alter nate diesel cooling system.

The licensee has already provided separate hose connections from the firewater system to provide an alternate source of cooling water for the diesel generator and the standby auxiliary feedwater system. The licensee has also replaced the doorway between the mechanical equipment room and the battery rooms with a watertight wall and to prevent accumulation of water in the mechanical equip-ment room in the event of a service water line break, a water relief valve has been provided between the mechanical equipment room and the turbine building.

The staff is concerned that postulated main steam and feedwater line breaks in the turbine building may adversely affect the atmospheric dump valves and main steam safety valves caused by fai lure of the concrete masonry block walls from overpressurization. In addition, postulated breaks in the intermediate building could result in jet impingement on some of these valves. During the evaluation of SEP Topic III-5. B, the licensee committed to install jet impingement shield-ing for one steam generator atmospheric dump valve and all main steam safety valves and to evaluate the effects of block wall failure. This was noted in Section 3.3. 1 of the IPSAR.

In letters dated June 16, 1983(a) and July 20, 1983, the licensee withdrew his commitment to provide such shielding on the basis of additional analyses of the consequences of jet impingement and block wall failure on the main steam and feedwater lines. The staff review of the licensee's analyses is presented in an evaluation forwarded by letter dated August 16, 1983. The staff concluded that the consequences of valve fai lures resulting from block wall collapse or jet impingement would be within the plant's capability to shut down safely and, therefore, the costs of plant modifications are not warranted.

As delineated in the evaluation of SEP Topic III-5.B, the staff is concerned that postulated breaks in steam heating lines in the auxiliary building could result in a steam environment in the building. The topic evaluation concluded that the equipment qualification program would address this issue; however, the licensee subsequently proposed to provide methods to attain safe shutdown even Ginna SER 3-7

if the auxiliary building experienced a steam environment from such a postulated break. The staff has reviewed the licensee's proposed methods and modifications and finds them acceptable. The staff will, however, require that the licensee implement the associated procedures and administrative controls, and the licensee has agreed.

The 'licensee has also committed to install pipe whip and jet impingement pro-tection for steam heating line risers on the i ntermediate floor to eliminate the possibility of postulated breaks in this location affecting nearby cable trays. The licensee has'committed to seal off the charging pump rooms from the rest of the auxiliary building to protect the pumps from experiencing a steam environment given a postulated steam line break. The staff finds these resolu-tions and commitments acceptable.

The topic 'evaluation concluded that the effects of pipe break in the relay room and air handling room were acceptable because adequate methods for detection were available. Subsequently, the licensee has decided to eliminate high-energy steam heating lines in these rooms to maintain a mild environment for equipment qualification purposes. The staff finds the licensee's proposal acceptable.

3.6 Seismic Desi n Considerations On the basis of the SEP Top'ic III-6 review of the original design analyses, the results of confirmatory analyses performed by the staff and its consultants, and the licensee's response to the SEP seismic-related safety issues, the fol-lowing conclusions can be drawn:

(1) Structure All safety-related structures and structural elements of the Ginna facility are adequately designed to resist the postulated seismic event. However, four sets of'teel bracing system were found to exceed the allowable stress level for the postulated safe shutdown earthquake (SSE). The licensee has committed to up-grade these structural members as part of any resultant modifications from the wind and tornado loading analysis. This analysis is currently being conducted by the licensee through his structural upgrade program (letter dated May 19, 1983). In a meeting on February 24, 1983, the analysis approach was described .

by the licensee. The staff concludes that the analysis techniques and accept-ance criteria proposed by the licensee are acceptable, as described in a sepa-rate eva'luation f'r the structural upgrade program '(Section 4.8), provided that the issues identified in IPSAR-Supplement 1 are confirmed in the application.

As a result of SEP seismic review, NRC Office of Inspection and Enforcement (IE) Bulletins 79-02'and 79-14, "and other NRC'requirements, the licensee imple-mented a seismic upgrade program.of all safety-related piping 2 in. in diameter and 1'arger using current NRC licensing criteria. Upgrading involves both modi-fication to and addition of pipe supports. The overall upgrade program is scheduled for completion by the end of the 1984 -refueling outage.

Ginna SER 3-8

According to the results of the SEP Topic III-6 piping audit analysis performed for the sampled piping systems, the piping systems will be capable of withstand-ing the postulated SSE upon completion of the seismic upgrade program.

(3) Mechanical E ui ment A total of 12 mechanical equipment items were sampled. Of the 12 items, 9 have been determined to be adequate and 3 were determined to be inadequate. The four essential service water pumps were determined to be not qualified with to structural and functional integrity because of the lack of 'egard the suction of the pumps that resulted in overstress in the pump casing support'ear, support. These pumps are unique to the service. water system. The licensee'has committed to upgrade the pumps, and the modifications are scheduled for comple-tion by June 30, 1984.

The staff's evaluation of three sampled safety-related tanks (component cooling surge tank, boric acid storage tank, and refueling water storage tank) showed that the support of the component cooling surge tank needs to be upgraded and the refueling water storage tank requires upgrading both with'regard to support and structural integrity. Because two of the sampled tanks were found to re-quire upgrading, the seismic review of other safety-related tanks was performed by the licensee to'emonstrate the design adequacy of the tanks.

On the basis of the staff's review, all safety-related field-erected tanks were found to be capable of withstanding the postulated seismic loadings except the three tanks discussed below:

(a) Refuelin Water Stora e Tank The licensee has committed to perform a detailed evaluation for demonstrating the design adequacy of this tank. The evaluation is scheduled to be completed by September 30, 1983,'nd any necessary modifications will be scheduled at that time.

(b) Sodium H droxide Tank According to the analysis performed,by the licensee (letter dated April 1l, 1983(b)), this tank would not meet the acceptance criteria for the seismic analysis. The licensee has committed to upgrade, this tank by the 1984 refueling outage.

(c) Volume Control Tank As discussed in the licensee's submittal (letter dated April ll, 1983(b)), this tank does not meet the acceptance criteria for the seismic review. However, the licensee does not consider that any modifications are necessary because (i) the predicted radiological consequences resulting from failure of this tank are substantially lower than that required by 10 CFR 100, (ii) the tank is not used for safe shutdown, and (iii) the tank volume is small (1,500 gal) and will not be of concern with respect to flooding of safety-related components. The staff agrees that this tank is not needed for safe shutdown and that the radiological consequences resulting from its Ginna SER 3-9

failure will not result in exposures substantially below the guide-lines set forth in 10 CFR 100. The issue of flooding caused by

. failure of tanks was reviewed as part of the resolution to. IPSAR Section 4.25.3. Therefore, this issue is resolved.

(4) Control Room Electrical Panels To demonstrate the structural integrity of panels'(load path from an internally mounted element to anchorage and support systems), the licensee has committed to conduct a low-impedance test for the main control board and, using the dy-namic characteristics measured from the test, perform a seismic analysis to demonstrate its design adequacy. According to the licensee, the test has already been completed and the analysis is being performed by his consultant.

The final report will be completed by the end of 1983.

(5) Electrical Cable Tra s Recently, the SEP Owners Group completed its full-scale shake-table laboratory testing program-of electrical cable trays. A report summarizing the results and general findings from the test Has submitted for review in April 1983 (URS/

John AD Blume). A second report that documents a set of cable tray evaluation criteria and guidelines has been developed, from the test. This report was, sub-mitted for review in August 1983 (URS/John A. Blume). For the evaluation of cable trays at.Ginna, the licensee committed to provide a program plan based on the critetia to be proposed by the Owners Group by the end of 1983. The, staff finds this approach acceptable and wi 11 review the Owners Group guidelines.

3.7 Desi n of Seismic Cate or I Structures 3.7. 1 Design Codes In SEP Topic III-7.B, the staff compared structural design codes used in the design of seismic Category I structures at Ginna with present codes. This was done through generic code versus code comparison without investigating specifi-cally how the original code was applied to the Ginna design; however, after reviewing drawings of structures at Ginna, the staff concluded that certain portions of the codes were not applicable to Ginna because the types of struc-tures to which the codes are referring were nonexistent at Ginna. The staff compared the loads and loading combinations used in the design of Ginna as described in the Ginna FSAR with those required today.

Code, load, and load combination changes affecting specific types of structural elements have been identified where existing safety margins in structures are significantly reduced from those that would be required by current'versions of the applicable codes and standards. Nineteen specific areas of design code changes potentially appl'i'cable to the Ginna plant have been identified where the current code requires substantially greater safety margins than the earlier version of the code, or where no original code provision existed.

The staff concluded in the IPSAR that design code changes potentially appli-cable, to the Ginna plant, where the current code requires substantially greater safety margins than the earlier version of the code or where no original code provision existed, should be evaluated to ensure adequate margins of safety.

Ginna SER 3-10

The licensee committed, in the, integrated assessment, to review the NRC evalua-tion and address the staff concerns in a structural upgrade program for his facility (see Section 3.2). .This program also addresses staff concerns raised in SEP Topics III-2, "Mind and Tornado Loadings," and III-4.A, "Tornado Missiles."

The staff's evaluation of the first phase of the structural upgrade program concluded that the methods proposed by the 1'icensee to address the design code changes were adequate. The licensee demonstrated that adequate margins already exist at the,facili,ty, proposed modifications to achieve adequate margins (e.g.,

diesel generator building shear walls), or committed to review the concerns and modify as necessary during phase 2 of the structural upgrade program.

3.7.2 Containment Structure The containment structure is a vertical prestressed concrete cylinder with a reinforced concrete flat base and a hemispherical dome. A welded steel liner (3/8 in. in thickness for the dome and cylinder and 1/4 in. for the base) is attached to the inside face of the concrete containment structure. The prin-cipal dimensions include an inside diameter of 105 ft 0 in. and 'a height (from top of base to spring line) of 99 ft 0'n. The nominal thickness 'dimensions of the reinforced concrete are 3 ft 6 in. for the wall and 2 ft 6 in. for the dome.

The concrete base slab is 2 ft thick, with an additional 2 ft of lean concrete fill over the bottom liner plate.

The containment at Ginna incorporates unique design features. It relies on prestressed, grouted rock anchors at the base to resist pressure and seismic loads. The grouted rock anchors are attached to'ertical, ungrouted tendons, in the walls. The containment only contains vertical prestressing tendons located in the side walls: a total of 160 tendons. Because there is no cir-cumferential curvature in the prestressing tendons at Ginna, there will be no mechanism to cause radial tension in the concrete resulting from prestressing forces. Therefore, the containment at Ginna will not experience delamination which is cracks in planes parallel to inner and outer concrete surfaces result-ing from prestressing forces (SEP Topic III-7.C).

The review of SEP Topic III-7.A revealed that the licensee's tendon survei 1-lance program does not specifically use the methodology described in Regulatory Guide 1.35, Revision 2. After reviewing the licensee's program, the staff identified a number of modifications to the surveillance program. The licensee has agreed to implement the modifications subject to the staff's approval of the containment vessel tendon evaluation program.

The containment liner is insulated with a l-l/4-in.-thick layer of polyvinyl chloride closed-cell foam from the base slab to 15 ft above the spring line.

This insulation limits the temperature rise in the liner and inside surface of the concrete in the cylindrical position of the structure in the event of a loss-of-coolant accident. The liner in the dome, however, is not insulated.

This. thermal discontipuity will cause high thermal stresses in the liner and can result in the buckling and failure of the liner.

The staff concluded in the IPSAR that further evaluation of the containment liner plate is required to ensure that it will maintain its leaktight integrity Ginna SER 3-11 W

when subjected to postulated loadings. The licensee submitted his analysis of this issue in a letter dated April 28, 1983(b). The staff reviewed the li-censee's submittal and subsequently issued jts evaluation in a letter dated August 1, 1983. In that evaluation, the staff concluded that although stud "

fai lure may occur, there is reasonable assurance that any such failure will occur in the stud and that the liner will retain its leaktight integrity.

Under SEP Topic III-7.D, the staff compared Ginna's containment structural integrity test procedure and the assessments of measurements with current criteria. Although deviations were identified, it is the staff's judgment that they are not significant and will not affect the assessments that were made in the licensee's structural integrity test report (GAI Report ¹1720, October 3, 1969). Therefore, the staff concludes that the test procedure used is adequate and the test results provide a basis to ensure that the containment structure will safely perform its intended functions and will withstand the design pres-sure load of 60 psig.

3.8 Floodin of Safet -Related E ui ment As implemented by SRP Section 3.6. 1, GDC 4 requires, in part, that structures, systems, and components important to safety be appropriately protected against dynamic effects, such as pipe whip and discharging fluids that may result from equipment failures.

In Section 4. 25 of the IPSAR, the staff concluded that the licensee should evaluate the design of various tanks in the auxiliary building to ensure their integrity to preclude flooding safety-related equipment. By letters dated.

April 11, 1983(b), and June 16, 1983(b), the licensee provided the necessary evaluations.

In those submittals, the licensee committed to qualify the three chemical volume and control system (CVCS) holdup tanks or prevent their fai lure from affecting the required safe shutdown equipment. The licensee demonstrated that the failure of all the remaining nonqualified tanks would result in a water level that is lower than the minimum elevation of equipment required for safe plant shutdown.

The staff's evaluation dated July 8, 1983, concluded that the licensee's analysis and commitments adequately resolve the issue.

Ginna SER 3-12

4 REACTOR

4. 1 Fuel S stem Desi n The Ginna reactor consists of 121 assemblies, each having a 14 x 14 fuel rod array. Each assembly contains 179 fuel rods, 16 rod cluster control (RCC) guide tubes, and 1 instrumentation tube. The fuel rods consist of slightly enriched UO~ pellets inserted'nto Zircaloy tubes.

The Ginna reactor was originally fueled with Westinghouse fuel. Beginning with Cycle 8 (1978), Exxon fuel was used to refuel the reactor. The use of Exxon (XN) fuel was evaluated and found to be acceptable as documented in the staff's safety evalution in support of Ammendment 19 to Ginna's Provisional Operating License (POL), forwarded in a letter dated May 1, 1978. The RCC guide tubes and the instrumentation tube in batches XN-1, XN-2, and XN-3 are made of stain-less steel, type 304L. The RCC guide tubes and instrument tubes in batches XN-4, XN-5, and XN-6 are Zircaloy. Each Exxon, Nuclear Corporation (ENC) assem-bly contains nine Zircaloy spacers with Inconel springs; eight of the spacers are located within the active fuel region.

Of the 121 assemblies, 4 contain mixed-oxide (Pu0~ plus UO~) bearing fuel rods.

The mixed-oxide (MOX) assemblies consist of three enrichment zones of PuO~ using natural UO~ as the diluent. RGEE inserted the four MOX assemblies into the Ginna reactor in 1980 as the culmination of experiments carried out as part of a research, demonstration, and development program initiated by RG5E. The staff's evaluation is documented as part of the safety evaluation for Amend-ment 32 of Ginna's POL, forwarded in a letter dated April 15, 1980.

In June 1983, Ginna began operating in Cycle 13.

4.2 0 eration With Less Than All Loo s in Service In a safety evaluation supporting Amendment 10 .to the POL, forwarded by letter dated March 30, 1976, the staff approved the shutdown margin analysis and issued the proposed Technical Specifications allowing operation with 1'ess than all loops in service up to 8.5X (130 MWt) of full power provided a predeter-mined shutdown margin can be maintained.

Current staff criteria require that, unless an emergency core cooling system (ECCS) analysis is performed and approved, a reactor must be shut down within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one of the operative loops becomes inoperable and cannot be returned to operation within that time period. The Ginna Technical Specifications state that if the appropriate shutdown margin cannot be maintained, the reactor will be brought to a hot shutdown condition until the proper shutdown margin can be established. There is no time limit on how long the plant can remain at 8.5X power in the n-1 loop configuration. However, on the basis of the low power level, the staff believes that the amount of stored energy in the fuel and the decay heat generated after shutdown following a postulated loss-of-coolant accident (LOCA) will be sufficiently reduced to 'ensure that peak cladding Ginna SER

temperatures are less than those calculated in the ECCS performance analyses.

This is based on the fact that peak cladding temperatures are strongly affected by the stored energy of the fuel and the decay heat. If the power level is reduced to less than lOX of full power, the stored energy in the fuel is propor-tionally reduced resulting in peak cladding temperatures significantly below those calculated for the accidents at full power. Fuel burnup and the cladding gap affect the relationship between power level and stored energy in the fuel.

These, however, are secondary effects, so that reducing power level by a factor of 10 reduces the stored energy. by approximately 10 with a substantial decrease in calculated peak cladding temperature during 'the LOCA. The staff finds that although the restrictions at Ginna are different from those in its criteria, the difference is not significant. Therefore, the staff concludes. that conti-nued operation in this mode is acceptable.

4.3 Loose-Parts Monitorin The requirements of 10 CFR 50 (GDC 13), as implemented by Regulatory Guide 1. 133, Revision 1, and SRP Section 4.4, prescribe a loose-parts monitoring program for the primary system of light-water-cooled reactors. Ginna does not have a loose-parts monitoring program that meets the criteria of Regulatory Guide l. 133.

However, the licensee did install a metal impact monitoring system on the sec-ondary side of the steam generators following the steam generator tube rupture.

A description of that system is contained in NUREG-0916.

A loose-parts monitoring program could provide an early detection of loose parts in the primary system that could help prevent damage to the primary system.

Such damage relates primarily to (1) damage to fuel cladding resulting from reheating or mechanical penetration (2) jamming of control rods (3) possible degradation of the component that is the source of the loose part to where it cannot properly perform its safety-related function.

Backfitting of a loose-parts monitoring program is being considered in Revi-sion 1 to Regulatory Guide 1. 133. When a decision to implement the recommenda-tions of this revision is reached, then the need to implement a loose-parts monitoring program on operating reactors, including Ginna, will be addressed generically. For the following reasons the need to backfit immediately was not recommended in SEP Topic III-8.A.

(1) A summary of 31 representative loose-parts incidents at 32 reactors (from value-impact statement of Revision 1 to Regulatory Guide l. 133) indicates that structural damage had occurred as a result of loose parts in only 9 incidents. None of these incidents, however, caused a safety-relat'ed accident.

(2) Most loose parts can be detected during refueling inspections. Backfitting is; therefore, not recommended for Ginna.

Ginna SER

4. 4 Reactor Materials In SEP Topic III-8.C, the staff reviewed the.materials used in the constru'ction of the reactor internals. The staff found that the materials specified for the Ginna plant have been proven adequate according to current standards by exten-sive tests and satisfactory performance. In addition, the. staff reviewed the effects of using sensitized stainless steel in the internal structures and RG8E's inservice inspection program for the reactor internal structures.

P

.On the basis of the SEP review, the staff concluded that the integrity of the reactor internal structures at Ginna has not been degraded through the use of sensitized stainless steel. Furthermore, the staff concludes that the integrity of the internal structures will be ensured by the licensee's inservice inspec-tion program, which,js in accordance with the requirements of 10 CFR 50.55a(g).

4.5 Reactivit Control S stems Protection A ainst Sin le Failure I

In SEP Topic IV-2 the staff evaluated the Ginna reactivity control systems to ensure that the design basis is consistent with analyses performed 'to verify that the protection system meets GDC 25. GDC 25 requires that the reactor protection system be designed to ensure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as.,accidental withdrawal of control rods. Reactivity control systems need not be single failure proof. However, the protection system must be capable of ensuring that acceptable fuel design limits are not exceeded in the event of a single failure in the reactivity control systems. The review was limited to the identification and evaluation of 'inadvertent control rod withdrawals and malpositioning of control rods that may occur as a result of single failures in the electrical circuits of the reactivity control systems.

On the basis of the information provided by the licensee, the staff concluded

. that the fo) lowing may occur as a result of single failures:

(1) Two control rod banks may be simultaneously withdrawn.

(2) Two banks may overlap at other than the design value.

t' This conclusion is based on the availability of alarm and interlock circuits associated with the rod control system so, that certain consequential effects of single failures within the rod control system are precluded by the operability

,of these interlocks and alarms. -, The basis for the assumption that these alarms and interlocks will be operable is that a failure in the alarm and interlock circuits will be identified and corrected during routine maintenance or as a result of system fault investigation., The effects of single failures occurring after an undetected failure has occurred in the alarm and interlock system are not included in the evaluation. This is consistent with the basis. used for plants currently under operating-license review.

r Each of.,the two reactivity control system malfunctions has been addressed as part of SEP Topic XV-8, "Control Rod Misoperation," to verify that specified acceptable fuel design limits are not exceeded.

Fuel design limits are not exceeded for either of the above two malfunctions, and thus, GDC 25 is met insofar as electrical failures within reactivity control systems are concerned.

Ginna SER 4-3

4.6 0 eratin Problems This section discusses. the more significant problems that affected operation of the reactor.

4. 6. 1 Fuel Failures Shortly after the reactor reached its authorized power level of 1,300 MWt in March 1970, a gradual rise in primary coolant radioactivity concentration was observed. This indicated the development of defects or leaks in the cladding of the fuel rods. The activity continued to increase until July 1970 and then leveled off at a steady-state value, of about 60 pCi/cc (approximately 25K of the Technical Specification limit). Following a shutdown with a quick return to power, the'activity would peak as high as 125 pCi/cc before returning to its steady-state value. In March 1971, the fuel was inspected and 32 fuel assemblies were found to contain 1 or more leaking fuel rods. The failures were attributed to moisture contamination during the fuel manufacturing, which can result in the formation of zirconium hydride and thereby can cause fai lure of the rod cladding. Twelve of the assemblies were replaced with new assemblies and Cycle 1 was continued.

The primary coolant system radioactivity concentration continued to rise so that just before the power rating increase in March 1972, the steady-state radioactivity level was approximately 65%%uo'f the Technical Specification limit.

During the power escalation program, the steady-state activity level increased to 88K of the Technical Specification limit at 1,455 MWt and the power increase to 1,520 MWt was delayed until just befo're the refueling shutdown in April 1972.

~

During this refueling, most of the known leaking fuel rods were removed from the core. Cycle 2 was then operated with the primary coolant activity level ranging between 10K and 20K of the Technical Specification limit. Cycle 3 and Cycle 4 were operated with activity levels about 10K to 15K of the limit.

Therefore, it appears that the control of moisture contamination in'the reload assemblies had been successful in reducing the number of fuel rod cladding failures.

4. 6. 2 Fuel Densification, Cladding Collapse, and Fuel Rod Bowing During the spring 1972 refueling, the licensee found fuel assemblies with small sections of some rods collapsed and a few bowed rods. Sixty-one fuel assemblies were replaced, and Cycle 2 was authorized with interim conditions for operation that limited the power level to 1,266 MWt (83K of rated power,). The interim conditions for operation provided -limits that ensured that safety margins were not reduced. In October 1972, RGEE removed the remainder of the nonprepres-surized fuel from the reactor. Since November 1972, the plant has operated without cladding collapse. The remaining assemblies that were susceptible to rod bowing because of insufficient end clearance between the top nozzle and the top of the fuel rods were also removed in October 1972. Following the release of the U. S., Atomic Energy Commission's "Technical Report on Densification of Light Water Reactor Fuels" (letter dated November 20, 1972), RGLE submitted proposed changes to the Technical Specifications-that were in conformance with the requirements of that report and accounted for the effects of operation with densified fuel. By letter dated June 29, 1973', these proposed changes were approved and issued.~ Full rated power (1,520 MWt) was again authorized with the incorporation of these- changes.

Ginna SER

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5. 1 Summar Descri tion As shown in Figure 5. 1, the reactor coolant system (RCS) consists of two simi-

>lar heat transport loops connected to the reactor pressure vessel. Each loop contains a reactor coolant pump, steam generator, and associated piping. In addition, the system includes a pressurizer, a pressurizer relief tank, inter-connecting piping, and .instrumentation necessary for operational control. All of these components are located within the containment building.

During operation, the RCS transfers the heat generated in the core to the steam generators where steam is produced to drive the turbine generator. Borated de-mineralized water is circulated in the RCS at a flow rate and temperature con-sistent with achieving the reactor core thermal-hydraulic performance. The coolant also acts as a neutron moderator and reflector and as a solvent for the neutron-absorbing boric acid used for chemical shim control.

The RCS pressure boundary provides a second barrier against the release of radioactivity generated within the reactor and is designed to ensure a high degree of integrity throughout the life of the plant.

The RCS pressure changes during normal operation are controlled by the use of the pressurizer where water and steam are maintained in equilibrium by elec-trical heaters and water spray. Spring-loaded safety valves and power-operated relief valves are mounted on the pressurizer and discharge to the pressurizer relief tank where steam is condensed and cooled by mixing with water.

5.2 Inte rit of Reactor Coolant Pressure Boundar The staff compared the codes used for the design, fabrication, erection, and testing of the Ginna plant with current codes under SEP Topic III-1. The resolution of that topic is discussed in Section 3. 1 of this SER, with the conclusion that older designs provide comparable levels of safety.

I To ensure the integrity of the reactor coolant pressure boundary, RG&E has developed an inservice inspection program that meets the requirements of'0 CFR 50.55a(g). The program was reviewed by the staff and approved by letter dated May 17, 1977. In addition, RG&E's inservice valve testing program was approved by letter dated May 26, 1981. The staff is reviewing the licensee's latest revision of the inservice pump testing.

Under SEP Topic V-5, the staff reviewed the licensee's ability to detect reactor coolant pressure boundary leakage. The licensee has all three systems required by Regulatory Guide 1.45. Two of'he three systems meet the sensitivity re-quirements. The third system (sump "A" level monitoring) can measure approxi-a 2-gpm leak in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in comparison to the Regulatory Guide requirement 'ately of a 1-gpm leak in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In addition to the three leakage-detection systems recommended by the Regulatory Guide, Ginna also'incorporates six other diverse Ginna SER 5" 1

systems. Taking all of these systems into consideration, the staff believes that a 1-gpm leak from the reactor coolant pressure boundary to the containment can be detected within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, as required by the Regulatory Guide.

However, none of the three systems required are seismically qualified to func-tion following earthquakes up to the operating-basis-earthquake (OBE) level, as detailed in the acceptance criteria of SRP Section 5.2.5. Ginna does have, as one of its diverse systems, a sump "B" level monitoring system that is seismic Category I and can measure a 10.5-gpm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The staff believes that the containment sump "B" level monitoring system will adequately~

meet the leak-detection needs of Ginna following a seismic event, including <

the safe shutdown earthquake. On the basis of the above, backfitting was not recommended during the integrated assessment.

During the 1978 refueling outage, RG8E modified the actuation circuitry of the existing air-operated pressurizer relief valves to provide a low-pressure set-point at 435 psig during startup and shutdown conditions. The new low-pressure power-operated relief valve (PORV) actuation circuitry uses multiple pressure sensors, power supplies, and logic trains to improve system reliability. Each of the two PORVs is manually enabled using two keylock switches, one to line up

.the air supply and the other to enable the low-pressure setpoint. When the re-actor vessel is at low temperatures with the overpressure protection system (OPS) enabled, a pressure transient is terminated below the Appendix G, 10 CFR 50, limit by automatic opening of the PORVs. An enabling alarm monitors the RCS temperature, the position of the keylock switches (two per channel), and the upstream isolation valve position. The OPS is enabled at a temperature of 330 F during plant cooldown and is disabled at the same temperature during plant heat-up. The enabling alarm alerts the operator in the event the RCS temperature is below 330~F and OPS valve or switch alignment has not been completed. The staff review of the OPS and the Technical Specifications associated with it are documented in Amendment 26 to Ginna's POL issued, by letter dated April 18, 1979.

5.3 Reactor Vessel The reactor vessel is cylindrical in shape with a hemispherical bottom and a flanged and gasketed removal upper head. The vessel is designed in accordance with Section III (" Nuclear Vessels" ) of the ASME "Boiler and Pressure Vessel Code" (1965) and GDC 1, 2, 5, 9, 14, 16, 33, 34, 36, and 40. The materials of construction and the design data of the reactor vessel are given in Tables 5. 1 and 5.2, respectively.

A one-piece thermal shield, concentric with the reactor core, is located between the core barrel and the reactor vessel. The shield, which is cooled by the coolant on its downward pass, protects the vessel by attenuating much of the gamma radiation and some of the fast neutrons that escape from the core.

This shield minimizes thermal stresses in the vessel, which result from heat generated by the absorption of gamma energy.

Surveillance specimens made from reactor vessel. steel are located between the reactor vessel wall and the thermal shield. These specimens are examined intervals to evaluate changes in the nil ductility transition tempera- at'elected ture of reactor vessel material.

Ginna SER 5-2

In October 1979, the NRC issued NUREG-0569, "Evaluation of the Integrity of SEP Reactor Vessels." In this evaluation, the only item that was found that may affect the integrity of the Ginna reactor vessel for operation past tive full-power years (EFPY) is the vessel beltline welds that are predicted ll effec-to have low upper-shelf toughness. It was recommended that the NRC review the fracture toughnes's of the vessel beltline-materials at about 10 EFPY and deter-mine if these materials meet the requirements of Appendix G, 10 CFR 50 licensee submitted the material test results of the third capsule of the

'he reactor pressure vessel material surveillance program for Ginna in a letter dated December 8, 1982.' The staff has reviewed the test results and has deter-mined that the licensee's proposed pressure-temperature limits are acceptable until the intermediate-to-lower shell weld accumulates a neutron fluence of 1.5 x 10 n/cm (E > 1 MeV). The staff's review of the licensee's radiation analysis and'neutron dosimetry indicates that a number of items need to be addressed further by the licensee before the staff can determine the number of EFPYs that the proposed pressure-temperature limits may be used. RG8E has agreed to provide additional information.

The licensee has requested that the date for removal of the next reactor vessel surveillance capsule be revised to the refueling outage that corresponds to 17 EFPYs. According to Westinghouse Topical Report WCAP 10086, 17 EFPYs corre-spond to a capsule neutron fluence of 4. 10 x 10'. n/cm , which is the approxi-,

mate fluence at the inner surface location at the end of life of the reactor vessel. The removal of the next reactor vessel material surveillance capsule when its fluence reaches the value estimated for the inner surface location at the reactor'vessel end of life is considered acceptable by the staff.

5. 4 Com onent and Subs stem Desi n 5.4. 1 Reactor Coolant Pump's 5.4. 1. 1 Reactor Coolant Pump Flywheel Integrity GDC 4 requires, in part, that nuclear power plant structures, systems, and com-ponents important to safety be protected against the effects of missiles that might result from equipment failures. Because reactor coolant pump flywheels have large masses and rotate at speeds of approximately 1,200 rpm during normal operation, a loss of flywheel integrity could result in high-energy missiles and excessive vibration of the reactor coolant pump assemby. The safety con-sequences could be significant because of possible damage to the reactor cool-ant system, the containment, or the engineered safety features.

'fstaffstaff The the pump has reviewed the flywheels at material, fabrication, design, and inspection aspects Ginna for compliance with Regulatory Guide 1. 14. The concluded in SEP Topic III-10.8 that the requirements for fabrication and the margins against flaw-induced fracture and yielding, required by Regulatory Guide 1. 14, have been satisfied for the flywheels. The present inservice in-spection program for the Ginna pump flywheels does meet the requirements of Regulatory Guide 1. 14. Compliance with the guide provides a basis acceptable to the staff for satisfying, in part, the requirements of GDC 4, "Environmental and Missile Design Bases."

Ginna SER 5-3

Additionally, under SEP Topic Y-7, the staff reviewed the possibl,ity of a struc-tural failure of a reactor coolant pump resulting from overspeed. The probabil-ity of attaining an overspeed following a LOCA that is sufficient to cause loss of flywheel integrity is very remote. This probability would be the product of the conditional probabilities of a break of, a large primary system coolant pipe, the probability of failure of the pipe restraints so that the break could become a double-ended guillotine break (calculations show a significantly smaller overspeed for a realistically constrained guillotine break), and the probability of a loss of electric power to the pump so that there is no- electric braking effect and the pump is permitted to accelerate freely. Also, the pump would have to remain free spinning. Seizing of the shaft or motor components could prevent overspeed.

On the basis of the low probability of the sequence of events necessary and because of the licensee's performance of an inservice inspection program for the flywheel, continued operation of the Ginna reactor is acceptable.

\

5.4. 1.2 Reactor Coolant Pump Seals The Westinghouse Owners Group initiated a test program in June 1983 to deter-mine the survivability of the reactor coolant pump (RCP) seconda'ry seals under conditions wherein seal cooling is lost. The secondary seals were thought to be the weak components in the design of the RCP seal. Their failure under a loss-of-cooling condition could lead to a small-break LOCA resulting from failure of the complete RCP seal.

The secondary seal survivability test program has now progressed through test number four. These tests have not been successful in confirming the ability of the secondary seals, which consist of 0-ring and channel seals, to survive under loss-of-cooling conditions. Two of the four tests resulted in 0-ring blowout with complete loss of sealing ability; two tests, although maintaining sealing ability, resulted in moderate to severe damage to the seals.

Westinghouse believes that their 0-ring and channel seal test rig does not rep-resent the actual design or loss-of-cooling conditions in the RCP; therefore, Westinghouse believes that the results of these tests are inconclusive. Westing-house is now proposing to the Owners Group that changes be made to the test rig which will eliminate these differences. The staff currently does not have enough information to either agree or disagree with Westinghouse conclusions regarding the results of the first four tests.

Experience data from Westinghouse pumps show RCP seal survival under loss-of-cooling conditions for a period of up to 65 minutes with no abnormal leakage.

There are other instances of interruptions of cooling function for periods of 45, 30, and 10 or less minutes also with low leakage. In addition, Westinghouse has presented 0-ring experience data from valve tests at 550~F and 2,250 psig which show survival of ethylene propylene 0-rings up to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

Westinghouse has committed to supply the staff with additional information that will include a schedule for further testing and analysis. The staff is inde-pendently pursuing an analysis of the problem as well as further testing of the RCP seal failure mechanisms. The staff will meet with Westinghouse, and the Westinghouse Owners Group to discuss the RCP seal problem and decide on a Ginna SER

course of action for further testing and analysis. In addition, the staff has involved Region IV in this problem and they are scheduling audits of both the Westinghouse pump manufacturing facility and their seal vendor, Stein Seal Company. NRC staff members will, accompany Region IV personnel on these. audits.

In any event,.the results of any further Westinghouse Owners Group testing and analysis will most probably not be complete until the end of 1983. Determina-tion of any action necessary for Ginna will be made when the staff has had an opportunity to gain access to these results and other information being developed.

5. 4. 2 Steam Generators 5.4.2. 1 Steam Generator Design The Ginna plant is a two-loop pressurized-water-reactor system with one Westinghouse Model 44 (44,000 ft~) vertical U-tube steam generator per reactor coolant loop. The steam generators were fabricated by the Westinghouse Heat Transfer Division and are similar to those at H. B. Robinson, Point Beach Units 1 and 2, and Indian Point Unit 2,, and the original steam generators at Turkey Point Units 3 and 4.

Figure 5.2 shows a cutaway view of this type of steam generator. The pirimary channel head, the tubesheet, and the 3,260 U-tubes form the reactor coolant pressure boundary within the steam generator. Feedwater enters the secondary side of the, steam generator through an inlet nozzle and internal feed ring.

The preheated, water flows down the space formed, by the tube wrapper and the shell, (the downcomer annulus), and then flows between the end of the wrapper and the tubesheet into the tube bundle. The U-tubes provide the surface area needed to transfer heat from the primary system to the secondary system.

Moisture separators and dryers are located above the tube bundle. These compo-nents function, to improve steam quality, by removing entrained moisture from the steam'eaving the tube bundle. The moisture removed from the steam is recycled through the downcomer annulus for mixing with the .incoming feedwater.

C The tubes of each steam generator are 7/8 in. in diameter with a 0.050-in.-

thick wall, and are made of a nickel-chromium-iron alloy (Inconel 600). The tubes were rolled and welded to the tubesheet, but were .not expanded in the tubesheet beyond the length expanded in shop fabrication.

The tubesheet is a 22-in.-thick, low-alloy steel forging that is clad with a nickel-chromium-iron alloy- and has holes drilled to accept the U-tubes. The tubesheet is welded to the stub barrel section of the shell.

I The shell is made of low-alloy steel plate. It forms the, outer boundary of the downcomer annulus and provides support for the upper internals of the generator.

5 '.2.2 Secondary-Side Modifications In March 1975, modifications were made to the steam generators to increase lateral flow velocities across the tubesheet and to improve blowdown efficiency.

The purpose of the increased flow velocity and blowdown was to reduce the amount Ginna SER

of slu'dge that had accumulated on the tubesheet. The sludge provided the con-centrating medium where corrosion of the tubing had been observed.'hese modi-fications included removal of the downcomer flow resistance plates by oxyacety-lene cutting, modifications to the feed ring, replacement of the moisture sepa-rator'orifice rings, and installation of blowdown lane flow blockers.

\

Additional modifications to the moisture separators were made in February 1976 to reduce moisture carryover. In February 1979, the bottom drain holes on the feed ring were plug welded and J-tubes were installed on top of the feed ring to provide added assurance against waterhammer.

5.4.2.3 Steam Generator Operating Environment The Ginna steam generator began operating in 1970 with coordinated phosphate (NaP04) secondary-side water chemistry control to reduce corrosion and con-tinued on this regime unti 1 1974. During this period the industry, in general, experienced early difficulties with phosphate wastage corrosion and stress-corrosion cracking where sludge had accumulated on the tubesheet. These diffi--

culties were attributed to problems in adequately controlling phosphate concen-trations and to impurities carried into the steam generators by th'e feedwater.

Wastage corrosion was identified during inservice inspections performed at Ginna in spring 1974. Following a Westinghouse recommendation to its customers in a letter dated August 29, 1974, Ginna converted from phosphate to all-volatile treatment (AVT) secondary-side water chemistry control during a shutdown that began in November 1974. AVT consisted of the addition of hydrazine (N~H4) to the condensate water for the purpose of scavenging oxygen. At this time, blow-down rates were increased from the maximum of 16 gpm per steam generator to 64 gpm to reduce cation conductivity.

The licensee reports that he has closely monitored condenser tube integrity and performed preventive repairs to damaged condenser tubing during shutdowns to provide assurance of feedwater purity. In addition, full-flow polishing de-mineralizers were installed and have been in service since 1978. Blowdown rates were further increased to 70 gpm in February 1979.

The blowdown chemistry comparison shown in Table 5.3 demonstrates the continued improvement in bulk water chemistry over the last 8 years.

Since November 1969, oxygen concentration has generally been less than 5 parts per billion (ppb). Several times over the last 12 years, dissolved oxygen has been as high as 40 ppb for several days. From July 1978 to the present, feedwater-dissolved oxygen has been less than 1 ppb.

5.4.2.4 Steam Generator Surveillance and Repair Requirements Pressurized-water reactor (PWR) steam generators supplied by the various nu-clear steam supply system vendors have, over a number of years, experienced a variety of tube-degradation problems. The problems have included degradation caused by tube corrosion, vibration, mechanical wear or impact, and the phenom-enon known as "denting." These mechanisms have been discussed in detail in NUREG-0886. To provide assurance that PWR steam generators can be operated safely, the primary objective of the NRC has been that degraded steam generator Ginna SER 5-6

tubes retain adequate integrity against a gross tube failure or burst over the full range of normal and postulated accident conditions. To meet this objec-tive, steam generator tube surveillance requirements have been established.

These include'requirements for inservice inspection of steam generator tubes, acceptance criteria beyond which degraded steam generator tubes must be~removed from service by plugging or repaired by 'sleeving, and primary-to-secondary rate limits beyond which the plant must be shut down for appropriate 'eakage corrective action.

Appendix B of 'the Ginna Station'uality Assurance Manual presents the licensee's inservice inspection program. This document 'commits to the requirements of 10 CFR 50.55a and follows the guidance of Section XI of the ASME Code. The selection of steam generator tubes for examinations and the extent of these examinations are as descr ibed in Regulatory Guide 1.83, Revision 1, dated 1975.

Plants currently undergoing review for'n operating license are required by the staff to incorporate a steam generator inspection program into the plant Tech-nical Specifications which complies with the Standard Technical Specifications for Mestinghouse pressurized-water reactors (NUREG-0452) in addition to Regula-tory Guide 1.83. The Standard Technical Specifications contain more stringent tube sampling requirements (i.e., number of tubes to be, inspected) than Regula-tory Guide 1.83 for cases where initial sampling inspectioq results in the find-ing of a certain number of degraded or defective tubes; However, the sample sizes actually implemented for steam generator inspections at Ginna in recent years have generally met or exceeded the Standard Technical Specifications.

Steam generator inspection pr'ograms implemented at Ginna since April 1980 have included lOOX sampling of tubes on the hot-le'g side, and at least a 25K samp'le on the cold-leg side.

Eddy current testing (ECT) .is the principal method used for performing tube at Ginna. The staff's review of RG5E's inspection program is dis- 'nspections cussed in,NUREG-0916. The staff's conclusion is that the inspection techniques used at Ginna meet or exceed ASME Co'de requirements and conform to the state of the art. 1 5.4.2.5 Plugging Limits Appendix B of the Ginna Station guality Assurance Manual'stablishes the bases for interpretation of the eddy current inspection results and specifies a 40K limit on allowable percent through-wall penetration by tube flaws (such as cor-rosion or mechanical wear). The 40K plugging limit is intended to ensure that tubes accepted for continued service will retain adequate structural margins against a gross tube failure under normal operating and accident conditions.

Tubes exhibiting eddy current indications in excess of this limit must be plugged or repaired by sleeving. The plugging repair procedure involves the insertion of plugs at the inlet and outlet ends of the tube, rendering the tube inactive as a primary pressure or heat transfer boundary.

8 t By letters dated April 1, 1983, June 13, 1983, and July 25, 1983, sleeving was approved at Ginna as an alternative to plugging. A sleeve repair involves the insertion of smaller diameter tubes (sleeves) inside the parent tube so as to span the degraded portion of'the parent tube and then sealing th'e sleeve ends against the tube wall. Sleeves are designed to restore the original integrity of the degraded tube, while allowing the tube to remain functional.

Ginna SER 5-7

5.4.2.6., Steam Generator Tube Rupture On January 25, 1982, one of the 3,260 steam generator tubes in the B steam gen-erator ruptured. There was no indication of primary-to-secondary leakage before

,the rupture occurred. The plant transient resulting from the tube failure in-cluded a significant primary system depressurization, actuation of the safety injection system, and a minor release of radioactive materials from the plant.

An investigation concluded that the rupture occurred as a result of wall thin" ning which it was postulated resulted from contact wear with an adjacent, plugged, severed tube. An evaluation of the tube rupture event and the staff's subsequent SER are provided in NUREG-0909 and NUREG-0916.

5.4.3 Residual Heat Removal System The residual heat removal (RHR) system consists of a single drop line from the reactor coolant system (RCS) (hot 'leg) through two pumps and their associated abilityy.

heat exchangers and back to the RCS through a single header. Each pump can be manually cross-connected to the alternate heat exchanger for increased reli-Normal cooldown,of the RCS is accomplished by operating both pumps and-heat exchangers; however, a lesser cooldown rate can be achieved with only one pump. One heat exchanger can effect cooldown approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after shutdown. Each RHR pump is supplied power from separate redundant 480-V emer-gency buses. The system is normally operated from the control room.

The single RHR cooling suction line from the RCS and single discharge line to the RCS render the RHR system susceptible to single failure of the in-line suction valves (700 and 701) in the closed position and passive failures of either suction or discharge lines. (Valves 700 and 701, which are inside con-tainment, can be manually operated to overcome a motor operator or power supply failure.) Although these fai lures would render the RHR mode of decay heat, removal inoperable, the alternate means of decay heat removal using the steam generators is still available as a backup. For the case of a failure of valves 700 or 701 or a pipe break downstream of these valves, an alternate flow path for core cooling is available through the RHR cooling discharge line and the high-pressure safety injection (HPSI) pumps. There are other means of core decay heat removal should the RHR, or component cooling water, system become inoperable. These methods have a low heat removal capability but could be used to supplement steam generator heat removal until the decay rate was low enough.

These methods are heat removal through the chemical and volume control system nonregenerative and excess letdown heat exchangers (requires component cooling water) and cooldown flow from the pressurizer to the containment through the pressurizer relief valves with coolant injection from the safety injection or chemical and volume control systems. If a pipe break upstream of valves 700 and 701 should occur (i.e., a LOCA), the core could be adequately cooled by means of the RHR containment recirculation mode,via the containment sump.

The staff reviewed the RHR system under SEP Topics V-10.A, V-10.B, V-ll.A, and V-ll.B to ensure reliable plant shutdown capability using safety-grade equip-ment and used the guidelines of SRP Section 5.4.7, Regulatory Guide 1. 139, and Branch Technical Position (BTP) RSB 5-1 (NUREG-0800). The Ginna systems have been compared with these criteria, and the staff has concluded that the systems meet the criteria except for the requirement, for overpressure protection of the shutdown cooling system and for operating procedures to shut down and cool down using safety-grade systems.

Ginna SER 5-8

Overpressure relief capacity is required for the RHR system when operation; that is, when it it is in.

is not isolated from the reactor coolant system.

The overpressurization protection system (OPS) fulfills this function and is required by Technical Specifications. There is no procedural requirement in the Technical Specifications that ensures that the, OPS is, in service whenever the RHR system, is in service., During cooldown, present procedures place the RHR system -into service at 350 F and 360 psi, whereas the OPS is not required to be in service until the temperature, is. 330'. F. The OPS- must be in service when the reactor coolant system is less, than 330,.F, according to the Technical Specifications.,

If the RHR system were placed II I into, service before the OPS and a pressure transient occurred, there could be a pipe break in the RHR system outside containment.

The licensee has proposed to revise the Technical Specifications to require that the OPS be..in service whenever, the RHR system is not isolated from the reactor coolant system. The basis for this requirement, is to,prevent the over-pressurization, of the RHR system that could lead to a. LOCA outside containment.

The staff finds this acceptable.

10 CFR 50 (GDC 19 and 34), as implemented by SRP Section 5.4.7, BTP RSB 5-1, and Regulatory Guide 1. 139, requires that the plant can be taken from normal operat-ing conditions to cold shutdown using only safety-grade systems (assuming a single failure) and either onsite or offsite power through the use of suitable procedures. The Ginna plant has safety-grade plant systems capable of safe shutdown under these conditions; however, the plant operating procedures rely on other nonsafety-grade systems and do not specify how the cooldown would be accomplished by the operator if nonsafety-grade systems failed.

The licensee uses safety-grade as well as nonsafety-grade systems to bring the plant to a cold shutdown. Nonsafety-grade systems may not be available follow-ing either a seismic or loss-of-power event. Lack of sufficient information in operating procedures may adversely affect the time required to shut down the reactor plant; for instance, failure of the nonsafety-grade air system could result in inoperability of the steam atmospheric dump valves. Manual opening of the steam atmospheric dump valves would be required if the backup nonsafety-grade nitrogen system also failed.

The licensee has proposed to develop appropriate documented procedures for oper ation of safety-grade systems and components to achieve cold shutdown nonsafety-grade systems are unavailable. The staff finds this acceptable.

if The procedures will be developed and implemented in coordination with Item 7 of Supplement 1 to NUREG-0737, "Upgrade Emergency Operating Procedures."

5.4.4 Pressurizer Power-Operated Relief Valves Overpressure protection for the pressurizer and the reactor coolant system is provided by the power-operated relief valves (PORVs) and the pressurizer safety valves. The PORVs prevent the opening of the safety valves for mild pressure transients and can be manually opened from the control room. The motor-operated block valve upstream of the PORV provides a means of isolating a stuck-open or leaking PORV.

Ginna SER 5-9

During the transient, as a result of the steam generator tube rupture (SGTR) on January-25, 1982, the pressurizer PORV malfuntioned after opening and closing successfully three times. Mhen trying to close the valve after the fourth time it was opened, the PORV started to close then returned to.the full open posi-tion. The licensee performed extensive tests on the PORV and on the valves that control the air that operates the PORV and determined that an alteration performed on one of the air control valves in 1979 was not in accordance with the manufacturer s installation and maintenance instructions. The air control valve and its associated piping have been modified in accordance with the valve manufacturer's recommendations to ensure proper operation of the PORV. In NUREG-0916, the NRC staff concurred with the modification and concluded that future performance of the pressurizer PORV will be acceptable.

The staff has recently concluded that a capability for rapid primary system depressurization is needed for PMRs to effectively mitigate the design"basis SGTR accident. Furthermore, the staff ha4 concluded that the components and systems to provide this depressurization capability should be safety gr ade.

Mestinghouse plants do not have a safety-grade capability. The staff ac'tions on this issue regarding -operating reactors are unde'r consideration.

Ginna SER 5-10

c

~ I E(

inside Outside

~ cv Auxiliary Spray IIC PC V PCV oI 43S e36 I

Vent Header 203 Letdown Relief 31S Ptessurlter Sptay FC Seal Return Relief FC Demineralized Spray FC Hester Conuoi Accumulator rr Ruptwe Disc ~ I~

~ ss SI6 Presswir sr A etrst Tsnk Assrdust Hest Reactor Cooisnt Pressunrsr Swee bne Pump LH SIS S IS Stssm Gsnerstor Reactor Cootsnt Pump

~ I Loop B Chs reine flow CVCS Steam Generator Rsudusl Loop A Hest Removsl Rssctor loop Vessel SIS Accumulator 0

M o MOtOI-Operated ICV

~ Ir t

Valve LH letdown FO tine SIS ICVCSI SIS Auxiliary Charging Figure 5.1 Ginna reactor coolant system diagram

PWR STEAM GENERATOR Steam Outlet to Turbine Generator

<<Manway Upper Shell Foodwater Inlot r

~ I Downcomor Flow Resistance Plato Antivibration Bars Lower Shell

'Tube Supports Wrapper Tuba Bundle

,. 1! i)~ First Tube Support Plate I

Secondary.Side Inspection Handhole

)IIIIIIIIIIII Tubesheet Partition Manway Primary Coolant Pi ry Cool t Inlet Outlet Primary Channel Head Figure 5. 2 Ginna steam generator Ginna SER 5-12

Table 5. I Materials 'of construction of the.

reactor vessel Section Materials Dome plate (top and bottom) SA-302, Grade II B

Shell and,nozzle forgings A-508, Class2 Cladding, stainless steel weld rod Type 304 equivalent, Thermal shield and internals A-240, -Type 304 Table 5.2 Reactor vessel design data Parameter 'Value Design/operating pressure, psig 2,485/2,235 Hydrostatic test pressure, psig 3,110 Design temperature, F 660 Overall height of vessel and closure head, ft-in. 39-1.3 Water volume (with core and.internals in place), fts 2,473 Thickness of insulation, minimum, in. 3.0 Number of reactor closure head studs 48 Inner diameter (ID) of flange, in. 121. 81 ID at shell, in. 132 Inlet nozzle ID, in. 27. 47 Outlet nozzle ID, in. 28. 97 Core flooding water nozzle, in. 3.5 Clad thickness, minimum, in. 0. 156 Lower head thickness, minimum, in. 4. 125 Vessel belt-line thickness, minimum, in. 6.5 Closure head thickness, in. 5. 375 Ginna SER 5-13

Table 5.3 Comparison of blowdown chemistry, 1974-1981 Parameter 1974-1977 1978 1981 Gati on conducti v i ty, pmhos 0. 7-2. 5 0. 2-0. 4 0. 12-0. 2 Chloride, ppb <50 <10 3-5 Sodium, ppb 5-15 5-15 3-8 Silica, ppb 20-50 15-30 5-10 pH 8. 6-9. 0 8. 7-8. 9 8. 7-8 ~ 9 Ginna SER 5-14

6 ENGINEERED SAFETY FEATURES Ginna is a 1,520-MWt Westinghouse pressurized-water reactor (PWR) which uses a dry cylindrical reinforced concrete type containment. The engineered safety features provided include the containment air recirculation system, containment spray system, and safety injection system. The safety injection system consists of two passive accumulators, three high-pressure pumps, and two low-pressure pumps. In the event of loss of offsite power and failure of one diesel genera-tor, minimum safety injection is provided by two high-pressure pumps and one low-pressure pump, and minimum containment heat removal is provided by one con-tainment spray pump and two fan coolers.

6. 1 En ineered Safet Features Material The staff reviewed the plant design to ensure that organic materials, such as organic paints and coatings, used inside containment do not behave adversely during accidents when they are exposed to high radiation fields. On the basis of its review of SEP Topic VI-1, the staff concluded that the organic materials used in the plant are acceptable and will not interfere with the operation of engineered safety features under accident conditions. gualification tests demon-strate that the types of organic coating materials used in the containment will maintain their integrity and remain in serviceable condition after exposure to the severe environmental conditions of a design-basis accident (DBA). Insignif-icant quantities of organic gases and of hydrogen would be generated under these conditions.

To provide further assurance that delamination, flaking, or peeling of coating materials will,not interfere with the operation of engineered safety features, the licensee has proposed an acceptable inspection program for coated surfaces in containment.

Additionally, as part of SEP Topic VI-1, the staff reviewed the methods available to raise or maintain the pH of solutions expected to be recirculated within con-tainment after a DBA. Low pH solutions may accelerate chloride stress-corrosion cracking and increase the volatility of dissolved iodines.

The containment spray system, used both to reduce post-DBA containment pressure and to remove postaccident fission products from the containment atmosphere (especialy radioactive iodine), is automatically actuated by a high-high con-tainment pressure signal.

The containment spray system uses borated water, with a concentration ranging from 2,000 to 2,300'pm of boron, from the refueling water storage tank, and a sodium hydroxide solution of 30K by weight from the chemical spray. The result-ing solution drains to the containment sump. During the recirculation phase of containment spray or emerging core cooling systems (ECCSs), the containment solution is recirculated. An inert cover gas of nitrogen is provided for'ump the sodium hydroxide tank, which is housed in the heated auxiliary building.

Ginna SER 6-1

The staff has independently evaluated the pH of the containment sump solution that results from mixing of the containment spray solution with the reactor coolant and ECCS fluids in the containment sump during recirculation. The staff verified by independent calculations that sufficient sodium hydroxide is avail-able to raise the pH of the containment sump solution above the minimum level of 7.0, consistent with the guidance of Branch Technical Position (BTP) MTEB 6-1, to reduce the likelihood of stress-corrosion cracking of stainless steel compo-nents. The staff has also verified that the sump Naximum pH will not exceed a value of 10.5 as specified in SRP Section 6.5.2., The calculations were based on the volumes and concentrations provided by the licensee.

The plant Technical centrationn Specifications provide for demonstration of operability of the containment spray system and monthly testing of the sodium hydroxide con-in the chemical additive storage tank,. consistent with the Standard Technical Specifications for Westinghouse pressurized-water reactors (NUREG-0452).

In addition, the plant Technical Specifications contain limiting conditions for operation which specify the minimum volume and boron concentration for the re-fueling water storage tank, accumulators, and boric acid tanks and the minimum volume and sodium hydroxide concentration for the spray additive tank. The staff has reviewed the plant Technical Specifications and has determined that they provide sufficient assurance that the pH values will be within the range specified in SRP Section 6..5.2 and are acceptable.

6.2 Containment S stems The containment structure encloses the reactor system and is the final barrier

,against the release of radioactive fission products in the event of an accident.

The containment structure must, therefore', be capable of withstanding, without loss of function, the pressure and temperature conditions resulting from postulated loss-of-coolant accidents (LOCAs) and steam line break accidents.

Furthermore, equipment with a postaccident safety function must be environ-mentally qualified for the resulting adverse pressure and temperature conditions.

For PWR plants, the high-energy-line breaks that must be analyzed include primary system pipe breaks and secondary system pipe breaks. A break on the primary side generally results in the most severe pressure response in the con-tainment; a break on the secondary side results in the most severe temperature conditions in the containment.

The containment analysis for a postulated pipe break consists of two separate calculations. The first calculation include's the mass and energy release anal-ysis, which, for -primary system pipe breaks (LOCAs), includes the blowdown, reflood, and post-reflood phases, and results in mass and energy release rates into the containment. The second calculation is the containment response anal-ysis, which results in the containment temperature and pressure response to the mass and ener gy release from the postulated break.

The acceptance criteria used to evaluate the Ginna containment functional design analysis are based on the SRP. (NUREG-. 0800). For the containment analysis to be found acceptable, both the mass and energy release 'and the 'containment response calculations must meet the acceptance criteria specified in the SRP.

ll The staff's review of the containment functional design of the Ginna plant, as reported in SEP Topics VI-,2.D and VI-3, identified deviations from current Ginna SER 6-2

safety criteria. To assess the significance of these deviations independent containment analyses were performed. The results of the analyses show that the containment design conditions are not exceeded for postulated LOCAs, but are exceeded for the main steam line break (MSLB) accident. The staff noted how-ever, that the MSLB analysis was very conservative and that a more refined MSLB analysis might show that the containment design pressure would not be exceeded.

The licensee provided a, more realistic analysis, which showed that, for the worst-case MSLB, the calculated peak containment pressure is less than the de-sign pressure of 75 psia (60 psig) but the temperature exceeds the design-basis-temperature profile used in qualifying electrical equipment. However, the time over which this occurs. is short; that is, less than 60 seconds. Also, the Ginna plant has an automatic spray system designed to accommodate. a single active failure. Based on this fact and the guidance in the Division of Operating Reactors (DOR) Guidelines, for equipment qualification (letter dated February 19, 1980), the containment temperature/pressure profiles for the worst-case LOCA are acceptable for use in equipment qualification.

The containment isolation system of a nuclear power plant is an engineered safety feature that functions to allow the normal or emergency passage of fluids through the containment boundary while preserving the ability of the boundary to prevent or limit the escape of fission products to the environs that'ay result from postulated accidents. The safety review criteria for the contain-ment isolation system have changed since Ginna began operation in 1971. SEP Topic VI-4 documents the deviations from current safety- criteria as they relate to the containment isolation system. The resolution of these differences is provided in the Integrated Plant Safety Assessment Report (NUREG-0821). The staff has required, or the licensee has proposed, a number of modifications to be performed on various containment isolation penetrations. The staff finds the containment isolation system at Ginna acceptable pending completion of these modifications.

\

Containment leakage testing requirements are contained in Appendix J to 10 CFR 50. By letter dated March 28, 1978, the staff issued Amendment No. 17 to Provisional Operating License (POL) No. DRR-18, which modified the Ginna Technical Specifications regarding containment leakage testing and granted certain exemptions from the requirements of Appendix J. The staff's letter dated May 6 1981, documented the completion of the Appendix J review for Ginna.

6. 3 Emer enc Core Coolin S stem i

k Emergency core cooling is provided by the safety, injection system (SIS), which constitutes the ECCS. The SIS components operate in three modes - passive accumulator injection, active safety injection, and residual heat removal recir-culation. The primary purpose of the SIS is to deliver cooling water automati-cally to the reactor core to limit the fuel cladding temperature and thereby ensure that the.,core will remain intact and. in place with its heat transfer geometry preserved; This protection is prescribed for all breaks (up to and including a hypothetical instantaneous double-ended rupture of the reactor coolant pipe), for a rod ejection accident, and for a steam generator tube.,

rupture. A more 'complete description of,the SIS is given in the staff's eval-.

uation of SEP Topic VI-7.A.3.

Ginna SER 6-3

In 1975, the staff reviewed the Ginna ECCS with respect to the single-failure criterion and issued Amendment No. 7 to POL No. DPR 18 (letter dated May 14, 1975). , As part of SEP Topics VI-7.C and VI-7.C.2, the staff determined that the initial safety evaluation accompanying the amendment was sufficient and that no further review or analysis was necessary.

Originally, LOCA analyses for all Westinghouse re'actors were conducted assuming that the temperature of the water in the upper-head region of the reactor vessel was the same as the inlet water temperature because of a bypass flow from the downcomer to the upper head. Temperature measurements made by Westinghouse indicate that the actual temperature of the upper-head fluid is almost as high as the hot-leg (outlet) temperature. All operating reactors were required to resubmit LOCA analyses using hot-leg temperature for the upper-head volume.

The staff's safety evaluation of Amendment No. 19 to POL No. DPR-18, forwarded by letter dated May 1, 1978, accepted revised analyses with upper-head fluid temperature equal to the hot-leg (outlet) temperature.

The Amendment No. 19 safety evaluation also addressed upper plenum injection (UPI) at Ginna. The generic evaluation model assumes that all safety injection water is introduced directly into the lower plenum. For two-loop units such as Ginna, the safety injection water is injected into the upper plenum. Thus, the staff was concerned that the Westinghouse model did not consider interaction between UPI water and steam flow. After plant-specific submittals by licensees operating two-loop plants were reviewed, the staff concluded that the calcula-tions provided by the licensees (with certain modifications to: the staff's model) are acceptable on an interim basis for continued safe operation of Westinghouse two-loop plants. Long-term efforts for developing a model specifically treating UPI are continuing. For the Ginna plant the calculations that specifically con-sidered UPI,using the modified version of the staff model resulted in a change of only 15F'ompared with those using the generic model in which the UPI-core interaction was not specifically considered. In the interim (before 'these models are developed), Ginna provided a modification to the current Westinghouse model that accounts for UPI-core interaction. It was demonstrated'hat modifi-cation resulted in an increase of peak cladding temperature (PCT) of 15F~.

Since for the Ginna plant both Exxon Nuclear Corporation (ENC) WREM-II and Westinghouse models predict similar PCTs (1,922 F for'ENC WREM-II and 1,957 F for Westinghouse), it can. be expected that the UPI modification, when applied to the ENC WREM-II model, will result in about the same increase in PCT.

By letter dated February 17, 1983, RGKE submitted by reference an Exxon-developed ECCS model that specifically treats upper plenum'njection. The staff is currently reviewing this submittal and will issue its safety evalua-tion of this subject separately. P SIS testing is performed at each reactor refueling interval, with the reactor coolant system pressure less than or equal to 350 psig and temperature less than or equal to 350 F. A test signal is applied to initiate operation of the system. The safety, injection and residual heat'removal (RHR) pump motors are prevented from starting during the test: The system is 'considered satisfactory if control board indication and visual observations indicate that all valves.

have jeceived the safety injection signal and have completed their travel..

Except during cold or refueling shutdowns, the safety injection pumps and RHR pumps are started at intervals not to exceed'1 month. The acceptable level of Ginna SER

performance for the RHR pumps is 200 gpm at the minimum discharge pressure of 140 psig. The acceptable level of performance for the safety injection pumps is 50 gpm at the minimum discharge pressure of 1,420 psig. The spray additive valves are tested at intervals not to exceed 1 month. Mith the pumps shut down and the valves upstream a'nd downstream of the spray additive valves closed, each valve is opened and closed by operator action. The accumulator check valves are checked for operability at each refueling outage. In SEP Topic VI-7.A.3, the staff found the licensee's testing program of the ECCS acceptable.

Ginna SER

7 INSTRUMENTATION AND CONTROLS

7. 1 Reactor Protection S stem The reactor protection system (RPS) automatically trips the reactor to protect against reactor coolant system damage caused by high system pressure and to protect the reactor core against fuel rod cladding damage caused by a departure from nucleate boi ling (DNB) under the,.following conditions:

(1) Reactor power reaches a preset limit.

(2) Excessive temperature rise occurs across the core.

(3) Pressurizer pressure reaches an established minimum or maximum limit.

(4) Pressurizer level reaches an established maximum.

(5) There is a loss of reactor coolant flow.

The basic reactor, tripping philosophy is to define a region of power and coolant temperature and pressure conditions allowed by 'the primary tripping functions (overpower high hT trip, overtemperature high hT trip, and nuclear overpower trip). The allowable operating region within these trip settings is provided to prevent any combination of power, temperature, and pressure that would result in a DNB with all reactor coolant pumps in operation.

1 ill Additional tripping functions such as a high pressurizer'pressure trip,.low pressurizer pressure trip, high pressurizer water level trip, loss-of-flow trip, steam and feedwater flow mismatch trip, steam generator low-low water level trip, turbine trip, safety injection trip, nuclear source and intermediate range trips, and manual trip are provided to back up the primary tripping functions for specific accident conditions and mechanical- failures.

The Ginna reactor possesses high-speed Westinghouse magnetic-type control rod drive (CRD) mechanisms. The reactor internal components, fuel assemblies, rod cluster control assemblies, and drive systems components are designed as seismic Category I equipment.

Two reactor trip breakers are provided, to interrupt power to the- CRD mechanisms.

The breaker main contacts are connected in series with the power supply to the mechanism coils. The trip breakers are opened by the undervoltage coils on both breakers (normally energized), which become deenergized by any one of the several trip signals. Each protection channel actuates two separate trip logic trains, one for each reactor trip breaker undervoltage trip coil. The electri-cal state of the devices providing signals to the circuit breaker undervoltage trip coils causes these coils to trip the breaker in the event of reactor trip or power loss. Opening either breaker interrupts power to the magnetic latch mechanisms on each CRD, causing them to release the rods and.allowing the rod clusters to insert by gravity into the core. The reactor shutdown function of the rods is completely independent of the normal control functions because the trip breakers completely interrupt the power supply to the rod mechanisms and thereby negate any possibility of response to control signals. The control rods must be energized to remain withdrawn from the core. An automatic reactor trip occurs on loss of power to the control rods.

Ginna SER 7" 1

As a result of the Salem anticipated transients without scram (ATWS) events, the Commission published NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant." In a letter dated July 8, 1983, the staff identified the actions licensees needed to take based on NUREG-1000. The actions address issues related to reactor trip system reliability and general management capability. Resolution of these actions will be reported seperately.

The RPS is designed on a channelized basis to achieve isolation and independence between redundant protection channels. The coincident trip philosophy is carried out to provide a safe and reliable system because a single failure will not defeat the function of the channel and also will not cause a spurious plant trip. Channel independence is carried throughout the system from the sensor to the relay providing the logic. The channelized design that applies to the ana-log as well as the logic portions of the protection system is discussed below.

Isolation of redundant analog channels originates at the process sensors and continues back through the field wiring and containment penetrations to the analog protection racks. When the safety and c'ontrol functions are combined, both functions are fully'solated in the remaining part of the channel; control isf derived from the primary safety signal path through an isolation amplifier.

1 The staff. concluded in SEP Topic VII-1.A that the isolation devices satisfy the current licensing -criteria and, as such, a failure in the control circuitry does not affect the safety channel. This approach also 'is used for pressurizer pressure and water level channels, steam generator water level and bT channels, steam flow-feedwater flow and nuclear power range channels.

Physical separation is used to achieve isolation of redundant transmitters.

Separation of field wiring is achieved using separate wireways, cable trays, conduit runs, and containment penetrations for each redundant channel. Analog

.equipment is separated by locating redundant components in different protection racks.

The power supplies to the channels are fed from four instrument buses. 'wo of the buses are supplied by. constant voltage transformers, and two are'upplied by inverters. Each channel is energized from a separate ac power 'feed. Each reactor trip circuit is designed so that a trip occurs when the circuit is deenergized. An open circuit or the loss of channel power, therefore,'auses the system to go into its trip mode. Reliability and independence are obtained by redundancy within each tripping function. In a two.-out-of-three circuit, the three channels are equipped with separate primary sensors and each channel is energized from an independent electrical bus. A single failure may be applied in which a channel fails to deenergize when required; however, such a malfunction can affect only one channel. The trip signal furnished by the two remaining channels is unimpaired in this event.

All reactor protection channels are supplied with sufficient redundancy the capability for channel calibration and testing at power. Bypass to'rovide removal of one trip circuit is accomplished by placing that circuit in a half-tripped mode; that is, a two-out-of-three circuit becomes a one-out-of-.,two circuit. Testing does not trip the system unless a trip condition concurrently exists in a redundant channel.

Ginna SER 7-2

Certain reactor trip channels are automatically bypassed- at low power, to allow for ~such conditi'ons as startup and shutdown, and where they are not required for safety. Nuclear source range and intermediate range trips, which specifically provide protection at low power or subcritical operation, are bypassed at power operation to prevent spurious reactor trip, signals and to improve reliability.

The reactor trip bistables are mounted in the protection racks and are the final operational components in an analog protection channel. Each bistable drives two logic relays (C and 0). The contacts from the C relays are interconnected to form the required actuation logic for trip breaker No. 1 through dc power, feed No. 1. The transition from channel identity to logic identity is made at the logic relay coil/relay contact interface. As such, there is both electrical and physical separation between the analog and the logic portions of the pro-tection system. The above logic network is duplicated for trip breaker No. 2 using dc power feed No. 2 and the contacts from the 0 relays. Therefore, the two redundant reactor trip logic channels will be physically separated and electrically isolated from one another. Overall, the RPS consists of identifi-able channels, which are physically, electrically, and functionally separated and isolated from one another. A description of the various RPS trip functions is given in SEP Topic VI-lO.A.

7.2 En ineered Safet Features Actuation S stem ESFAS The engineered safety features (ESF) as defined by SRP Section 7. 1-III are those systems that are required to mitigate the consequences of a postulated accident.

The systems for Ginna, as identified in the Technical Specifications, which are actuated by the ESFAS are (1) safety injection (2) steam line isolation (3) containment containment isolation spray'4)

(S) feedwater isolation The safety injection system (SIS) delivers borated water to the reactor core following a LOCA. The principal components of the SIS are two passive accumu-lators (one for each loop), three safety injection pumps, two residual heat removal (RHR) pumps, and the essential piping and valves. The, accumulators, being passive devices, discharge into the cold leg of each loop and do not require automatic initiation.

The SIS may be actuated by two-out-of-three (2/3) low pressurizer pressure signals, 2/3 low steam line pressure signals, 2/3 high containment pressure signals, or it can be actuated manually. Any of the SIS signals will open the SIS isolation valves, start the high-head safety injection pumps and the low-head RHR pumps.

The steam line isolation valves are closed upon receipt of high steam line flow in conjunction with an SIS signal, by containment pressure, or by manual

'nitiation.

The containment spray system consists of two pumps, one spray additive tank, valves, piping, and spray nozzles. Containment spray is initiated by coinci-dent signals from two sets of 2/3 high containment pressure signals monitoring Ginna SER 7"3

containment: high-high pressure. "

The actuation signal starts the pumps and opens the discharge valves to the spray header. Valves for the 'spray additive

'ank open after a time delay and may be controlled by the operator.

Containment isolation is, initiated by an SIS signal or manually. Actuation of containment isolation trips the containment sump pumps, closes all containment isolation valves (containment sump pump discharge isolation valves, steam generator isolation blowdown valves, reactor coolant 'drai'n tank vent header and pump suction valve, containment ventilation purge valves, containment depres-surization valves, containment air test supply valve, and containment air test valves), and trips the purge supply and exhaust fans. Containment ventilation valves are also isolated on high containment activity or from a manual contain-ment spray signal.

The feedwater isolation system consists of the four main feedwater and four feedwater bypass isolation valves. These valves close when they receive an SIS signal or an ESF sequence initiation signal. They fail closed if power or air is lost.

In SEP Topic VII-2, the st'aff concluded that with the exception of the contain-ment radiation monitors, which do not qualify as Class 1E equipment, the ESF actuation system meets current licensing criteria. The qualification of the radiation monitors will be addressed in the implementation of Regulatory Guide 1.97 as part of Supplement 1 to NUREG-0737.

discussed in SEP Topic VI-7.C.1, the staff reviewed the licensee's modified t's emergency core cooling system (ECCS) to confirm that it is designed to meet the most limiting single failure (i.e., the loss of a single ac or dc onsite power system). The staff found that the ECCS did not satisfy the review criteria for the separation between redundant systems.

However, the short-circuit analysis provided in the licensee's letter of July 14, 1981, shows:

(1) fusing has been coordinated so that faults will be cleared before dc bus trans fer.

a 4 (2) The automatic transfer schemes for buses 14, 16, 17, and 18 and DG1A con-trol panel have electrical interlocks .to prevent the paralleling of-the two dc systems.

(3) The two dc systems can be paralleled when the two systems are purposely tied together during the test of one set of batteries or duri'ng the main-tenance or repair of a main 150-A charger unit.

(4) No credible component failure can cause the paralleling of the two dc systems through the manual switches on the 4 kV non-Class 1E buses.

(5) The automatic transfer scheme used for the main control board annunciators is designed so that only one of the two dc sources can be connected.

On the basis of its review of the licensee's calculations, the staff concludes that the present design and administrative controls provide an acceptable alternative to the criteria provided that fuse types and sizes, battery capacity, and electrical loads are not changed.

Ginna SER 7-4

7.3 S stems Re uired for Safe Shutdown In the SEP review of safe shutdown systems (Topic VII-3) for Ginna, the staff and the licensee developed a list of the minimum systems necessary to take the reactor from operating conditions to cold shutdown. Although other systems may be used to perform shutdown and cooldown functions, the following systems are the minimum number required to fulfill the requirements of BTP RSB 5-1:

(1) reactor protection system (2) auxiliary feedwater system (3) main steam system (4) service water system (5) chemical and volume control system (6) component cooling water system (7) residual heat removal system (8) electrical instrumentation and power systems for the above systems In SEP Topic VII-3, the staff noted that the systems required to take the reac-tor from hot shutdown to cold shutdown (assuming only offsite power is available or only onsite power is available with a single failure) are capable of initia-tion to bring the plant to safe shutdown and are in compliance with current licensing criteria and the safety objectives of SEP Topic VII-3, except that long-term cooling (RHR) is susceptible to single electrical instrumentation and control (EI&C) failures, which render this form, of long-term cooling inoperable.

However, other systems are available for the removal of decay heat.

The instrumentation available to control room operators to place and maintain the reactor in cold shutdown conditions meets current licensing criteria because no single EI&C,failures render vital parameters such as reactor pressure and temperature inoperable.

The capability to maintain the reactor in hot shutdown from outside the control room exists and is in compliance with the safety objectives of SEP Topic VII-3.

No procedure exists to take the plant from hot to cold shutdown from outside the control room. However, all the required systems and components could be operated at local stations throughout the plant and, therefore, are acceptable.

The staff, therefore, concludes that with the installation (during the 1983 refueling outage) of a redundant component cooling water, surge tank level indi-cation, Ginna satisfies all of the requirements for safe shutdown, including GDC 17, because of the number and quality of systems provided, an 8-hour battery capacity, and the capability to establish a delayed access line by backfeeding through the main transformer in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

7.4 Other S stems Re uired for Safet 7.4. 1 Overpressure Protection During Low Temperature Operation The low temperature overpressure protection system at Ginna is described =in Section 5.2. In its Safety Evaluation in support of Amendment No. 26 to Ginna's license (letter dated April 18, 1979), the staff concluded that the design of the low temperature overpressure protection system in the areas of electrical instrumentation and control (EI&C) is in accordance with the design criteria originally prescribed by the NRC and later expanded d'uring subsequent discus-sions with RG&E.

Ginna SER 7-5

The staff finds the EI8C aspects of the design acceptable because:

(1) The overpressure protection system complies with Institute of Electrical

'and'Electronics Engineers '(IEEE) Std. 279-1971 and seismic criteria as identified in Section 2.

(2) The system is redundant 'and satisfies the single-failure criterion.

(3) The design requires no operator action for '10 minutes after the operator receives an overpressure action alarm.

(4) The system is testable on a periodic basis.

The changes to the Technical Specifications reduce the probability of overpressurization events to acceptable levels.

7.4.2 Engineered Safety Feature Switchover From Injection to Recirculation Mode In SEP Topic VI-7. 8, the staff reviewed the operator action rhquired to realign the ECCS for'the recirculation mode following a LOCA to determine if automatic switchover is necessary to protect public health and safety. Item 19 of SRP Section 6.3 states that the complete sequence of ECCS operation from in'jection to long-term core cooling (recirculation) should be examined to see that minimal manual action is required, and that, where manual action is needed, sufficient time is available for the operator'o respond. The current Ginna procedures for switchover from injection to recirculation do not meet current NRC criteria for operator actions.

The limited probablistic risk assessment study performed in conjunction with the SEP integrated assessment concluded that there will only be a small reduc-tion in risk because of procedural modifications. However, the staff feels that improvement of the switchover procedure to preclude the interruption of core cooling flow is prudent.

The licensee has agreed to modify the switchover procedure in order to meet the necessary operator action criteria.'he actual procedures will be developed and implemented in coordination with Item 7 of Supplement 1 to NUREG-0737.

7.4.3 Accumulator Isolation Valves Power and Control System Design For many LOCAs, the performance of the ECCS in PWR plants depends on the proper functioning of the accumulators. The motor-operated isolation valve, provided between the accumulator and the primary system, must be considered an "operating bypass" (IEEE Std. 279-1971) because, when closed, it prevents the accumulator from performing its intended protective function. Accordingly, the motor-operated isolation valve should be designed against a single failure that can result in a loss of capability to perform a safety function.

An additional operational requirement for these valves is that they be closed to permi,t primary system depressurization during reactor shutdown. (See the staff's discussions of SEP Topics V-3 and VII-3.)

The staff's review of SEP Topic VI-7. F found that the Ginna accumulator isola-tion valve power and control system design meets the requirement of BTP ISCB 18, Ginna SER

Part 2, but that the plant does not satisfy BTP ICSB 18, Part 4, because after locking out power, one channel of valve position indication is available instead of two. However, the staff concluded that the single valve position indicator combined with the administrative procedures for removing power from the valve and the verification of valve position meets the intent of redundant valve position indicators and is, therefore, acceptable.

7.4.4 Frequency Decay Issue 9 of NUREG-0138 states that the staff should require that a postulated rapid decay of the frequency of the offsite power system be included in the accident analysis and that the results be demonstrated to be acceptable.

Alternatively, the reactor coolant pump (RCP) circuit breakers should be designed to protection system criteria and tripped to separate the pump motors from the offsite power system because rapid decay of the frequency of the off-site power system has the potential for slowing down or braking the RCP, thereby reducing the cooling flow rates to levels not considered in previous analyses.

Oak Ridge National Laboratory (ORNL), under a technical assistance program, reviewed the frequency decay rate phenomena and its effects on RCPs. The results of the review are presented in Section 4 of NUREG/CR-1464. In summary, the report shows that the conditions required for dynamic braking of RCPs are a sustained and rapid decrease in frequency while bus voltage is maintained.

These conditions are only realized in a highly capacitive system using large amounts of buried transmission cables (such as Long Island). The RG8E system does not use large amounts of buried transmission cables. Therefore, the necessary conditions are not present in the Ginna offsite electrical distribu-tion system. Accordingly, in SEP Topic VII-6 the staff concluded this issue is not applicable to Ginna.

7. 4. 5 Electric Type Valve Operators In U.S. Atomic Energy Commission AEC Bulletin 72-3 dated December 1, 1972, the NRC notified RG8E that Limatorque valve operator models SMB-00 and SMB-000 were considered to have a common-mode failure in that a weak "torque switch torsion spring" could cause malfunction of the valve operator torque switch. The valve operators in question were identified as a specific production 'imatorque group that was manufactured between 1969 and mid-1971. RG8E was requested to notify the staff of any valve operators of the described make, model, and vin-tage installed in the plant and the corrective actions that were being taken.

By letter dated December 18, 1972, RGKE notified the NRC that although many electric type valve operators of that make and model are installed in the plant, all of them were manufactured before 1969.

Additionally, the staff reviewed the electric type valve operators as part of SEP Topic III-10.A. The staff found that although there have been some torque switch problems, they have resulted in less than 9.4X of all motor-operated valve failures. Therefore, the staff found that the Ginna designs satisfy the current licensing criteria for all safety-related valve functions'inna SER 7-7

8 ELECTRIC POWER SYSTEMS

8. 1 Potential E ui ment Failures Associated With De raded Grid Volta e tibilityy By letter dated June 3, 1977, the staff requested that RG8E assess the suscep-of the Ginna Class lE electrical equipment to sustained degraded voltage conditions at offsite power sources and to the interaction between the offsite and onsite emergency power systems. In addition; the staff requested that the licensee compare the current design of the emergency power systems at the plant facilities with the NRC staff positions as stated in the letter of June 3, 1977, and that the licensee propose plant modifications, as necessary, to meet the staff positions, or provide a detailed analysis that shows that the facility design has equivalent capabilities and protective features. Further, the staff required that certain Technical Specifications be incorporated into all facility operating licenses.

RG8E proposed certain design modifications and additions to the licensee's Technical Specifications. These design modifications include the, installation of a degraded voltage protection system for the Class 1E equipment. The pro-posed additions to the Technical Specifications are in regard to the setpoints, calibrations, and surveillance requirements associated with the proposed voltage protection system.

On the basis of information provided by RG8E, the staff determined in the safety evaluation for Amendment No. 38 to Ginna's license (letter dated March 26, 1981) that the modifications comply wi.th all of the staff's requirements and design-basis criteria have been met. The voltage setting and time delays protect the Class 1E equipment from a sustained degraded voltage condition of the offsite power source.

The license has retained the load-shed feature while the onsite sources are supplying the Class lE buses and has included in the Technical Specifications the maximum limits of the setpoint values of the loss-of-voltage (load-shed feature) relay. A review of the setpoint values, limits, and logic circuitry has determined that there will be no adverse interaction of the onsite sources with the load-shed feature during load sequencing. The additions to the Technical Specifications and the method of testing the logic circuitry have been reviewed and found to meet the staff's requirements.

The staff, therefore, concludes that 'the licensee's modifications meet the criteria established by the NRC for the protection of Class 1E equipment from grid voltage degradation.

8. 2 Ade uac of Station Electric Distribution S stem Volta es As part of SEP Topic VIII-l.A, RG8E reviewed the electric power system at Ginna to (1) determine analytically the capacity and capability of the offsite power system and onsite distribution system to automatically start as well as Ginna SER 8-1

operate all required loads within their required voltage ratings in the event of (a) an anticipated transient or (b) an accident (such as LOCA) without manual shedding of any electric loads (2) determine if there are any events or conditions that could result in the simultaneous or consequential loss of both required circuits from the off-site network to the onsite electric distribution system, thus violating the requirements of GDC 17 A detailed review and technical evaluation of the licensee's submittals were performed by Lawrence Livermore Laboratory (LLL) under contract to the NRC, general supervision by NRC staff. -This work is reported by LLL in Tech- 'ith nical Evaluation Report (TER), "Adequacy of Station Electric Distribution System Voltages for the R. E. Ginna Nuclear Power Station Unit 1," dated November 5, 1981.

The staff has reviewed the LLL TER and concurs 'in the findings that:

Under worst-case conditions, the Class lE equipment wi 11 automatically start and continue to operate within its voltage design rating.

(2) The voltage of the Class lE equipment will not exceed the upper design voltage rating under maximum offsite voltage and minimum plant loading conditions.

(3) The analysis submitted was verified by test. The test data indicate that the analytical results are lower than actual measured values; thus, the model is conservative with acceptable percentage error differences.

(4) Spurious.,trips will not occur for the voltages and plant operating condi-tions analyzed.

8. 3 Onsite Emer enc Power S stems - Diesel Generator Under SEP Topic VIII-2, the staff reviewed the onsite ac generators for the Ginna plant to determine if they had sufficient capacity and capability to supply the required automatic safety loads during anticipated occurrences and/or in the event of postulated accidents after a loss of offsite power. The staff evaluated the loading of the diesel generator and the bypass of the protective trips during accident conditions and periodic testing. The staff found that automatic diesel generator loading .is in compliance with current licensing criteria. The bypass of the diesel generator 'protective trips is in agreement with current NRC staff guidelines. In the safety evaluation accompanying Amendment No. 41 to the Ginna license (letter dated April 23, 1981(a)),'he staff approved revised Technical Specifications for diesel generator testing.

The licensee has modified his annunciators to ensure that there are no condi-tions that are not alarmed that might render the diesel generators incapable of automatic start. The staff, therefore,'concludes that the Ginna onsite emer-gency power systems meet current licensing criteria.

8.4 Station Batter Ca acit Test Re uirements To ensure that the onsite Class lE battery capacity is adequate to supply dc power to all safety-related loads required by the accident analyses and is Ginna SER 8-2

verified on a periodic basis, the staff reviewed the Ginna Technical Specifica-tions, including the test program,, with regard,to the requirement for periodic surveillance testing of onsite Class lE batteries and the extent to which the test meets Section 5.3.6 of IEEE Std. 308-1974 to determine battery capacity.

The Ginna battery surveillance requirements are included in Section 4.6.2 of the plant's Technical Specifications. There is no periodic battery discharge test required. Therefore, the Ginna plant does not comply with the current licensing requirements for station battery capacity tests. The staff' posi-tion in SEP Topic VIII-3.A was that the battery discharge test is a more severe test than that which was being used and should be adopted.

In response to the staff's position, the licensee performed the battery dis-charge test during the spring 1982 refueling outage and has committed to pro-pose appropriate changes to the Technical Specifications for battery testing.

Therefore, the staff concludes that this item is adequately resolved.

8.5 DC Power S stem Bus Volta e Monitorin and Annunciation To ensure the design adequacy of the dc power system battery and bus voltage monitoring and annunciation schemes so that the operator can (1) prevent the loss of an emergency dc bus or (2) take timely corrective action in the event of loss of an emergency dc bus, the staff reviewed the dc power system battery, battery charger, and bus voltage monitoring and annunciation design with respect to the indication of dc power system operability status to the operator.

The staff' review of SEP Topic.VIII-3.B determined that the Ginna control room has no indication of battery current, battery charger current, charger output voltage, battery high discharge rate, under/overvoltage, or battery or charger breaker/fuse status. The limited PRA, performers in conjunction with the SEP integrated assessment, determined that the batteries are an important contribu-tor to risk. The staff's position is that, at a minimum, battery current and charger output current have local indication and a dc trouble alarm in the con-trol room so that the operator will be alerted to the operability status of the power system. Also, the breaker/fuse status should be administratively con-trolled.

The licensee agreed to provide additional dc system monitoring to assure the control room operator that a low-resistance tie exists between the vital .bat-teries and the dc loads. A "dc system trouble" alarm annunciated in the control room and local current and switch position indications in the battery rooms were installed during the spring 1983 refueling outage.

8.6 Electrical Penetrations of Reactor Containment Under SEP Topic VIII-4, electrical penetrations in the containment structure were reviewed to ensure that they do not fail from electrical faults during a high-energy-line break. As part of the SEP, the staff performed an audit,'com-paring sample containment electrical penetrations with current licensing criteria for protection against fault and overload currents following a postu-lated accident.

The initial topic review showed that with a loss-of-coolant-accident environment inside containment, the backup protection for some penetrations did not conform Ginna SER 8-3

to current licensing criteria. The licensee has provided the necessary circuit protection to ensure that electrical penetrations conform to current licensing criteria. The staff, therefore, concludes that the electrical penetrations in containment are acceptable.

Ginna SER 8-;4

9 AUXILIARY SYSTEMS The purpose of SEP Topic IX-1 was to review the storage facility for new and irradiated fuel, including the cooling capability and seismic classification of the fuel pool cooling system of the spent fuel storage pool in order to ensure that'new and irradiated fuel is stored safely with respect to criticality, cooling capability, shielding, and structural capability.

The staff reviewed the spent fuel pool modifications as described in Amendment No. 11 to POL No. OPR-18 (letter dated November 15, 1976). The staff determined that the safety evaluation supporting the amendment was performed in accordance with current licensing criteria. This review satisfies the aspects of Topic IX-1 relating to criticality and the structural capability of the storage racks.

By letter dated November 3, 1981, the staff issued a Safety Evaluation regard-ing the proposed spent fuel pool cooling system (SFPCS) modifications. The evaluation concluded that the proposed SFPCS modifications and the existing

backup systems were acceptable.

The structural response of the Ginna plant with respect to seismic capability has been reviewed and presented in NUREG/CR-1821. Although the spent fuel pool structure was not specifically evaluated during the seismic review, the overall conclusion was that the Ginna plant structures and structural elements are adequately designed to withstand the postulated earthquake.

The new fuel storage area is located in the auxiliary building. New fuel is dry in the fuel storage area. The primary concern would be flooding of 'tored the storage area with the potential for inadvertent criticality.

The new fuel storage facility is designed to provide center-to-center spacing of 21 in., which would maintain Keff if ff <0.90 even the facility were filled with unborated water. In addition, the new fuel storage area is covered with locked steel plates that would prevent sudden flooding of the area. Leakage through the steel plates would be removed through a drain in the new fuel enclosure.

On the basis. of the above considerations, the staff concludes that the new fuel storage facility meets the guidance of SRP Section 9. 1. 1.

K In SEP Topic IX-3, the staff reviewed the licensee's component cooling water (CCW) system and service water system (SWS) to ensure that. the systems have the capability to meet the design objectives and, in particular, to ensure that:

,(1) Systems are provided with adequate physical separation so that there are no adverse interactions among those systems under any mode of operation.'inna SER 9-1

(2) Sufficient cooling water inventory has been provided or that adequate pro-visions for makeup are available.

(3) Tank overflow cannot be released to the environment without monitoring and unless the level of radioactivity is within acceptable limits.

(4) Vital equipment necessary for achieving a controlled and safe shutdown is not flooded as a result of the failure of the main condenser circulating water system.

On the basis of the staff's review of the service and cooling water systems for Ginna, the staff has concluded that the essential systems and functions are (1) component cooling water: residual heat removal heat exchanger cooling and emergency core cooling system pump cooling (2) service water system: all components supplied by the critical" supply headers The staff has determined that the design of the above systems is in conformance with current regulatory guidelines and with GDC 44 regarding capability and redundancy of the essential functions of the systems, except for the SWS Technical Specification, the CCW surge tank level, and the pressure sensor on the CCW pumps.

The present Technical Specifications require that two of the four SMS pumps and SWS,loops be operational. However, if this plant were operating with the mini-mum number of required SWS pumps aligned to one bus and an accident occurred, the possibility exists that no SWS pump would be available. This is based on the assumption that one of the two emergency diesel generators would fail to start. Although the licensee's operation requires that an SWS pump be aligned to each bus, the staff's position is that the Technical Specifications be made more explicit to ensure that the two operating SWS pumps are not serviced by the same diesel generator. The licensee has agreed to modify the Technical Specifications accordingly.

The CCW surge tank level is measured by a single transmitter with indication provided in the control room. This does not satisfy Section 4.20 of IEEE Std. 279-1971.

The surge tank level is an important parameter to measure because it gives an anticipatory indication of possible CCW loss of water, which would lead to loss of the system function.'ailure of the CCM system could affect the ability to safely shut down. The basis for this is that failure of the existing single sensor in the "high" condition could give an erroneous and misleading indica-tion to the operator if the surge tank level was low. The operator might not detect a low CCW surge tank level under this condition in sufficient time to correct a problem in the CCW system.

The limited PRA study performed by Sandia National Laboratories reported that the importance of this issue was low. However, the staff considered the uncer-tai'nties in the PRA study and additional factors during the SEP integrated assessment and concluded that the relative importance of this issue is suffi-ciently high to warrant the installation of additional alarms and a transmitter.

Ginna SER 9-2

The licensee has proposed to add during the next refueling outage another transmitter to the surge tank with level alarms that is independent of the pre-sent indicator in the control room. The staff finds this acceptable.

The output pressure of the CCW pumps is measured by a single pressure sensor and transmitter with indications provided in the control room. This does not satisfy Section 4. 20 of IEEE Std. 279-1971.

Failure of the single pressure sensor or transmitter could give an erroneous indication to the operator that the CCW system is operating properly. Failure of the CCW system could affect the ability to shut down safely.

I The need for a redundant pressure sensor is considered low because there are other means of detecting inadequate cooling water flow in the CCW system (flow meters, temperature monitors, and alarms on components cooled by CCW such as the CCW heat exchanger, the control rod drive mechanisms, and the primary coolant pump seals). The limited PRA study found that resolution of this issue has a small impact on the availability of the CCW system and ranked the issue of low importance. Therefore, during the integrated assessmeht, the staff did not recommend backfitting this item.

The systems also meet the requirements of GDC 45 and 46 regarding system design to permit periodic inspection and testing.

As discussed in Section 3.6, the CCW surge tank will be structurally upgraded to withstand the effects of natural phenomenon such as earthquakes.

9.3 Boron Addition S stem Following a LOCA, boric acid solution is introduced into the reactor vessel by two modes of injection. In the initial injection mode, borated water is pro-vided from the boric acid tanks, the refueling water storage tank, and the accumulators. After this. initial period, which lasts at least 20 minutes for a large-break LOCA and longer for smaller breaks, the emergency core cooling system (ECCS) is realigned for the recirculation mode. In this mode borated water is recirculated from the containment sump to the reactor vessel and back to the sump through the break. A portion of the water introduced into the reactor vessel is converted into steam by the decay heat generated in the core.

Because the steam contains virtually no impurities, the boric acid content in the water that was vaporized remains in the vessel. The concentration of boric acid in the core region, therefore, will continuously increase, unless a di lu-tion flow is provided through the core. Without the dilution flow the concen-tration of boric acid will eventually reach the saturation limit and any further increase in boric acid inventory will cause its precipitation. Boric acid deposited in the core may clog flow passages and seriously compromise the per-formance of the ECCS. SEP Topic IX-4 reviewed the boron addition system, in particular with respect to boron precipitation during the long-term cooling mode of operation following a LOCA, to ensure that the ECCS is designed and operated in such a manner that a sufficient throughflow is provided before the concentration of boric acid reaches its saturation limit.

The design of the Ginna reactor is different from current Westinghouse designs in two areas that affect boron precipitation. One is that the residual heat removal (RHR) injection feeds directly into the upper plenum rather than into Ginna SER 9-3

the cold or, hot legs. This means that a switchover from cold- to hot-leg injection cannot be used to dilute- boron in the RHR system. The second area of difference is that several valves may .be flooded following a LOCA. Once flooded, the valves may not work, and no credit is given for operation of flooded valves. The valve lineup on the Ginna high head injection system is set for cold-leg injection with power removed to prevent 'spurious operation of the flooded valves. This means that switchover from cold- to hot-leg injection cannot be used to prevent boron precipitation in the high head injection system.

To prevent boron precipitation, the Ginna plant uses simultaneous injection from the RHR and high head injection systems. The simultaneous injection takes place within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following the LOCA, and requires the pr imary system to be cooled to RHR conditions. However, even if the system is not cooled to RHR conditions, it is unlikely that boron precipitation would occur because the solubility is greater at higher temperatures. Furthermore, cooldown to RHR operating conditions will not be a problem with a large-break LOCA. The staff, therefore, concludes that the Ginna method for prevention of boron precipita-tion is acceptable.

9.4 Chemical and Volume Control S stem The chemical and volume control system (CVCS) takes water from the reactor coolant system (RCS) and passes it through a regenerative heat exchanger,'n orifice to reduce its pressure; and a nonregenerative heat exchanger before reducing its pressure further by the use of a pressure control valve. After filtering and cleanup, the water may be returned to the RCS by the use of the charging pumps, which increase the water pressure and pass it through the regenerative heat exchanger to either the hot or cold legs of the RCS or to the pressurizer auxiliary spray l,ine.

The isolation of the CVCS suction line is provided by a manually operated sole-noid valve in series with three parallel solenoid-operated valves. Each of these valves is operated from the control room and has valve position indicated.

None of the valves have interlocks to prevent opening or to close automatically if the pressure exceeds the design rating of the,low-pressure portions of the system.

The isolation of the CVCS discharge line is provided by a common discharge line check valve. and a branch check valve in each of the three branches downstream of the common check valve. Drain fittings on the discharge line upstream of each check valve can allow the valves to be tested. There is no position indication in the control room for the check valves. Solenoid isolation valves in each discharge line branch have position indication in the control room, but these valves have no inter locks to prevent system overpressurization.

The CVCS is not in compliance with current licensing requirements for isolation of high- and low-pressure systems contained in BTP EICSB 3 because the suction line solenoid-operated valves have no interlocks to prevent system overpres-surization, and the discharge line check valves have no position indication in the control room.

Ginna SER 9-4

Because of the severe consequences of a LOCA outside of containment and the lack of assurance that these isolation valves could be closed against signifi-cant flow under the resulting environmental conditions, the staff proposed that the CVCS suction valves be modified to satisfy the functional requirements of BTP EICSB 3 and that position indication be installed on the charging pump dis-charge valves.

The licensee responded to the staff's proposals in a letter dated June 23, 1981. In summary, the licensee argued that the isolation of the letdown system and -the failure of the relief valve (which is not tested) has been shown to be acceptable based on SEP Topic XV-16 and that the probability of failure of a charging pump and the two check valves in the discharge path presents an acceptable small risk without further design changes.

The staff has concluded that the basis for not requiring interlocks on the CVCS discharge valves is that the check valves in series with the positive displace-ment pumps satisfy the single-failure criterion as long as check valve position is known (or functionabi lity demonstrated) and the pump capacity is verified periodically. Gross failure of a positive displacement pump at the high-pres-sure/low-pressure interface is not within the scope of this topic because it is an extremely improbable event.

The results of a letdown line break outside containment (evaluated under SEP Topic XV-16) are the basis for not requiring, modifications in the letdown subsystem because the radiological consequences satisfy the regulatory limits of 10 CFR 100.

The charging pump discharge valves were evaluated further under SEP Topic VI-4.

By current criteria each of the charging pump discharge lines would requi re two containment isolation valves because the charging system is not required for postaccident function. However, during the SEP integrated assessment, back-ifitting was not recommended for these containment penetrations because:

(1) The piping system is designed to operate at 2,250 psi, significantly above containment design pressure.

(2) The piping is seismic Category I.

(3) The charging pumps are positive displacement pumps, and, therefore, leak-age back through the pumps is expected to be minimal.

The staff, therefor e, concludes that the CVCS is acceptable.

9' Ventilation S stems To ensure that the ventilation systems at Ginna have the capability to provide a safe environment for plant personnel and for engineered safety features, the staff reviewed the design and operation of these systems under SEP Topic IX-5.

The systems reviewed included the (1) control room air ventilation system (2) spent fuel pool area ventilation system (3) auxiliary building and radwaste area ventilation system (4) turbine area ventilation system (5) engineered safety feature ventilation system Ginna SER 9-5

The staff determined that the ventilation systems for the Ginna plant are in conformance with current criteria with the exception of the auxiliary building system, which has a potential backflow problem.

10 CFR 50 (GDC 60), as implemented by SRP Section 9.4.5, requires that the plant include a means to suitably control the release of radioactive materials in gaseous and liquid effluents. Current criteria require that the capability exist to direct ventilation air from areas of- low radioactivity to areas of progressively higher radioactivity. There are two scenarios that can possibly violate this requirement; both can occur when the main exhaust fans shut down when offsite power is not available and the plant is operating on emergency diesel power.

The first condition is one in which exhaust air, with a higher radioactivity potential, could leak into the intermediate building housing the controlled access area. With the main exhaust fans shut down, the positive pressure created on the input side of the high-efficiency particulate air (HEPA) filter could cause exhaust leakage into the intermediate building if there was insuf-ficient partial vacuum created by the plant vent stack.

The second possibility can occur under the same main exhaust fan shutdown con-ditions with the plant vent stack providing insufficient partial vacuum on the system. With four separate exhaust subsystems discharging to a common point at the HEPA filter input, the flow-pressure characteristic of the fans could be sufficiently mismatched to produce backflow through an operating fan (isolation dampers open) and thus introduce higher radioactive exhaust to an area of generally lower radioactivity potential.

In both cases the leakage would be to a controlled area with no increased releases to the environment. The limited PRA study found this issue to be a low contributor to risk. The staff, therefore, concluded during the SEP inte-grated assessment that modifications to the ventilation system are not necessary.

9.6 Fire Protection Following a fire at the Brown's Ferry Nuclear Station in March 1975, the NRC initiated an evaluation of the need for improving the fire protection programs at all licensed nuclear power plants. As part of this continuing evaluation, the NRC, in February 1976, published the report by a special review group entitled, "Recommendations Related to Browns Ferry Fire," NUREG-0050. This report recommended that improvements in the areas of fire prevention and fire control be made in most existing facilities and that consideration be given to design features that would increase the ability of nuclear facilities to with-stand fires without the loss of important functions. To implement the report's recommendations, the NRC initiated a program for reevaluation of the fire pro-tection programs at all licensed nuclear power stations and for a comprehensive review of all new licensee applications.

The NRC issued new guidelines for fire protection programs in nuclear power plants that reflect the recommendations in NUREG-0500. All licensees were requested to (j.) compare their fire protection programs with the new guidelines and (2) analyze the consequences of a postulated fire-in each plant area.

Ginna SER

The staff reviewed the licensee's analyses and visited the plant to examine 'the relationship of safety-related components, systems, and structures to both com-bustible materials and the associated fire detection and suppression systems.

The Safety Evaluation supporting Amendment No. 24 to Ginna's license (letter dated February 14, 1979) summarizes the results of the staff's review of the fire protection program at Ginna. In three supplements to the Fire Protection Safety Evaluation (letters dated December 17, 1980, and February 6 and June 22, 1981(a)), the staff resolved all but one of the open items remaining from the original review. The open item was related to the safe shutdown capability at Ginna following a potential fire.

In February 1981, the Fire Protection Rule (10 CFR 50.48 and Appendix R to 10 CFR 50) became effective. RG8E provided the staff with the descriptions of the various means used to achieve and maintain safe shutdown conditions without equipment or cabling in any fire area, and proposed modifications or alterna-tives required because of the unacceptable interactions caused by a fire. The staff's review of RG8E's safe shutdown capabilities is contained in the Safety Evaluation on this topic issued by letter dated April ll, 1983(a). The staff concluded that the proposed modifications will meet the requir'ements of Appendix R to 10 CFR 50, Sections III.G.3 and III-L, for those areas identified in the Safety Evaluation.

However, the licensee has proposed in conversations with the staff that an alternate system be used instead, of the currently approved system because of the high costs of the originally proposed modifications. The resolution of this proposal will be the result of future licensing actions.

Ginna SER 9-7

II a

I J,

10 STEAM AND POWER CONVERSION SYSTEM

10. 1 Steam and Feedwater S stem Steam from each of the two steam generators supplies the turbine, where the steam expands through the high-pres'sure turbine and then flows through reheaters and intercept valves to two, double-flow, low-pressure turbines, all in tandem (Figure 10. 1). Five stages of extraction are provided; two from the low-pres-sure turbine, 'one of which is the exhaust, and three from the low-pressure tur-bine. The feedwater heaters for the lowest three stages are. located in the condenser neck. .All feedwater heaters are horizontal, half-, size units (two trains), except those for the lowest two extraction stage points, which are of the duplex type. The feedwater system is the closed type with deaeration ~

accomplished in the condenser (Figures 10.2 and 10.3). The rejected heat is dissipated to Lake Ontario by the open cycle circulating water system..

Condensate is taken from the condenser hotwell through the condensate booster pumps and full-fl'ow demineralizers to the suction of the condensate pumps, through the hydrogen coolers, air ejectors, gland steam condenser, and low-pressure heaters to the suction of the.feedwater pumps. The feedwater pumps then, send feedwater through the high-pressure heaters to the steam generators.

Effluent from the high-pressure heater drains and effluent from the four reheater drains cascades to the second-stage extraction, low-pressure heater and then to a drain tank. The moisture separators also drain to this tank.

The heater drain pump discharges to the feedwater pump suction. Effluent from the three lower pressure heater drains cascades to the condenser.

The steam piping is des,igned to. ensure correct steam distribution and pressures to all steam-consuming equipment for all turbine loads. In addition, the steam and feedwater lines with their supports and structures from the stea'm generators to their respective isolation, valves are seismic Category I. A,failure of any main steam or feedwater line or malfunction of a valve installed therein or any consequential damage will not impair the reliability of the auxiliary feedwater system, render inoperative any engineered safety feature, initiate a loss-of-coolant condition, or cause failure of any other steam or feedwater line.

The main steam lines have four ASME Code-approved safety valves on each line, which provide pressure relief for the steam generators. There is also one power-operated relief valve on each line for long-term plant cooldown by atmos-pheric steam discharge .if the condenser steam dump is not available. Each steam line is equipped with a fast-closing isolation valve and a nonre'turn check valve. The isolation and nonreturn valves are located outside the containment.

These valves prevent reverse flow in the steam lines resulting from an upstream steam line break or isolate a downstream steam line break at the common header.

The eight main steam safety valves have a total combined rated .capability of 6,580',000 lb/hour. The total full-power steam flow is 6,260,000 lb/hour; therefore, eight main steam safety valves will be able to relieve the total steam flow if necessary.

Ginna SER 10-1

A steam flowmeter (flow Venturi tube) is provided in the line from each steam generator. The steam flow signals are used by the automatic feedwater flow control system. The flow Venturi tube also serves to limit the steam flow rate in the unlikely event of a steam line break outside the containment.

Steam pressure is measured upstream of the isolation and nonreturn valves.

Each isolation valve contains a free swinging disc, which is normally held up out of the main steam flow path by an air piston. The isolation valves are closed by either a containment isolation signal'r a steam line isolation signal. The isolation valves are designed to close in less than 5 seconds.

The feedwat'er lines are:equipped with a nonreturn check valve and an isolation valve in each line. The nonreturn valve is the boundary between seismic Cate-gory I and Category III fe'edwater piping and prevents 'the steam generator from blowing back through the feedwater lines if damage occurs to the Category III portion.

10.2 Turbine Disc Cracks Since 1979 the staff has known of the stress-corrosion problems in low-pressure rotor discs in Westinghouse turbines. The manufacturer has conducted a program to establish the methods and intervals for inservice inspection of the bore and keyway areas in these discs. Westinghouse has developed a method for predicting the rate of crack growth based on all of the cracks found to date in Westinghouse turbines, past history of similar turbine disc cracking, and results of labora-tory tests. This prediction method takes into account two main parameters, the yield strength of the disc, and the temperature of the disc at the bore area where the cracks of concern are occurring. The higher the yield strength of the material and the higher the temperature, the faster the rate of crack growth will be.

By letter dated August 28, 1981, the staff advised RG8E that the inspection schedules based on the recommendations of Westinghouse will provide an accept-ably high degree of assurance that discs would be inspected before cracks could grow to a size that could cause disc failure at speeds up to design speed. RG8E committed to the inspection schedules by letter dated September 16, 1981. The staff, therefore', considers this issue to be complete.

10.3 Secondar Water Chemistr The Ginna steam generator began operation in 1970 with coordinated phosphate secondary water chemistry control to reduce corrosion and continued on this regime until 1974. During this period the industry in general experienced early difficulties with phosphate-wastage corrosion and stress-corrosion cracking where sludge had accumulated on the tubesheet. These difficulties were attri-

. buted to problems in adequately controlling phosphate concentrations and to impurities carried into the steam generators by the feedwater.

Wastage corrosion was identified during inservice inspections performed at Ginna in the spring of 1974. Following a Westinghouse recommendation to its customers in a letter dated August 29, 1974, Ginna converted from phosphate to all-volatile treatment (AVT) secondary water chemistry control during a shutdown that began in November 1974. AVT consisted of the addition of hydrazine to the condensate water for the purposes of scavenging oxygen. At this time, blowdown Ginna SER 10-2

rates were increased from the maximum, of 16 gpm per steam generator to 64 gpm to reduce cation conductivity.

The licensee determined that to- maintain the new water chemistry specifications associated with AVT,- full-flow, deep-bed, condensate demineralizers along with blowdown heat recovery would'e required. The installation of the full-flow condensate demineralizers included (1) a regenerative system, (2) a waste, treatment system, (3) a condensate booster pump system, (4) modification of the existing blowdown heat recovery, and (5) a building addition to house these systems. The demineralizers are designed to treat 9,400 gpm of condensate at 100'-400 F and,425 psig on a continuous basis for the remaining life of the plant., The system has been in operation since 1978.,

The licensee reports that he has closely monitored condenser tube integrity and performed preventive repairs to damaged condenser tubing during shutdowns to provide assurance of feedwater purity. Blowdown rates were further increased to 70 gpm in February 1979.

The blowdown chemistry comparison in Table 5.3 demonstrates the continued improvement in bulk water chemistry over the last 8 years.

N 10.4 Auxiliar and Standb Auxiliar Feedwater S stems The auxiliary feedwater system and all three pumps are located in the inter-mediate building. There are two motor-driven pumps and one steam-driven pump.

These pumps are only used during startup and normal or emergency shutdown of the plant. They are susceptible to damage from the effects of breaks in the main steam and feedwater lines and the auxiliary steam and feedwater lines.

To ensure the heat removal capability for core cooling, the licensee proposed and later installed a standby auxiliary feedwater system adjacent to the aux-iliary building along the south wall. The standby auxiliary feedwater pump-house is a seismic Category I concrete structure supported by caissons.

The standby auxiliary feedwater system consists of two, independent 100K capacity subsystems in a new structure remote from high-energy lines. The dis-charge piping from the pumps is routed through the auxiliary building, enter s the containment through penetrations remote from the main steam and feedwater lines, and connects to the feedwater lines near each steam generator with check valves near the connection'o minimize the amount of line pressurized during normal plant operation.

The pumps take suction from the service water loops inside the auxiliary building, are motor driven from the engineered safety features buses, and are manually started from the control room in the event that the auxiliary feed-water pumps, which start automatically, are not operable. The analysis'per-formed by the licensee assumes that feedwater is not available for 10 minutes following the'orst-case line break. This is ample time for the control room operator to take action because alarms and indications are available in the control room to alert the operator to the lack of effective auxiliary feedwater flow and the standby pumps can be put into operation from the control room.

In the event of loss of o'ffsite power, the pumps would be powered by the diesel generators. The diesel generators have sufficient capacity for this additional 225-kW load. However, to prevent an overload of the feedbreakers tying the Ginna SER 10-3

diesel generators to the" buses, an interlock has been installed to prevent starting a standby pump when its associated auxiliary pump is running on the diesel generator.

I The staff concluded in 'the Safety Evaluation related to Amendment No. 29 to Ginna's license (letter dated August 4, 1979) that the modifications provide an acceptable backup to the auxiliary feedwater system for maintaining the plaht in a safe shutdown condition. Following the accident at TMI, the staff re-reviewed the auxiliary and standby auxiliary feedwater systems at Ginna under TNI Action Items II. E. 1. 1 and II. E. l. 2. The evaluations of these Action Items (lettei s dated January 29, 1981, June 16, 1982, and August 18, 1982) concluded that the Ginna auxiliary and standby auxiliary feedwater systems meet NRC's requirements.

Ginna SER 10-4

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SUPERINTENDENT TRAINING MANAGER ASSISTANT ASSISTANT TRAINING SUPERINTENDENT COORDINATOR (S)

INSTRUCTORS ADMINISTRATIVE HEALTH PHYSICS MAINTENANCE OP ERAT. IONS NUCLEAR TECHNICAL MANAGER AND CHEMISTRY MANAGER MANAGER (SRO) ASSURANCE MANAGER MANAGER MANAGER

, Figure 13.2 R. E. Ginna Nuclear Power Plant organization chart

HEALTH PHYSICS ADMINISTRATIVE AND CHEMISTRY MANAGER MANAGER COST CONTROL ADMINISTRATIVE OFFICE HEALTH COORDINATOR COMPUTER CHEMIST PHYSICISTS RADIOCHEMIST SUPERVISOR SYSTEMS ANALYST COMPUTER CENTRAL CHEMISTRY RADIATION CHEMISTRY TECHNICIANS RECORDS TECHNICIANS PROTECTION TECHNICIANS FOREMAN RADIATION PROTECTION TECHNICIANS I/SHIFT Figure 13.3 R. E. Ginna Nuclear Power Plant administrative and health physics and chemistry sections

MAINTENANCE NUCLEAR MANAGER ASSURANCE MANAGER MATERIALS OPERATIONAL QUALITY F IRE PROTECTION MAINTENANCE I 8 C COORDINATOR ASSESSMENT CONTROL AND SAFETY SUPERVISOR SUPERVISOR ENGINEER ENGINEER COORDINATOR I 8 C AND SHIFT TECHNICA QUALITY CONTRO STOCKROOM MAINTENANCE INSPECTION FOREMAN FOREMAN ELECTRIC ADVISOR FOREMAN I/SHIFT SUPERVISOR FITTERS, TECHNICIANS STOCKIKEEPERS MECHANICS. REPAIRMEN TECHNICIANS HANDYMEN ELECTRICIANS WHEN RCS AVERAGE TEMPERATURE i EIIP F Figure 13.4 R. E. Ginna Nuclear Power Plant maintenance and nuclear assurance sections

OPERATIONS TECHNICAL MANAGER (SRO) MANAGER OPPERATIONS RESULTS REACTOR TECHNICAL UPERVISOR (SRO AND TEST ENGINEER PROJECTS ENGINEER (S)

SUPERVISOR SUPERVISOR SHIFT TECHNICIANS TECHNICAL SUPERVISOR REPAIRMEN COMPUTER I/SHIFT (SRO) SYSTEMS ANALYST HEAD CONTROL OPERATOR 1/SHIFT (RO) COMPUTER TECHNICIANS CONTROL OPERATOR I/SHIFT (RO)

AUXILIARY OPERATORS 2/SHIFT Figure 13.5 R. E. Ginna Nuclear Power Plant operations and technical sections

14 INITIAL TEST PROGRAM The preoperational testing program and the operational and transient tests for operation up to 1,300 MWt were successfully completed and reported in the "Technical Supplement Accompanying Application To Increase Power, February 1971" (letter dated February 8, 1971). The staff reviewed and reported on these results in the Safety Evaluation issued by letter dated January 20, 1972, in conjunction with Amendment No. 2 authorizing the power increase.

A testing program for power escalation to 1,520 MWt was successfully completed in March 1972, and RG&E reported the results in a letter dated August 14, 1972.

The power escalation program consisted pf a number of tests and measurements at power levels of 1,300, 1,380, 1,455, and 1,520 MWt. At each of these power levels, incore flux maps, delta T measurements, containment vessel radiation surveys, and primary coolant activity measurements were perfor'med. Additional flux maps were obtained at 1,455 MWt to calibrate the axial offset monitoring.

The flux maps, delta T measurements, and the containment vessel radiation surveys all showed very good agreement with predictions. Primary coolant activity, however, increased more than was expected. This was attributed to the failure of a few more fuel rods during the power escalation program. The operation at 1,520 MWt was limited to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> just before a refueling outage during which the fuel assemblies that showed evidence of leaking rods were removed. A dis-cussion of the fuel densification problems is given in Section 4.6. 1 of this report.

Ginna SER 14-1

15 ACCIDENT ANALYSIS h

As part of SEP the staff reevaluated the ability of Ginna to withstand normal and abnormal transients and a broad spectrum of postulated accidents without undue hazard to the health and safety of the public. The results of these analyses are used to show conformance. with GDC 10 and 15.

During its review of the transients and accidents analyses of Section 15, the staff has considered GDC 21, 27, and 28 and Regulatory Guides 1.53 and 1. 105 as they apply to the events analyzed to ensure that the applicable requirements have been met.

For each event analyzed the worst operating conditions were assumed, and credit was taken for minimum engineered safeguards response. Parameters specific to individual events were conservatively selected.

Two types of events were analyzed:

(1) those incidents that might be expected to occur during the lifetime of the reactor (anticipated transients)

(2) those incidents not expected to occur that have potential to result in significant radioactive material release (accidents)

The following is a list of those events reviewed by the staff:

SEP, Number Ti tl e XV-1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of a Steam Generator Relief or Safety Valve XV-2 Spectrum of Steam System Piping Failures Inside and'Outside Contain-ment (PWR) j XV-3, Loss of External Load, Turbine Trip, Loss of Condenser Vacuum, Closure of Main Steam Isolation Valve (BWR), and Steam Pressure Regulator Failure (Closed)

XV-4 Loss of Nonemergency AC Power to the Station Auxiliaries XV-5 Loss of Normal Feedwater Flow XV-6 Feedwater System Pipe Breaks Inside and Outside Containment (PWR)

XV-7 Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break C XV-8 Control Rod Misoperation (System Malfunction or Operator Error)

Ginna SER 15-1

XV-9 Startup of an Inactive Loop or Recirculation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate XV-10 Chemical and Volume Control System Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant (PWR)

XV-12 Spectrum of Rod Ejection Accidents (PWR)

XV-14 Inadvertent Operation of Emergency Core Cooling System and Chemical and Volume Control System Malfunction That Increases Reactor Coolant Inventory XV-15 Inadvertent Opening of a PWR Pressurizer Safety/Relief Valve or a BWR Safety/Relief Valve XV-16 Radiological Consequences of Failure of Small 'Lines Carrying Primary Coolant Outside Containment XV-17 Radiological Consequence of Steam Generator Tube Failure (PWR)

XV-19 Loss-of-Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary XV-20 Radiological Consequences of Fuel-Damaging Accidents (Inside and Outside Containment)

The staff's reevaluation of the above events noted two modifications that were necessary.

Under SEP Topic XV-17, the radiological consequences of a steam generator tube failure were reviewed. This topic encompasses those design features that limit the release of radioactivity, including Technical Specifications associated with coolant activity concentrations. The initial topic review showed that no maximum coolant concentration for an iodine spike was set and no yearly limit was established for the number of such spikes. Following the January 25, 1982, steam generator tube rupture (SGTR) at Ginna, the licensee has, by Amendment No. 51 to the Technical Specifications (letter dated May 22, 1982), implemented limits on coolant activity as required by the staff.

Under SEP Topic XV-19, the radiological consequences of a design-basis LOCA were reviewed. This review was to ensure that the radiological consequences from containment leakage and leakage from engineered safety features outside containment are within the exposure guideline values of 10 CFR 100. The initial topic review identified a 2-minute delay timer on the sodium hydroxide addition to containment spray that is contrary to current regulatory criteria. The licensee has set the sodium hydroxide addition valves delay timer to 1 second and instructed operators not to override sodium hydroxide addition in the event of a containment spray activation. This modification was completed during the spring 1982 refueling, and the staff finds it acceptable.

On the basis of the SEP topic safety evaluations of the transients l'isted above, the staff concludes that the analyses demonstrate the operation of the plant Ginna SER 15-2

will not result in any violation of fuel design or reactor coolant pressure boundary design limits, conform with GDC 10 and 15, and are, therefore, accept-able. Additionally, the staff concludes that the licensee has provided adequate protection systems to mitigate accidents in compliance with GDC 10, 15, and 20 and 10 CFR 50 and 100.

Also as a result of the SGTR, the staff required in NUREG-0916 that the licensee perform a number of modifications and analyses to confirm the staff's conclu-sion that there is reasonable assurance that the plant can be operated without endangering the health and safety of the public. These modifications and anal-yses have been performed and are currently under staff review. The staff's evaluations of these items will be reported as routine operating reactor licensing actions.

Ginna SER 15-3

16 TECHNICAL SPECIFICATIONS The Technical Specifications in a license define certain features, character-istics, and conditions governing the operation of a facility that cannot be changed without prior approval of the staff. The current Technical Specifica-tions for Ginna are part of the provisional operating license and will be made part of the full-term operating license. Included are sections covering defini-tions, safety limits, limiting safety settings, limiting conditions for opera-tions, surveillance requirements, design features, and administrative controls.

In the course of the staff's review of the individual SEP topics, the Ginna Tech-nical Specifications were compared with the Standard Technical Specifications for deviations. Where significant differences existed, they were identified and the staff considered them for upgrading. Table 4. 1 of the IPSAR (NUREG-0821) and Table 3. 1 of IPSAR Supplement 1 identify the Technical Specifications that the staff has identified as requiring upgrading. The other sections of the Tech-nical Specifications are reviewed only to the extent that reloads, license amend-ments, or generic problems require.

Ginna SER 16-1

17 EQUALITY ASSURANCE The quality assurance organization is responsible for ensuring that procedures and instructions comply with complete and adequate quality assurance, require-ments. In addition, quality assurance personnel should provide sufficient reviews, inspections, and audits to verify the effective implementation of. the entire quality assurance program.

The licensee has structured his quality assurance program for the operation phase so that it is in accordance with Appendix 8 to 10 CFR 50 and complies with the regulatory positions given in quality assurance-related regulatory guides and with the requirements of Americal National Standards Institute (ANSI) N45.2. 12. The quality assurance program is implemented by means of written policies, procedures, and instructions. These documents result in con-trol of quality-related activities involving safety-related items in accordance with the requirements of Appendix B to 10 CFR 50 and with applicable regula-tions, codes, and standards.

The licensee's quality assurance program requires that implementing documenta-tion encompass detailed controls for (1) personnel indoctrination and .training; (2) translating codes, standards, regulatory requirements, technical specifica-tions, engineering requirements, and process requirements into drawings, specifi-cations, procedures, and instructions; (3) developing, reviewing, and approving procurement documents, including changes; (4) prescribing all quality-related activities by documented instructions, procedures, drawings, and specifications; (5) issuing and distributing approved documents; (6) purchasing items and ser-vices; (7) identi,fying materials, parts, and components; (8) performing special processes; (9) inspecting and/or testing materials, equipment, processes, or services; (10) calibrating and maintaining measuring equipment; (ll) handling, storing, and shipping items; (12) identifying the inspection, test, and operat-ing status of items; (13) identifying and disposing of nonconforming items; (14) correcting conditions adverse to quality; (15) preparing and maintaining quality assurance records; and (16) auditing activities that affect quality.

guality is verified through checking, review, surveillance, inspection, testing, and audit of quality-related activities. The quality assurance program requires that quality verification be performed by individuals who are not directly responsible for performing the quality-related activities. Inspections are performed by qualified personnel in accordance with procedures, instructions, and checklists approved by the quality assurance organization.

The quality assurance organization is responsible for the establishment and implementation of the audit program. Audits are performed in accordance with preestablished written checklists by qualified personnel not having direct responsibilities in the areas being audited. Audits are performed to evaluate all aspects of the quality assurance program, including the effectiveness of the quality assurance program implementation.

The quality assurance program requires the review of the audit results by the person having responsibility in the area audited and corrective action where Ginna SER 17-1

necessary. Continued deficiencies, or failure to implement corrective action, will be reported in writing by the quality assurance organization to the appro-priate management within the Rochester Gas and Electric Corporation. Followup audits are performed to determine that nonconformance and deficiencies are effectively corrected and that the corrective action precludes repetitive occurrences. Audit reports, which indicate performance trends and the effective-ness of the quality assurance program, are prepared and issued to responsible management for, review and assessment; On the basis of its review and evaluation of the quality assurance program described in Revision 7 to Supplement IV to the "Technical Supplement Accom-panying the Application for a Full-Term Operating License for the R. E. Ginna Nuclear. Power Plant" (letter dated November 21, 1980), as supplemented by the licensee's letter of April 23, 1981(b); the staff concluded in the Safety Evaluation transmitted by letter dated June 22, 1981(b), that:

W (1) The organizations and individuals performing quality assurance functions have the required independence and authority to effectively carry out the quality assurance program without undue influence from those directly responsible for cost and schedules.

(2) The quality assurance program describes requirements, procedures, and controls that, when properly implemented, comply with the requirements of Appendix B to 10 CFR 50 and with applicable quality-assurance-related regulatory guides and ANSI standards.

(3) The quality assurance program covers activities affecting structures, systems, and components identified in the Final Facility Description and Safety Analysis Report (letter dated January 24, 1968). Items designated as seismic Class I on'he list given in the Safety Analysis Report are subject to the pertinent provisions of the quality assurance program. An evaluation of the acceptability of the list of structures, systems, and components is included in the scope of review of SEP Topic III-1, "Clas-sification of Systems and Equipment."

(4) Updating regulatory guide commitments to include Regulatory Guide 1. 146 and Revision 1 of Regulatory Guide 1. 58 will be addressed separately in response to NRC's Generic Letter 81-01.

Accordingly, the staff concluded that the licensee's description of the quality assurance program is in compliance with applicable NRC regulations.

In accordance with 10 CFR 50.54, the licensee submitted on June 6, 1983, his revised quality assurance program for staff review. The revised program commits to Regulatory Guides 1.146 and 1.58, Revision 1, as requested by Generic Letter 81-01 (dated May 14, 1981). The staff approved the revision by letter dated August 10, 1983. The effectiveness of the implementation of the Ginna quality assurance program will continue to be the subject= of routine NRC inspections.

Ginna SER 17-2

18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Ginna application for a full-term operating license is being reviewed by the Advisory Committee on Reactor Safeguards. The NRC staff will issue a supplement to this Safety Evaluation Report after the Committee report to the Commission is available. The supplement will append a copy of the Committee's report, will address comments made by the Committee, and will describe steps taken by the NRC staff to resolve any issues raised as a result of the Committee's review.

Ginna SER 18-1

19 COMMON DEFENSE AND SECURITY Rochester Gas and Electric Corporation (RG&E) is not owned, dominated, or con-trolled by an alien, a foreign corporation, or a foreign government. The activ-ities that will continue to be conducted do not involve any restricted data, but RG8E has agreed to safeguard any such data that might become involved in accordance with the requirements of 10 CFR 50. The licensee will continue to rely upon obtaining fuel as it is needed from sources of supply available for civilian purposes, so that no diversion of special nuclear material from mili-tary purposes is involved. For these reasons, and in the absence of any infor-mation to the contrary, the staff has found that issuance of the full-term operating license will not be inimical to the common defense and security.

Ginna SER 19-1

20 FINANCIAL QUALIFICATIONS On March 31, 1982, the NRC published in the Federal Re ister (47 FR 13750) amendments to its regulations that entirely eliminate the review relating to the financial qualifications of applicants for construction permits and oper-ating licenses. Because these amendments were effective immediately, there will be no further review of the financial qualifications of the Rochester Gas and Electric Corporation.

Ginna SER 20-1

21 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS Pursuant to the financial"'protection and indemnification provisions of the Atomic Energy Act of 1954, as amended (Section 170 and related sections), the Commission has issued regulations in 10 CFR 140. These regulations set forth the Commission's requirements with regard to proof of financial protection by, and indemnification of, licenses for facilities such as power reactors under 10 CFR 50.

Under the Commission's regulations, 10 CFR 140, a license authorizing the operation of a reactor may not be issued until proof of financial protection in the amount required for such operation has been furnished, and an indemnity agreement covering such operation has been executed. The amount of financial protection that must be maintained for the Ginna plant (which has a rated capac-ity in excess of 100,000 electrical kilowatts) is the maximum amount available from private sources (that is, the combined capacity of the two nuclear liabi l-ity insurance pools; this amount is currently $ 160 million).

The NRC and RG8E entered into Indemnity Agreement No. B-38 on March 21, 1969.

Therefore, the staff concludes that the licensee complies with the provisions of 10 CFR 140 applicable to operating licenses, including those that relate to proof of financial protection in the requisite amount and to execution of an appropriate indemnity agreement with the Commission.

Ginna SER 21-1

22 CONCLUSIONS On the basis of its evaluation of the application as set forth above, the staff has determined that (1) The application for a full-term operating license (FTOL) for the R. E.

Ginna Nuclear Power Plant filed by RG8E, dated August 15, 1972, as supple-mented and as revised, complies with the requirements of the Atomic Energy Act of 1954, as amended (Act), and the Commission's regulations set forth in 10 CFR Chapter I, except as duly exempted therefrom.

(2) Construction of the R. E. Ginna Nuclear Power Plant has been completed in conformity with Construction Permit No. CPPR-19, as amended, the applica-tion as amended, the provisions of the Act, and the rules and regulations of the Commission.

(3) The provisions of Provisional Operating License No. DPR-18 have been met.

(4) The facility will operate in conformity with the FTOL application as amended, the provisions of the Act, and the rules and regulations of the Commission.

(5) There is reasonable assurance (a) that the activities authorized by the FTOL can be conducted without endangering the health and safety of the public and (b) that such activities will be conducted in compliance with regulations of the Commission set forth in 10 CFR Chapter I.

(6) The licensee is technically qualified to engage in the activities autho-rized by the FTOL, in accordance with the regulations of the Commission set forth in 10 CFR Chapter I.

(7) The issuance of the FTOL will not be inimical to the common defense and security or to the health and safety of the public.

Ginna SER 22-1

rl

> ~

APPENDIX A REFERENCES Algermissen, S. T., et al. (USGS), Open File Report 82-1033, "Probabilistic Estimates of Maximum Acceleration and Velocity in Rock in the Contiguous United States," 1982.

Atomic Safety and Licensing Appeal Board, ALAB-444, 6 NRC,760, "Gulf States Utilities Co., River Bend, Units 1 and 2," Nov. 23, 1977.

I Code of Federal Re ulations, Title 10 (10 CFR), "Energy,"'.S. Government, Printing Office, Mashington, D. C. (includes General Design Criteria).

Dames and Moore, "Geologic and Geophysical Investigations Ginna. Site, Ontario, New York, for Rochester Gas and Electric Corp.", 1974.

---, "Nine Mile Point Nuclear Station, Unit 2 Geologic Investigations, for Niagara Mohawk Power Corporation," 1978.

Environmental Science Services Administration (ESSA), "1966 Report on the Seismicity of the Rochester, New York Aiea," forwarded in letter dated Feb. 16, 1966, to H. L. Price (USAEC) from J. C. Tilson (ESSA).

Federal Re ister (47 FR 13750), "Final Rule - Nuclear Power Plant - Elimination of Review of Financial gualifications of Electric Utilities in Licensing Hear-ings - 10 CFR Parts 2 and 50," U.S. Nuclear Regulatory Commission, Mar. 31, 1982.

Fenneman, N. M., Ph sio ra h of Eastern United States, McGraw Hill Book Co.,

New York, 1938.

GAI, Inc., "Structural Integrity Test of Reactor Containment Structure - R. E.

Ginna Nuclear Power Station," GAI Report ¹1720, Oct. 3, 1969.

Lawrence Livermo're Laboratory, "Technical Evaluation of the Adequacy of Station Electric Distribution System Voltages for the R. E. Ginna Nuclear Power Station, Unit 1," Nov. 5, 1981.

Letter, Nov. 1, 1965, from LeBoeuf, Lamb, and Leiby (on behalf of RG8E) to U. S. AEC, Attn: Division of Licensing and Regulation,

Subject:

Application for Licenses, Notarized Oct. 28, 1965.

---, Apr. 25, 1966, from R. Doan (AEC), to F. Drake (RG&E),

Subject:

Transmit-tal of Provisional Construction Permit.

---, Jan. 24, 1968, from LeBoeuf, Lamb; and Leiby (on behalf of RG8E) to U.S.

AEC, Attn: Division of Licensing and Regulation,

Subject:

Final Facility Description and Safety Analysis Report, Notarized Jan. 18, 1968 (Amendment 6 to Application for Licenses).

Ginna SER A-1

---, Mar. 21, 1969, from P. Morris (NRC) to LeBoeuf, Lamb, Leiby and MacRae (Attorneys for RG&E),

Subject:

Indemnity Agreement No. B-38.

---, June 24, 1969, from P. Morris (AEC) to F. Drake (RG&E),

Subject:

Proposed Issuance of Provisional Operating License -,- Technical Specifications and Safety Analysis Report Included.

---, Sept. 25, 1969, from P. Morris (AEC) to F. Drake (RG&E),

Subject:

Trans-mittal of Provisional Operating License No. DPR-18 which became effective Sept. 19, 1969. ,

---, Feb. 8, 1971, from LeBoeuf, Lamb, Leiby, and MacRae (on behalf of RG&E) to P. Norris (NRC),

Subject:

Petition Requesting Amendment of License and Extension of Expiration Date of Provisional Operating License.

---, Jan. 20, 1972, from D. Skovholt (NRC) to E. Nelson (RG&E),

Subject:

Proposed Issuance of Amendment to Provisional Operating License No. DPR-18.

---, Aug. 14, 1972, from K. Amish (RG&E) to E. Bloch (NRC),

Subject:

Power Escalation to 1520 MWT, Mar. 1972. ql

---, Aug. 15, 1972, from LeBoeuf, Lamb, Leiby, and MacRae (Attorneys for RG&E) to J. 0 Leary (AEC),

Subject:

=Application To Convert Provisional Operating License.

---, Nov. 20, 1972, from A. Giambusso (AEC) to E. Nelson (RG&E) and other operating reactor licensees identified on listing attached to letter, trans-mitting "Technical Report on Densification of Light Water Reactor Fuels," dated Nov. 14, 1972.

---, Dec. 18, 1972, from K. Amish (RG&E) to J. O'Reilly (AEC),

Subject:

AEC Bulletin 72-3, re: Electric Type Valve Operators.

---, June 29, 1973, from D. Skovholt (NRC) to E. Nelson (RG&E),

Subject:

Change No. 9 to License DPR-18.

---, Aug. 29, 1974, from E. R. Duhn (Westinghouse) to G. E. Platt (RG&E),

Subject:

Steam Generator Chemistry Control.

---, May 14, 1975, from R. Purple (NRC), to L. White (RG&E),

Subject:

Amend-ment No. 7 to Provisional Operating License No. DPR-18.

-, Mar. 30, 1976, from R. Purple (NRC), to L. White (RG&E),

Subject:

Amend-ment 10 to Provisional Operating License No. DPR-18.

---, Nov. 15, 1976, from A. Schwencer (NRC), to L. White (RG&E),

Subject:

Amend-ment 11 to Provisional Operating License No. DPR-18.

---, May 17, 1977, from V. Stello (NRC), to L. White (RG&E),

Subject:

Amendment No. 13 to Provisional Operating License No. DPR-18.

---, June 3, 1977, from A. Schwencer (NRC), to L. White (RG&E),

Subject:

Degraded Grid Voltage.

Ginna SER A-2

---, Mar. 28, 1978, from D. Ziemann (NRC) to L. White (RG&E),

Subject:

Amend-ment No. 17 to Provisional Operating, License No. DPR-18.

-, Hay 1, 1978, from D. Ziemann (NRC) to L. White (RG&E),

Subject:

Amend-ment 19 to Provisional Operating License No. DPR-18.

---, Jan. 17, 1979, from D. Ziemann (NRC) to L. White (RG&E),

Subject:

Amend-ment 22 to Provisional Operating Li'cense No. DPR-18.'--,

Feb. 14, .1979, from D. Ziemann (NRC) to L. White (RG&E),

Subject:

Amend-ment 24 to Provisional Operating License No. DPR-18.

---, Apr. 18, 1979, from D. Ziemann (NRC) to L. White (RG&E),

Subject:

Amend-ment 26 to Provisional Operating License No. DPR-18.

---, Aug. 4, 1979, from D. Ziemann (NRC) to L. White (RG&E),

Subject:

Amend-ment 29 to Provisional Operating License No. DPR-18.

---, Feb. 19, 1980, from G. Lainas (NRC) to A. Schwencer (NRC),

Subject:

Electrical Equipment Environmental qualification With Attachments Containing DOR Guidelines.

---, Apr. 15, 1980, from D. Ziemann (NRC) to L. White (RG&E),

Subject:

Amend-ment 32 to Provisional Operating License No. DPR-18.

---, Nov. 21, 1980, from LeBoeuf, Lamb, Leiby, and MacRae (on behalf of RG&E) to H. Denton (NRC),

Subject:

Revision 7 to Supplement IY to the Technical Supplement Accompanying the Application for.a Full-Term Operating License for the R. E. Ginna Nuclear Power Plant.

---, Dec. 17, 1980, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Ginna (Supplement 1).

Fire Protection

---, Jan. 29, 1981, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E)

Subject:

Auxiliary Feedwater System.

---, Feb. 6, 1981, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Fire Protection - Ginna (Supplement 2).

---, Feb. 10, 1981, from D. M. Crutchfield (NRC) to J. E. Haier (RG&E),

Subject:

Amendment 35 to Provisional Operating License No. DPR-18.

---, Mar. 11,. 1981, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Amendment 37 to Provisional Operating License No. DPR-18.

---, Mar. 26, 1981, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Amendment 38 to Provisional Operating License No. DPR-18.

---, Apr. 23, 1981(a), from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Amendment 41 to Provisional Operating License No. DPR-18.

---,,Apr. 23, 1981(b), from J. E. Maier (RG&E) to D. H. Crutchfield (NRC),

Subject:

equality Assurance Program, R. E. Ginna Nuclear Power Plant.

Ginna SER A-3

---, May 4, 1981, from D. G. Eisenhut (NRC) to All Licensees of Operating Plants and Holders of Construction Permits,

Subject:

gualification of Inspec-tion, Examination, and Testing and Audit Personnel (Generic Letter 81-01).

---, May 6, 1981, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Completion of Appendix J Review.

---, May 26, 1981, from D. M. 'Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Inservice Testing - Valves.

---, June 8, 1981, from D. M. Crutchfield (NRC) to All SEP Owners,

Subject:

Site Specific Ground Response Spectra For SEP Plants Located in the Eastern United States.

---, June 22, 1981(a), from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Fire Protection - Ginna (Supplement 3).

---, June 22, 1981(b), from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Ginna guality Assurance Program.

---, June 23; 1981, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topics V-10.B, V-ll.A, V-ll.B, VI-7.C.1, VII-3, and VIII-2, R. E. Ginna Nuclear Power Plant.

---, July 14, 1981, from J. E. Maier (RG&E), to D. M. Crutchfield (NRC), Sub-ject: SEP Topic VI-7.e.l.

l

---, Aug. 28, 1981, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Turbine Disc Cracking (Ginna).

---, Sept. 16, 1981, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC), Sub-ject: Turbine Disc Cracking.

---, Nov. 3, 1981, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Spent Fuel Pool Cooling System.

---, Nov. 30, 1981, from'D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Amendment 16 to Provisional Operating License No. DPR-18.

---, Apr. 29, 1982, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Structural Re-evaluation.

---, May 20, 1982, from G. Smith (NRC) to J. E. Maier (RG&E),

Subject:

.Emergency Preparedness Appraisal 50-244/81-22.

---, May 22, 1982, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Restart of R. E. Ginna Nuclear Power Plant.

---, May 26, 1982, from R. Grimm (FEMA) to B. Grimes (NRC),

Subject:

Evaluation of Offsite Radiological Emergency Preparedness for the Ginna Nuclear Power Station Exercise.

-, June 16, 1982, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Auxiliary Feedwater System Evaluation, NUREG-0737 Item II. E. 1. 1 - Ginna.

Ginna SER

---,,Aug. 18, 1982, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Auxiliary Feedwater System Automatic Initiation and Flow Indication (TMI Action Plan Item II.E.1.2).

I.

Sept. 28, 1982, from J. E. Maier (RG&E) to H. Denton (NRC),

Subject:

Request - Plant Staff Reorganization. 'mendment

---, Nov. 18, 1982, from J. Devine (USGS) to R. Jackson (NRC),

Subject:

1886 '*

Charleston Earthquake.

P p f 1

Dec. 8, 1982, from J. E. Maier (RG&E),to D. M. Crutchfield (NRC),

Subject:

Analysis of Reactor Vessel Surveillance Capsule T.

---, Jan. 11, 1983, from.J. E. Maier. (RG&E) to D. M. Crutchfield'(NRC),'Sub-ject: Management Organization Chart.

I

---, Feb. 17, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

ECCS, Upper Plenum Injection.

---, Mar. 22,, 1983, from J.. E. Maier (RG&E).to D. M. Crutchfield.(NRC),

Subject:

SEP Topic III-5.B, Pipe Break Outside Containment.

---,, Apr. 1, 1983, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Steam Generator Tube Sleeving.Repair Program.,

p

---, Apr. 11, 1983, from D. M. Crutchfield (NRC) to J.

S E. Maier (RG&E)," Sub-ject: Control Room Habitability (TMI Action Plan Item fl III.D.3.4).

II

---, Apr. 11, 1983(a), from D. M. Crutchfield (NRC) to J: E. Maier. (RG&E),

Subject:

Fire Protection Rule - 10 CFR 50.48(c)(5) - Alternate Safe Shutdown-Section III.G of Appendix R to 10 CFR .50 - (SEP Topic IX-6).

---, Apr. 11, 1983(b), from J. E. Maier (RG&E) to D. MD Crutchfield (NRC),

Subject:

SEP, Topics III-1, III-6 and IX-3.

---, Apr. 12, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topic III-5.A, High Energy Line Break Inside Containment.

---, Apr. 22, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC)',

Subject:

Structural Reanalysis Program, SEP Topics II-2.A, III-2, III-4.A, and III-7.B.

---, Apr. 28, 1983(a), from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topics II-2.A, III-2, III-4.A, and III-7.B, "Structural Reanalysis Program" - Block Mails.

---, Apr. 28, 1983(b), from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topics III-7.8, Load Combinations (Containment Liner Analysis).

---, May 19, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Structural Reanalysis Program, SEP Topics II-2.A, III-2, III-4.A, and III-7.B-R. E. Ginna Nuclear Power Plant, Docket No. 50-244.

---, May 20, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topic II-3.B, Deer Creek Flooding.

Ginna SER A-5

--.-, May 27, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topic III-7.B, Design Codes, Design Cr'iteria and Load Combinations.

---, June 6, 1983, from J. E. Maier (RG&E) to J. Allan (NRC),

Subject:

Sub-mittal of equality Assurance Program. h k

June 13, 1983, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Steam Generator Sl eeving Tubesheet -Sl eeving.

f June 16, 1983(a), from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topic III-5.B, Pipe Break Outside Containment.

---, June 16, 1983(b), from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topic III-6, Seismic qualification of Tanks.

---, June 17, 1983, from F. Miraglia (NRC) to J. E. Maier (RG&E),

Subject:

Environmental Evaluation.

---, July 8, 1983, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

IPSAR Section 4.25.3, Flooding Due to Failure of Tanks R. E. Ginna Nuclear Power Plant.

---, July 8, 1983(a), from D. G."Eisenhut (NRC) to All Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Per-mits,

Subject:

Required Actions Based on Generic Implications of Salem ATMS Events (Generic Letter 83-28).

---, July 15, 1983, from R. Starosteck (NRC) to J. E. Maier (RG&E),

Subject:

Systematic Assessment of Licensee Performance (SALP).

---, July 20, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

SEP Topic III-5.B, Pipe Break Outside Containment (IPSAR Section 3.3. 1. 1).

---, July 25, 1983, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

Modifications to Steam Generator Tube Sleeving Process Necessitated by Tube L'ockup.

---, Aug. 1, 1983, from.D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

IPSAR Section 4.17, Containment Liner Insulation - R. E. Ginna Nuclear Power Plant.

---, Aug. 10, 1983, from T. Martin (NRC) to J. E. Maier (RG&E),

Subject:

10 CFR 50.54 gA Program Description Change Review for R. E. Ginna Nuclear Power Plant.

---, Aug. 16, 1983, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

IPSAR Section 3.3. 1. 1, Pipe Break Outside Containment - R. E. Ginna Nuclear Power Plant.

---, Aug. 22, 1983, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E),

Subject:

R. E. Ginna Nuclear Power Plant, IPSAR Section 4.8, Mind and Tornado Loadings; Sectidn 4. 11, Tornado Missiles; Section 4. 17. 1, Design Codes, Design Criteria and Load Combinations.

Ginna SER A-6

---, Sept. 28, 1983, from D. M. Crutchfield (NRC) to J. E. Maier (RGLE), Sub-ject: Amendment 57 to Provisional Operating License No:'PR-18.

r Memorandum, Aug. 19, .1981, from Acting Director for Radiological Emergency Preparedness Division (FEMA) to Director, Division of Emergency Preparedness (NRC),

Subject:

Status of Off-Site'lanning in 'New York State.

Power Authority State of New York, "James A. Fitzpatrick Nuclear Power Plant Final Safety Analysis Report," 1972.

Sbar, M. L., and L. R. Sykes, "Contemporary Compressive Stress and Seismicity in Eastern North America: An'Example of Intra-Plate Tectonics," Geol. Soc.

Amer. Bull., 84: 1861-1882, 1973.

URS/John A. Blume and Associates, Engineers, URS/JAB 8050, "Shaking-Table Testing for Seismic Evaluation of Electrical Raceway Systems," Apr. 1983.

---, URS/JAB 8050, "Analytical Techniques, Models,'nd Seismic Evaluation of Electrical Raceway Systems," Aug. 1983.

il U.S. Army Corps of Engineers, EM 1110-2-1411, "Standard Project Flood Deter-mination Procedure," Mar. 1965.

Atomic Energy Commission (AEC), "Draft Environmental Statement Related to

'.S.

the Operation of R. E. Ginna Nuclear Power Plant, Unit 1," Apr. 1973.

---, "Final Environmental Statement Related to the Operation of R. E. Ginna Nuclear Power Plant Unit 1," Dec. 1973.

U.S. Geological Survey (USGS), '-'Geology arid Hydrology of the Proposed Brookwood Nuclear Station No. 1 Site, Wayne County, New York," forwarded in letter dated Feb. 28, 1966, to H. L. Price (USAEC) from the Acting Director (USGS).

U. S. Nuclear Regulatory Commission, NUREG-0050, "Recommendations Related to Browns Ferry Fire," Feb. 1975.

---, NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, A Guidance Manual for Users of-Standard Technical Specifications," Oct. 1978.

---, NUREG-0138, "Staff Discussion of Fifteen Technical Issues Listed in Attachment to November 3, 1976 Memorandum From Director, NRR, to NRR Staff,"

Nov. 1976.

---, NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors," Rev; 2, July 1979; Rev. 3, Aug. 1980; Rev. 4, Nov. 1981.

---, NUREG-0472, "Radiological Effluent Technical Specifications for Pressurized Water Reactors," Rev. 3, Sept. 1982.

---, NUREG-0569, "Evaluation of the Integrity of SEP Reactor Vessels," Oct. 1979.

Ginna SER A-7

---, NUREG-0654,, "Criteri a for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,"

Rev. 1, Nov. 1980.

---, NUREG-0737, "Clarification of TMI Action Plan Requirements: Requirements for Emergency Response Capabi 1-ity," Nov. -1980; Suppl. 1, Jan. 1983.

---, NUREG-0800 (formerly NUREG-75/087), "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants. LWR Edition," July 1981 (includes Branch Technical Positions).

---, NUREG-0821, "Integrated Plant Safety Assessment - Systematic Evaluation Program - R. E. Ginna Nuclear Power Plant," Dec. 1982; Suppl. 1, Aug. 1983.

---, NUREG-0886, "Steam -Generator Tube Experience," Feb. 1982.

---, NUREG-0909, "NRC Report on the January 25, 1982 Steam Generator Tube Rupture at R. E. Ginna Nuclear Power Plant," Apr. 1982.

---, NUREG-0916, "Safety Evaluation Report Related to the Restart of R. E.

Ginna Nuclear Power Plant,," May 1982.,

---, NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant," Vol. 1, Apr. 1983; Vol. 2, Aug. 1983.

---, NUREG/CR-1464, "Review of Nuclear Power Plant Offsite Source Reliability and Related Recommended Changes to the NRC Rules and Regulations," May 1980.

---, NUREG/CR-1821, "Seismic Review of the Robert E. Ginna Nuclear Power Plant as Part of the Systematic Eval.uation Program," Dec. 1980.

---, Regulatory Guide (RG) 1. 14, "Reactor Coolant Pump Flywheel Integrity,"

Rev. 1 for comment, Aug. 1975.

---, RG 1.26, "guality Group Classifications and Standards for Water-, Steam-,

and Radioactive-Waste-Containing Components of Nuclear Power Plants."

---, RG 1.33, "gA Program Requirements (Operation)," Rev. 2, Feb. 1978.

---, RG 1.35, "Inservice Inspection of Ungrouted Tendons in Prestressed Con-crete Containment Structures," Rev. 2, Jan. 1976.

---, RG 1.45, "Reactor Coolant Pressure Boundary Leakage Detection System,"

May 1973.

---, RG 1.53, "Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems."

---, RG r

1.58, "gualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel," Rev. 1, Sept. 1980.

---, RG 1.76, "Design Basis Tornado for Nuclear Power Plants," Apr. 1974.

Ginna SER A-8

---, RG 1.83, "Inservice Inspection of PMR Steam Generator Tubes," Rev. 1, July 1975.

---, RG 1.91, "Evaluations of Explosions Postulated To Occur on Transportation Route Near Nuclear Power Plants," Rev. 1, Feb. 1978.

---, RG 1.97, "Instrumhntation for Light-Mater-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident."

---, RG 1. 105, "Instrument Setpoints."

---, RG 1. 117, "Tornado Design Classification," Rev. 1, Apr. 1978.

---, RG 1. 127, "Inspection of Mater-Control Structures Associated With Nuclear Power Plants," Rev. 1, Mar. 1978.

---, RG 1. 133, "Loose-Part Detection Program for Primary, Systems of Light Water Cooled Reactors," Rev. 1, May 1981.

---, RG 1. 139, "Guide for RHR To Achieve and Maintain Cold Shutdown," Rev. 1, Sept. 1979.

---, RG 1.:146, "qualification of gA Program Audit Personnel for Nuclear Power Plants," Aug. 1980.

---, RG 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable,"

Rey. 3, June 1978.

---, SECY-76-545, "The Systematic Evaluation of Operating Nuclear Power Plants,"

Nov. 12, 1976.

---, SECY-77-561, "Systematic Evaluation of Operating Reactors - Phases I and II," Oct. 26, 1977.

U.S. Atomic Energy Commission, Directorate of Regulator'y Operations, Region 1, AEC Bulletin 72-3, "Malfunction of Electric Type Valve Operators," Dec. 1, 1972.

U.S. Nuclear Regulatory Commission, Office of Inspection and Enforceme'nt,.

IE Bulletin 79-02, "Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts," Rev. 2, Nov. 8, 1979.

---, IE Bulletin 79-14, "Seismic Analysis for As-Built Safety-Related Piping Systems," July 2, 1979.

Westinghouse Topical Report WCAP 10086, "Analysis of Capsule T From Rochester Gas and Electric Corporation, R. E. Ginna Nuclear Plant, Reactor Vessel Radiation Surveillance Program," Mar. 30, 1982.

Ginna SER A-9

APPENDIX 8 THREE MILE ISLAND - LESSONS LEARNED REQUIREMENTS The accident at Three Mile Island Unit 2 (TMI-2) resulted in requirements that were developed from the recommendations of several groups established to inves-tigate the accident. These groups include the Congress, the General Accounting Office, the President's Commission on the Accident at Three Mile Island, the NRC Special Inquiry Group, the NRC Advisory Committee on Reactor Safeguards, the Lessons Learned Task Force, the Bulletins and Orders Task Force of the'NRC Office of the Nuclear Reactor Regulation, the Special Review Group of the NRC Office of Inspection and Enforcement, the NRC Siting Task Force and Emergency Preparedness Task Force, an) the NRC Offices of Standards Development and Nuclear Regulatory Research. NUREG-0660, entitled "NRC Action Plan Developed as a Result of the TMI-2 Accident" (referred to as Action Plan), was developed to provide a comprehensive and integrated plan for the actions now judged neces-sary by NRC to correct or improve the regulation and operation of nuclear facil-ities. The Action Plan was based on the experience from the TMI-2 accident and the recommendations of the investigating groups.

With the development of the Action Plan (NUREG-0660), NRC has transformed the recommendations of the investigating groups into discrete scheduled tasks that specify changes in its regulatory requirements, organization, or procedures.

Some actions to improve the safety of operating plants were judged to be neces-sary before an action plan could be developed, although they were subsequently included in the Action Plan. Such actions came from the Bulletins and Orders issued by the Commission immediately after the accident, the first report of, the Lessons Learned Task Force, and the recommendations of the Emergency Pre-paredness Task Force. Before these immediate actions were applied to operating plants, they were approved by the Commission.

NRC has identified a discrete set of licensing requirements related to TMI-2 in the Action Plan for the Ginna plant. NUREG-0737 entitled "Clarification of TMI Action Plan Requirements" was issued in November 1980. This report identi-fies the specific items from NUREG-0660 that'ave been approved by the Commis-sion for implementation at nuclear power plants't also includes additional information about schedules, applicability, method of implementation review, submittal dates, and clarification of technical positions. By letter dated December 17, 1982, Supplement 1 to NUREG-0737 was issued to provide additional clarification regarding safety parameter display systems, detailed control room design reviews, Regulatory Guide 1.97 (Revision 2) application to emergency response facilities, upgrading of emergency operating procedures, emergency response faci lties, and meteorological data. The schedules for completing the topics in Supplement 1 are being negotiated with the licensees and wi 11 be con-firmed by'n NRC order.

Each requirement from NUREG-0737 related to TMI-2 is addressed in Tables B. 1, 8.2, and 8.3 and in the following sections. Table 8. 1 lists those TMI items that the NRC staff considers complete. Table 8.2 lists those TMI items that Rochester Gas and Electric Corporation (RG8E). has stated are complete but which Ginna SER 8-1

have not been formally closed out by the NRC. All of the items in Table B.2 are included in the March 14, 1983 order confirming RG&E's committments.

Table B.3 lists those TMI items that are covered by Supplement 1 to NUREG-0737.

The remaining open items are discussed sequentially in the following sections.

I.A.1.3.2 MINIMUM SHIFT CREW This position defines shift manning requirements for normal operation. The letter of July 31, 1980 from D. G. Eisenhut to all'power reactor licensees and applicants set forth the interim criteria for shift staffing. Subsequent rule-making has codified the minimum shift crew in 10 CFR 50.54 and requires all licensees to meet the requirements 'by January 1, 1984.

Discussion By letter dated October 13, 1980, RG&E provided a planned schedule to meet the July 1, 1982 interim cr'iteria. 'In a letter dated February 1, 1982, RG&E stated that contingent on the results of the April and December 1982 licensing exam-inations, RG&E would be able to meet the license criteria in full by September 1983. In a letter dated August 5, 1983, the licensee informed the NRC that because of a licensee-requested change in the examination schedule, he would.

not be able to meet the license criteria until January 1, 1984.

II.B. 1 REACTOR COOLANT SYSTEM VENTS Each applicant and licensee shall install reactor coolant system (RCS) and reactor vessel head high point vents remotely operated from the control room.

Although the purpose of the system is to vent noncondensible gases from the.

RCS which may inhibit core cooling during natural circulation, the vents must =

not lead to an unacceptable increase in the probability of a loss-of-coolant, accident (LOCA) or a challenge to containment integrity. Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to 10 CFR 50, "General Design Criteria." The vent system shall be designed with sufficient redundancy that ensures a low probability of inadvertent or irreversible actuation.

Discussion As described in RG&E's letter dated July 1, 1981, the reactor coolant head vent system has been installed on the Ginna reactor. The system is operable although the 'fuses in the control/power circuits for the solenoid valves have been removed in the control room. Permission to operate the vents must be granted by the Plant Superintendent or the Technical. Support Center Manager.

1 By letter dated February 25, 1982, the staff "requested additional information from RG&E to complete the review of the reactor coolant head vent system. RG&E provided the information in a letter dated May', 1982. The staff is completing its review of this topic and expects to issue a safety evaluation by the end of 1983.

I I. F. 2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for 'the plant to supplement existing Ginna SER B-2

instrumentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret.,indication of inadequate core. cooling.

A,description of the function'al design requirements for" the system shall also

,,be included., A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.

t Discussion In the March 10, 1983, response to Generic Letter 82-28 (dated December 10, 1982), RG8E committed to install a reactor vessel water level indicating device and upgrade core exit thermocouples. RGBE currently is developing plans to install the instruments and expects to provide the staff with an implementation schedule by the fall of 1983. RG8E stated that on the basis of experience at other utilities and,discussions with the vendors, a system could be installed no sooner than the seco'nd refueling outage after the selection of the instrument type to be installed even if. expedited engineering effort is initiated on the project.

II.K.2.13 THERMAL MECHANICAL REPORT - EFFECT OF HIGH-PRESSURE INJECTION ON VESSEL INTEGRITY FOR SMALL-BREAK LOSS-OF-COOLANT ACCIDENT WITH NO AUXILIARY FEEDWATER A detailed analysis shall be performed of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater,. V Discussion This Action Plan item required a detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater. Westinghouse (in support of the Westinghouse Owners Group) has performed analyses for generic Westinghouse plant groupings to address this issue. The generic study is applicable to the R. E. Ginna plant. A report of the study was sent to the NRC by letter dated December 30, 1981.

The staff is continuing its review of this Action Plan item and expects to issue a safety evaluation in Fiscal Year 1984.

II.K.2.17 POTENTIAL FOR VOIDING IN THE REACTOR COOLANT SYSTEM DURING TRANSIENTS The potential for voiding in the reactor coolant system during anticipated transients shall be analyzed.

Discussion This Action Plan item required that an analysis be performed to address the potential, for void formation during natural circulation cooldown and depressur-ization transients.. Westinghouse (in support of the Westinghouse Own'ers Group) has performed the required analysis. The results of the analysis, which are applicable to Ginna, have been submitted to the NRC by letter dated April 20, 1981.

Ginna SER 6-3

In addition, the Westinghouse Owners. Group has developed a natural'circulation cooldown guideline that takes the results of the study into account so as to preclude void formation in the upper head region during natural circulation cooldown and depressurization transients, and specifies those conditions under which upper head voiding may occur. These Westinghouse Onwers Group'eneric guidelines have been submitted to the NRC by letter dated November 30, 1981.

The generic guidance developed by the Westinghouse Owners Group (augmented as appropriate with plant-specific considerations) has been used in the implemen-tation of Ginna plant-specific operating procedures as documented in a letter dated November 13, 198l.

Following the steam generator tube rupture that occurred on January 25, 1982, RGEE was required by NUREG-0916 to provide a detailed thermal-hydraulic analysis of system behavior during the incident to verify phenomena, including void for-mation. This analysis was provided in a letter dated November 22, 1982.

The staff is continuing its review of the above documents and 'expects to issue a safety evaluation in Fiscal Year 1984.

II.K.3.1 INSTALLATION AND TESTING OF AUTOMATIC POWER-OPERATED RELIEF VALVE ISOLATION SYSTEM All pressurized-water-reactor licensees should provide a system that uses the power-operated relief valve (PORV) block valve to protect against a small-break loss-of-coolant accident. This system wi 11 automatically cause the block valve to close when the reactor coolant system pressure decays after the PORV has opened. Justification should be provided to ensure that failure of this system would not decrease overall safety by aggravating plant transients and accidents.

Each licensee shall perform a confirmatory test of the automatic block valve closure system following installation.

Discussion Implementation of this Action Plan item was.modified in the May,1980 version of NUREG-0660. The change delays implementation of this item until after the studies specified in TMI Action Plan Item II. K.3.2 have been completed, if such studies confirm that the subject system is necessary.

II. K.3.2 REPORT ON OVERALL SAFETY EFFECT OF POWER-OPERATED RELIEF VALVE ISOLATION SYSTEM (1) . The licensee should submit a report for staff review documenting the var-ious actions taken to decrease the probability of a small-break loss-of-coolant accident caused by a stuck-open power-operated relief valve (PORV) and show how those actions constitute sufficient improvements in reactor safety.

(2) Safety-valve failure rates based on past history of, the operating plants designed by the specific nuclear steam supply system (NSSS) vendor should be included in the report submitted in response to (3.) above.

Ginna SER B-4

Discussion This Action Plan item called for documentation of modifications, if required, for automatic isolation of PORVs. An analysis performed for the Westinghouse Owners Group (WOG) determined that automatic isolation of PORVs is not required.

The results of the analysis and this conclusion were submitted by letter dated March 13, 1981.

The staff is continuing its review of this item and expects to issue a safety evaluation in late 1983.

II. K.3.5 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING LOSS-OF-COOLANT ACCIDENT Tripping of the reactor coolant pumps in case of a loss-of-coolant accident (LOCA) is not an ideal solution. Licensees should consider other solutions to the small-break LOCA problem (e. g., an increase in safety injection flow rate).

In the meantime, until a better solution is found, the reactor coolant pumps should be tripped automatically in case of a small-break LOCA. The signals designated to initiate the pump trip are discussed in NUREG-0623.

Discussion (The following discussion is taken from Generic Letter No.83-10d dated February 8, 1983.)

The staff has completed its evaluation of the analyses of LOFT Test L3-6 per-formed by the Westinghouse Owners Group and concludes that the evaluations acceptably predict the test results. Therefore, the staff finds the currently approved Westinghouse (W) evaluation model for small-break LOCAs in continued conformance. with Appendix K to 10 CFR 50 for the case of limited reactor cool-ant pump (RCP) operation after reactor trip and for the range of licensed W reactor designs.

The staff has reviewed industry analyses and performed its own analyses to determine whether RCP trip is necessary, during LOCAs and evaluated the desir-ability of continued RCP operation during non-LOCA transients and accidents, including steam generator tube ruptures. The staff has concluded that there is a wide range of transients and LOCAs where it is beneficial for the operators to maintain forced circulation cooling and mixing through operation of the RCPs.

However, some of the calculations show that for certain small-break LOCAs, pri-marily those with only one of the two high-pressure safety injection (HPSI) pumps assumed available, continued operation of the RCPs or continued operation of the RCPs followed by delayed RCP trip could lead to core damage.

Some uncertainty in these conclusions remains. Specifically, there is a complex interrelationship among break size, break location, RCP trip delay time, avail-able safety systems, and peak cladding temperature (PCT),for each type of NSSS design. Moreover, although the staff's and each vendor's calculational models adequately predicted LOFT Test L3-6, there appear to be subtle differences embedded, in the computer models which, when applied to large, commercial, pres-surized-water-reactor designs, yield differing results regarding the necessity for RCP trip during 'small-break LOCAs.

Ginna SER B-5

Because of this, the staff places substantial weight on the views of the reactor designers and the utilities which are almost unanimous in asserting that for some small-break LOCAs with less than the maximum available HPSI flow, delayed RCP trip could lead to core damage. Some utilities indicated their preference to keep the RCPs running for all events;'owever, this view appeared to be based solely on the desire to maintain forced circulation and did not consider -the consequences of delayed RCP trip.

While acknowledging the industry's general conclusion that the RCPs should be tripped for small-break LOCAs, both the staff and the industry recognized that there are other accident sequences of much higher probability than the small-break LOCA where the absence of forced circulation makes the operator's job more difficult and can increase the likelihood of operator errors. For this reason, the staff believes that a balance should be struck between the competing risks associated with tripping the RCPs early and leaving them running following transient and accident events.

Based on discussions with both licensees and the reactor manufacturers and internal evaluations, the staff believes that appropriate pump trip setpoints can be developed by the industry that would not require RCP trip for those transients and accidents where forced circulation and pressurizer pressure control is a major aid to the operators, yet would alert the operators to trip the RCPs for those small-break LOCAs where continued operation or delayed trip might result in core damage.

In summary, the staff has concluded that the need for RCP trip following a transient or accident should be determined by each licensee on a case-by-case basis, considering the Owners Group input. However, the staff must ensure that whatever decision is made reliable operation of reactors regarding pump operation, it will result in safe, and wi 11 not adversely affect the ability of licensees to comply with the Commission's rules and regulations.

For plants with low head safety injection pumps such as Ginna, the staff under-stands that RCP trip is still expected to occur on the low-pressure trip set-points currently proposed by Westinghouse for the design-basis steam generator tube rupture.. The staff considers this unacceptable and these licensees-should

'identify a more discriminating criterion for RCP trip that would allow continued RCP operation for tube leaks up to the design-basis steam generator tube rupture In his April 22, 1983, response to Generic Letter 83-10d, the licensee stated that Westinghouse and the Westinghouse Owners Group will undertake a two-part program to address the requirements of Generic Letter 83-10d for the purpose of providing more uniform RCP trip criteria and methods of determining those criteria. In the first part of the program, revised RCP trip criteria will be developed which provide an indication to the operator to trip the RCPs for small-break LOCAs requiring such action but will allow continued RCP operation for steam generator tube ruptures -less than or equal to a double-ended tube rupture. The revised RCP trip criteria will also be evaluated against other non-LOCA transients and accidents. where continued RCP operation is desirable to demonstrate that a need to trip the RCPs will not be indicated to the oper-ator for the more likely cases. Because this study is to be used for emergency response guideline development, better estimate assumptions will be applied consideration of the more likely scenarios. The first part of the program in'he will be completed and incorporated into Revision 1 of the Emergency Response Ginna SER B-6

Guidelines developed by Westinghouse for the Westinghouse Owners Group. The scheduled date for completion of Revision 1 is September 1983.

The second part of the program is intended to provide the required justification for manual RCP trip. This part of the program must necessarily be done after the completion of the first part of the program. The schedule for completion of the second part of the program is the end of 1983.

tl 1

The staff will review the Westinghouse Owners Group program when i,t is submi'tted.

III.K. 3. 30 REVISED SMALL-BREAK LOSS-OF-COOLANT-ACCIDENT METHODS TO SHOW COMPLIANCE WITH 10 CFR PART,50, APPENDIX K The analysis methods used by nuclear steam supply system vendors and/or fuel suppliers for small-break loss-of-coolant accident (LOCA) analysis for compli-ance with Appendix K to 10 CFR 50 should be revised, documented, and submitted for NRC approval. The revisions should account for .comparisons with experi-mental data, including data from the LOFT Test and Semiscale Test facilities.

Discussion letter I'y dated January 19, 1982, RG8E presented the Westinghouse position that the small-break LOCA analysis model currently approved by the NRC for use on Ginna is conservative and is in conformance with Appendix K to 10 CFR 50; How-ever, Westinghouse believes that improvement in the realism of small-break cal-culations is a worthwhile effort and has submitted its revised small-break LOCA analysis model (WCAP 10054) to address NRC concerns. The staff is currently reviewing Westinghouse's submittals and plans to issue a safety evaluation during the second quarter of Fiscal Year 1984.

II.K.3.31 PLANT-SPECIFIC CALCULATIONS TO SHOW COMPLIANCE WITH 10 CFR PART 50.46 Plant-specific calculations using NRC-approved models for small-break loss-of-coolant accidents as described in Item, II. K.3.30 to show compliance with 10 CFR 50.46 should be submitted for NRC approval by all licensees.

Discussion, ll a letter dated June 6, 1982, the licensee stated that he will develop a sched-

'n ule for submitting revised analyses, if required, after the NRC has approved a revised model and the extent of the reanalysis, the number of utilities requir-ing analysis, and the work load of Westinghouse have been evaluated.

REFERENCES Code of Federal Re ulations, Title 10, "Energy," U.S. Government Printing Office Washington, D.C. (includes General Design Criteria).

Letter, July 31, 1980, from D. G. Eisenhut (NRC) to All Power Reactor Licensees,

Subject:

Interim Criteria for Shift Staffing.

---, October 13, 1980, from D. L. White (RG8E) to,D. M. Crutchfield (NRC),

Subject:

Shift Staffing Interim Criteria.

Ginna SER B-7

---, March 13, 1981, from R. W. Jurgensen (WOG) to J. R. Miller (NRC),

Subject:

WCAP 9804, "Probabilistic Analysis and Operational Data in Response to Item II. K. 3. 2 for Westinghouse NSSS Plants.

1

- ,April 20, 1981, from R. W. Jurgensen (WOG) to P. S. Check (NRC),

Subject:

St. Lucie Cooldown Event Report.

---, July 1, 1981, from J. E. Maier, (RG&E) to D. M. Crutchfield (NRC),

Subject:

NUREG-0737 Requirements.

---, November 13, 1981,,from J. E. Maier-(RG&E) to D. M. Crutchfield (NRC),

Subject:

Generic Letter No. 81-21, Natural'irculation Cooldown.

---, November 30, 1981, R. W. Jurgensen (WOG) to D. G. Eisenhut (NRC),

Subject:

Emergency Response Guidelines.

---, December 30, 1981, from 0. D. Kingsley (WOG) to H. R. Denton (NRC),

Subject:

Summary Report on Reactor Vessel Integrity for Westinghouse'perating Plants, WCAP 10019.

---, January 19, 1982, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Status of January -1, 1982 NUREG-0737 Items.

---; February 1, 1982 from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Shift Staffing Criteria.

-, February 25, 1982, from D. M. Crutchfield (NRC) to J. E. Maier (RG&E)',

Subject:

Reactor Coolant System (RCS) Vents (Item II.B. 1), Request for Addi-tional Information..

---, May 7, 1982, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Reactor Coolant System Vents (TMI Item II.B. 1).

i

---, June 6, 1982, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Post TMI Requirements (Generic Letter 82-10).

---, November 22, 1982, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Response to Safety Evaluation Report - NUREG-0916.

- , December 10, 1982, from D.-G. Eisenhut (NRC) to All Licensees of Operating Westinghouse and CE PWRs,

Subject:

Inadequate Core Cooling Instrumentation System (Generic Letter 82-28).

---, December 17, 1982, from D. G. Eisenhut (NRC) to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Per-mits,

Subject:

Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability (Generic Letter 82-33).

February 8, 1983, from D.G. Eisenhut (NRC) to All Licensees 'With Westing-

,house (W) Designed Nuclear Steam Supply Systems (NSSSs) (Except Yankee Atomic Electric .Company),

Subject:

Resolution of TMI Action Item II. K. 3. 5, "Automatic Trip of Reactor Coolant Pumps" (Generic Letter 83-10d).

Ginna SER B-8

---, March 10, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Inadequate Core Cooling Instrumentation System (Generic Letter 82-12).

---, March 14, 1983, from D. M. Crutchfield (NRC), to J., E. Maier (RG&E),

Subject:

Order Confirming Licensee Commitments on Post-TMI Related Issues (Generic Letters 82-05 and 82-10).

---, April 22, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Response to Generic Letter No.83-10d, "Automatic Trip of Reactor Coolant Pumps."

- , August 5, 1983, from J. E. Maier (RG&E) to D. M. Crutchfield (NRC),

Subject:

Shift Staffing Criteria.

U.S. Nuclear Regulatory Commission, NUREG-0623, "Generic Assessment of Delayed Reactor Coolant Pump Trip During Small Break Loss-of-Coolant Accidents in Pres-surized Water Reactors," November 1979.

---, NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident,"

May 1980..

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

---, NUREG-0916, "Safety Evaluation Report Related to the Restart of R. E. Ginna Nuclear Power Plant," May 1982.

---, Regulatory Guide (RG) 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors."

---, RG 1.4, "Assumptions Used for Evaluating the Potential Consequences of a Loss of Coolant Accident for Pressurized Water Reactors."

---, RG 1.97, Rev. 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident."

Westinghouse Electric Corporation, WCAP 10054, "Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code," December 31, 1982.

Ginna SER B-9

Table B. 1 TMI items completed TMI Action Date of NRC Item Shortened title closeout letter I.A.1.1 Shift Technical Advisor 1/12/82 I.A.1.2 Shift Supervisor Responsibilities 7/7/80 I. A. 1. 3. 1 Shift Manning Overtime Limits 11/16/81 I.A.2.1 Immediate Upgrading of Reactor Operator and 7/2/82 Senior Reactor Operator Training and gualifi-cations - Modify Training I.A.2.3 Administration of Training Programs 10/31/80 I.C.2 Shift and Relief Turnover Procedures 7/7/80 I.C.3 Shi ft Supervi sor Responsibi 1 i ty 7/7/80 I. C.4 Control Room Access 7/7/80 I. C.5 Feedback of Operating Experience 11/16/81 I.C.6 Verify Correct Performance of Operating 11/16/81 Activities II. B.4 Training for Mitigating Core Damage 7/2/82 II. D. 3 Valve Position Indication 5/11/81 II. E. 1. 1 Auxiliary Feedwater System Reliability 6/16/82 II. E.1.2 Auxiliary Feedwater System Initiation 8/18/82 and Flow I I. E. 3. 1. 1 Emergency Power for Pressurizer Heaters- 7/7/80 Upgraded Power Supplies I I. E. 3. 1. 2 Emergency Power for Pressurizer Heaters- 5/11/81 Technical Specifications II. E.4.1 Dedicated Hydrogen Penetrations 7/20/81 I I. E. 4. 2. 1-4 Improve Diverse Isolation 7/7/80 I I. E.4. 2. 5 Containment Pressure Setpoint 1/13/82 II. E.4.2.6 Containment Purge Valves 12/15/82 II. E.4.2.7 Radiation Signal on Purge Valves 12/15/82 II. E.4.2.8 Containment Isolation Technical Specifications 5/11/81 II. F. 1. Noble Gas Monitor 10/20/81 II. F. l. Iodine/Particulate Sampling 10/20/81 II. F. 1. Containment High-Range Monitor 1/ll/82 II. G. 1 Power Supplies for Pressurizer Relief 5/11/81 Valves, Block Valves, and Level Indicators II. K. 1 IE Bulletins 5/28/80 II. K. 2. 19 Benchmark Analysis of Sequential 6/29/81 Auxiliary Feedwater Flow Ginna SER B-10

'Table B. 1 (Continued)"

. TMI Action Date of NRC Item Shortened title closeout letter II. K.3.3 Reporting Safety Valve and Relief Valve 8/18/82 Failures and Challenges II. K.3.9 Proportional Integral Derivative-'Controller 8/25/81 I I. K. 3. 10 Proposed Anticipatory Trip Modifications 8/25/81 II. K. 3. 12 Anticipatory Trip on Turbine Trip 8/25/81 II. K.3.17 Report on Outages of Emergency Core Cooling 8/17/83 Systems Licensee Report and Proposed Technical Specifications II . K. 3. 25 Power on Pump Seals 7/2/S2 I I I. A. 1. 1 Emergency Preparedness, Short Term 10/31/80 I I I. A. 2. 1 Emergency Preparedness, Long Term 5/25/83 I I I. D. 1. 1 Primary Coolant Outside Containment 5/11/81'/23/82 III.D. 3 ~ 3 Inplant Radiation Monitoring III. D. 3. 4 Control Room Habitability 4/11/83 Table B.2 TMI items completed according to Rochester Gas and Electric Corporation (RG8E) but not yet reviewed by NRC TMI Action Date of NRC Item Shortened title closeout letter I.A.3.1 Review Scope and Criteria for Licensing 4/23/82 Examinations II. B.2 Design Review of Plant Shielding and Environ-mental gualification of Equipment for Spaces/

Systems Which May Be Used in Postaccident Operations II. B. 3 Postaccident Sampling Capability I I. F. 1. 4 Accident Monitoring - Containment Pressure 4/23/82 I I. F. 1. 5 Accident Monitoring - Containment Water Level 4/23/82 I I. F. 1. 6 Accident Monitoring Containment Hydrogen 4/23/S2 II. D. 1. 2 Relief Valve and Safety Valve Test Programs 3/4/83 II. D. 1. 3 Block Valve Testing 6/11/82 "Completed in accordance with confirmatory order dated March 14, 1983.

Ginna SER B-ll

Table B.3 TMI covered by Supplement,l to NUREG-0737 TMI Action Item Shortened title I.C.1  : Guidance for the Evaluation and Development of Procedures for Transients and Accidents I.D.1 Control Room Design Reviews I.D.2 Plant Safety Parameter Display Console II I. A. 1. 2 Upgrade Emergency Support Facilities I I I. A. 2. 2 Meteorological Data Ginna SER B-12

t APPENDIX C UNRESOLVED SAFETY ISSUES C. 1 Introduction The NRC staff evaluates the safety requirements used in its reviews against new information as it becomes available. Information related to the safety of nu-clear power plants comes from a variety of sources including experience from operating reactors; research results; NRC staff and Advisory Committee on Reactor Safeguards (ACRS) safety reviews; and vendor, architect/engineer, and utility design reviews. After the accident at Three Mile Island (TMI), the Office for Analysis and Evaluation of Operational Data was established to pro-vide a systematic and continuing review of operating experience. Each time a new concern or safety issue is identified from one or more of these sources, the need for immediate action to ensure safe operation is assessed. This assessment includes consideration of the generic implications of the issue.

In some cases, -immediate action is taken'o ensure safety,"for, example, the derating of boiling-water reactors (BMRs) as a result of the channel box wear problems in 1975. In other cases, inter im measures, such as modifications to operating procedures, may be sufficient to allow further study of the issue before licensing decisions are made. In most cases, however, the initial assessment indicates that immediate licensing actions or changes in 'licensing criteria are not necessary. If the issue applies to several, or a class of, plants, it is evaluated further as a "generic safety issue." This evaluation considers the safety significance of the issues, the cost to implement any changes in plant design or operation, and other significant and relevant fac-tors to establish a priority ranking of the issue. On the basis of this rank-ing, the issue is (1) scheduled for near-term resolution, (2) deferred until resources become 'available, or (3) dropped from further consideration.

The issues with the highest priority ranking are reviewed to determine whether they should be designated as "unresolved safety issues" (NUREG-0410, "NRC Pro-gram for'the Resolution of Generic Issues Related to Nuclear Power Plants,"

dated January 1, 1978). However, as discussed above, such issues are considered on a generic basis only after the staff has made an initial determination that the safety significance of the issue does not prohibit continued operation or require licensing actions while the longer-term generic review is under way.

C.2 ALAB-444 Re uirements These longer-term generic studies were the subject of a Decision by the Atomic Safety and Licensing Appeal Board of the Nuclear Regulatory Commission. The Decision was issued on November 23, 1977 (ALAB-444) in connection with the Appeal'oard's consideration of the Gulf States Utility Company application for the River Bend Station, Units 1 and 2.

Ginna SER C-1

In the view of the Appeal Board, In short, the board (and the public as well) should be in a position to ascertain from the SER itself--without the need to resort to extrinsic documents--the staff's perception of the nature and extent of the relationship between each significant unresolved generic safe-ty question and the eventual operation of. the reactor under scrutiny.

Once again, this assessment might well have a direct bearing upon the ability of the licensing board to make the safety findings required of it on the construction permit level even though the generic answer to the question remains in the offing. Among other things, the fur-nished information would likely shed light on such alternatively impor-tant considerations as whether: (1) the problem has already been resolved for the reactor under study; (2) ther e is a reasonable basis for concluding that a satisfactory solution will be obtained before the reactor is put in operation; or (3),the problem would have no safety implications until after several years of reactor operation and", should it not be resolved by then, alternative means will be available to ensure that continued operation (if permitted at all) would not pose an undue risk to the public.

This appendix is specifically included to respond to the decision of the Atomic Safety and Licensing Appeal Board as enunciated in ALAB-444, and as applied to an operating license proceeding ~Vir inia Electric and Power ~Cpm an (North Anna Nuclear Power Station, Unit. Nos. 1 and 2), ALAB-491, 8 NRC 245 (1978).

C.3 Unresolved Safet Issues In,a related matter, as a result of Congressional action on the NRC budget for Fiscal Year 1978, the Energy Reorganization Act of 1974 was amended (PL 95-209) on December 13, 1977 to include, among other things, a new Section 210 as follows:

UNRESOLVED SAFETY ISSUES PLAN SEC. 210. The Commission shall develop, a plan providing for specifi- >>

cation and analysis of unresolved safety issues relating to nuclear reactors and shall take such action as may be necessary to implement corrective measures with respect to such issues. Such plan shall be submitted to the Congress on or before January 1, 1978 and progress reports shall be included in the annual report of the Commission thereafter.

The Joint Explanatory Statement of the House-Senate Conference Committee for the Fiscal Year 1978 Appropriations Bill (Bill S. 1131) provided the following additional information regarding the Committee's deliberations on this portion of the bill:

SECTION 3 - UNRESOLVED SAFETY ISSUES The House amendment required development of a plan to resolve generic safety issues. The conferees agreed to a requirement that the plan Ginna SER C" 2

be submitted to the Congress on or before January 1, 1978. The con-ferees also expressed the intent that this plan should identify and describe those safety issues, relating to nuclear power reactors, which are unresolved on the date of enactment. It should set forth:

'(1) Commission actions taken directly or indirectly to develop and implement corrective "measures; (2) further actions planned concerning such measures;. and (3) timetables and cost estimates of such actions.

The Commission should indicate the priority it has assigned to each issue, and the basis on which priorities have been assigned.

In response to the reporting requirements of the new Section 210, the NRC staff submitted to Congress on January 1, 1978, a report (NUREG-0410) that described the NRC generic issues program. The NRC program was already in place when PL 95-209 was enacted and is of considerably broader 'scope than the Unresolved Safety Issues Plan required by Section 210. In the letter transmitting NUREG-0410 to the Congress on December 30, 1977, the Commission stated, "The progress reports, which are required by Section 210 to be included in future NRC annual reports, may be more useful to Congress if they focus on the specific Section 210 safety items."

It is the NRC's view that the intent of Section 210 was to ensure that plans were developed and implemented on issues with potentially significant public safety implications. In 1978, the NRC undertook a review of'ore than 130 generic issues addressed in the NRC program to determine which issues fit this description and qualify as unresolved safety issues for reporting to the Congress. The NRC review included the development of proposals by the NRC staff and review and final approval by the NRC Commissioners.

This review is described in NUREG-0510, "Identification of Unresolved Safety Issues Relating to Nuclear Power Plants - A Report to Congress," dated January 1979. The report provides the following definition of an unresolved safety issue:

An Unresolved Safety Issue is a matter affecting a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants it affects.

Further, the report indicates that in applying this definition, matters that pose "important questions concerning the adequacy of existing safety require-ments" were judged to be those for which resolution is necessary to (1) compen-sate for a possible major reduction in the degree of protection of the public health and s'afety or (2) provide a potentially significant decrease in the risk to the public health and safety. guite simply, an unresolved safety issue is potentially significant from a public safety standpoint, and its resolution is likely to result in NRC action in the affected plants.

All of the issues addressed in the NRC program were systematically evaluated against this definition as described in NUREG-0510. As a 'result, 17 unresolved safety issues addressed by 22 tasks in the NRC program were identified.

An in-depth and systematic review of generic safety concerns identified between January 1979 and March 1981 was performed by the staff to determine if any of Ginna SER C-3

these issues should be designated,.as unresolved safety issues. The candidate issues originated from concerns identi.fied in NUREG-0660, "NRC Action Plan as a Result of the TMI-2 Accident"; ACRS recommendations; abnormal occurrence re-ports; and other operating experience. The staff's proposed list was reviewed and commented on by the ACRS,,the Office of Analysis and Evaluation of Opera-tional Data (AEOD), and the Office of Policy Evaluation. The ACRS and AEOD also proposed that several additional unresolved, safety issues be considered by the Commission. The Commission considered the above information and approved the four Unresolved Safety Issues A-45 through A-48. A description of the re-view process of candidate issues, together with a list of the issues considered is presented in NUREG-0705, "Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants, Special Report to Congress," dated March 1981.

An expanded discussion of each of the new unresolved safety issues is also con-tained in NUREG-0705. In. addition to the four issues identified above, the Commission approved another issue, A-49, "Pressurized Thermal Shock," as an unresolved safety issue in December 1981.

The issues are listed below. The, number(s) of the generic task(s) (for example, A-1) in the NRC program addressing each issue is (are) indicated in parentheses following the title.

Unresolved Safet Issues A licable Task Nos.

(1) Waterhammer (A-1)

(2) Asymmetric Blowdown Loads on the Reactor Coolant System (A-2)

(3) Pressurized Water Reactor Steam Generator Tube Integrity (A-3, A-4, A-5)

(4) BWR Mark I and Mark II Pressure Suppression Containments (A-6, A-7, A-8, and A-39)

(5) Anticipated Transients Without Scram (A-9)

(6) BWR Nozzle Cracking (A-10)

(7) Reactor Vessel Materials Toughness (A-ll)

(8) Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports (A-12)

(9) Systems Interaction in Nuclear Power Plants (A-17) h (10) Environmental qualification of Safety-Related Electrical 'Equipment (A-24)

I (ll)'Reactor Vessel Pressure Transient Protection (A-26)

I h II (12) Residual Heat Removal Requirements- (A-31) p Jl

.(13) Control of, Heavy Loads Near Spent Fuel (A-36) t I

(14) Seismic Design Criteria (A-40) 1 "h I 'll Ginna SER C-4

(15) Pipe Cracks at Boiling. Water Reactors (A-42)

(16) Containment Emergency Sump Reliability (A-43)

(17) Station Blackout (A-44)

(18) Shutdown Decay Heat Removal Requirements (A-45)

(19) Seismic gualification of Equipment in Operating Plants (A-46)

(20) Safety Implications of Control Systems (A-47)

(21) Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment (A-48)

(22) Pressurized Thermal Shock (A-49)

In the view of the staff, the unresolved safety issues (USIs) listed above are the substantive safety issues referred to by the Appeal Board in ALAB-444 when it spoke of "... those generic problems under continuing study which have...

potentially significant public safety implications." Nine of the tasks iden-tified with the USIs are not applicable to Ginna, and six of these nine tasks (A-6, A-7, A-8, A-10, A-39, and A-42) are peculiar to boiling-water reactors.

Tasks A-4 and A-5 address steam generator tube problems in Combustion Engineer-ing and Babcock 8 Wilcox plants. Task A-48 is related to pressurized-water reactor (PWR) plants with ice condenser containments or BWR plants with III-type containment. With regard to the remaining 18 tasks that are 'ark applicable to this facility, the NRC staff has issued NUREG reports providing its proposed resolution of 7 of these issues. The table below lists those issues and the NUREG reports.

Task No. NUREG report and title A-2 NUREG-0609, "Asymmetric Blowdown Loads on PWR Primary Systems" A-9 NUREG-0460, Vol. 4, "Anticipated Transients Without Scram for Light Water Reactors" A-ll NUREG-0744, "Resolution of the Task A-ll Reactor Vessel Materials Toughness Safety Issue," Vols. I and II, Rev. 1, October 1982 A-24 NUREG-0588, Rev. 1, "Interim Staff Position on Environmental gualification of Safety-Related Electrical Equipment" A-26 NUREG-0224, "Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors and BTP RSB 5-2 A-31 SRP Section 5.4.7 and BTP RSB 5-1 "Residual Heat Removal Systems" incorporate requirements of USI A-31

'-36 '"" I NUREG'-0612, "Control of Heavy Loads at Nuclear Power Plants" Ginna SER C-5

The remaining issues applicable to this facility are listed below.

(1) A-1 Waterhammer (2) A-3 Westinghouse Steam Generator Tube Integrity (3) A-12 Potential for Low Fracture Toughness and Lamell.ar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports (4) A-17 Systems Interactions in Nuclear Power Plants (5) A-40 Seismic Design Criteria (6) A-43 Containment Emergency Sump Reliability (7) A-44 Station Blackout (8) A-45 Shutdown Decay'eat Removal Requirements (9) A-46 Seismic qualification of Equipment in Operating Plants (10) A-47 Safety Implications of Control Systems (11) A-49 'Pressurized Thermal Shock With the exception of Task A-12, all task action plans for the generic tasks up to and including A-40 above are included in NUREG-0649, "Task Action Plans for Unresolved Safety Issues Related to Nuclear Power Plants." NUREG-0577, which represents staff resolution of USI A-12, was issued for comment in November 1979. The NUREG contained the task action plan for A-12. Task action plans for later tasks were issued individually as indicated in the table below.

Task action plans for selected USIs, Task Issue task number action plan A-43 01/81 Q-44 07/80 A-45 10/81 06/82 (Rev. 1)

A-46 05/82 A-47 06/82 A-49 03/82

'I Each task action plan provides a description of the problem; 'the staff's ap-proach to its resolution; a general discussion of the bases, on which continued plant licensing or operation can"proceed pending completion of the task; the technical organizations involved in the task and estimates of the manpower Ginna SER C-6

required; a description of the interactions with other NRC offices, the Advisory Committee on Reactor Safeguards, and outside organizations; estimates of fund-ing required for contractor-supplied technical assistance; prospective dates for completing the task; and a description of potential problems that could alter the planned approach or schedule.

In addition to the task action plans, the staff issues the "Office of Nuclear Reactor Regulation Unresolved Safety Issues Summary" (Aqua Book) (NUREG-0606) on a quarterly basis, which provides current schedule information for each of the unresolved safety issues. It also includes information relative to the implementation status of each unresolved safety issue for which technical resolution is complete.

The staff has reviewed the 18 USIs listed above as they relate to Ginna. Dis-cussion of each of these issues is provided in Section C.4. On the basis of its review of these items, the staff concludes that there is reasonable assurance that Ginna can continue to be operated before the ultimate resolution of these generic issues without endangering the health and safety of the public.

However, Tasks A-9, A-ll, A-24, A-26, A-31, and A-36 are accepted subject to the resolution of those open items identified in Section C.4.

C.4 Discussion of Tasks as The Relate to Ginna This section provides the NRC staff's evaluation of the Ginna facility for each of the applicable unresolved safety issues. This includes the staff bases for continued operation before ultimate resolution of these issues. 'he staff's conclusions are based in part on information provided by the licensee in his letter of October 19, 1982.

A-1 Waterhammer Waterhammer events are intense pressure pulses in fluid systems caused by any one of a number of mechanisms and systems conditions. Since 1971 there have been approximately 150 incidents involving waterhammer reported for pressurized-water reactors and boiling-water reactors. The waterhammers have involved steam generator feed rings and piping, decay heat removal systems, emergency core cooling systems, containment spray lines, service water lines, feedwater lines, and steam lines. However, the PWR systems most frequently affected by water-hammer effects are the feedwater systems. The most significant waterhammer events have occurred in the steam generator feed rings of pressurized-water reactors.

A number of factors at the Ginna plant reduce the likelihood of steam generator waterhammers. These are (1) limiting auxiliary feedwater flow to less than 150 gpm when steam generator levels are low and there is no safety requirement for more feedwater (2) automatic start of auxiliary feedwater on (a) loss of all feedwater (b) loss of offsite power (c) low low level in any one steam generator (d) safety injection Ginna SER C-7

(3) existence of only a short length of feedwater piping between a steam generator and its loop seal The likelihood of waterhammer was further reduced by installation in 1979 of "J" tubes on the Ginna steam generator feed rings. The staff's SER for Ginna (forwarded by letter dated December 20, 1979) relative to steam generator water-hammer concluded that "...the means for reducing the potential for steam genera-tor, waterhammer at, this facility [Ginna] are adequate... and no fur ther action is required of the licensee with regard to steam generator waterhammer." Use of "J" tubes has been demonstrated to be effective in minimizing conditions lead-ing to waterhammer.

Waterhammer occurrences and underlying causes have been evaluated through USI A-1. The staff's technical findings are reported in NUREG-0927. Actions taken by the licensee (e. g., installations of "J" tubes in the steam generator feed ring and limits on auxiliary feedwater flow rates) are consistent with generic findings that support the use of such design features and controls for minimizing, or eliminating, waterhammer occurrence. In addition, no water-hammer occurrences have been reported for the Ginna plant since 1975, follow-ing corrective design actions related to feedwater control valve instabilities.

On the basis of experience and plant design and operational changes implemented by the licensee, the occurrence of waterhammer is judged to be low. Nonetheless, should a pipe break occur, the emergency core cooling systems (which are based on the design-basis accident (DBA)) will provide adequate core cooling. Should a waterhammer occur at a future date at the Ginna plant, plant-specific evalua-tions and actions will be taken. Generic evaluations have not identified the need for any additional measures other than those already implemented.

On the basis of the Ginna design, operating experience, and operating proce-dures, the staff concludes that the waterhammer issue is properly addressed for the Ginna plant and that operation can continue without undue risk to the health and safety of the public.

A-2 S mmetric Blowdown Loads on the Reactor Coolant S stem In January 1978, all licensees for pressurized-water reactor (PWR) plants were required by NRC to provide an assessment of the adequacy of their reactor vessel supports and other affected structures to withstand combinations of re-sponse to asymmetric loss-of-coolant-accident (LOCA) loads and the safe shut-down earthquake.

Rochester Gas and Electric Corporation (RG&E) is an active participant in the Westinghouse A-2 Owners Group, which addresses this issue. The following Westinghouse topical reports, which are applicable to Ginna and relate to the "leak-before-break", concept, have been submitted,to the NRC, and are'currently being evaluated by, the NRC staff and its contractor,-.Edgerton, Germeshausen &

Grier:

(1) WCAP 9558 through Rev. 2, May 1981 (2) WCAP 9570, October 1979 and June 1980 (3) WCAP 9628, November 1979 (4) WCAP 9662, January and February 1980-Ginna SER C-,8,

(5) WCAP 9748, June 1980 (6) WCAP 9749, June 1980 (7) WCAP 9787 th'rough Rev. 1, May 1981 I

Although the staff has not yet issued. the final acceptance of the Westinghouse A-2 Owners Group and RG8E analyses, it is expected that this acceptance will complete all open issues related to Task A-2.'he analyses have already been discussed with the Advisory Committee on Reactor Safeguards.

The staff concludes that there is reasonable assurance that until this generic issue is resolved, the Ginna plant can continue operation without undue risk to the health and safety of the public.

A-3 Steam Generator Tube Inte rit In 1977, the U.S. Nuclear Regulatory Commission established task action plans (TAPs) A-3, A-4, and A-5 to evaluate the safety significance of tube degrada-tion in Westinghouse, Combustion Engineering, and Babcock 8 Wilcox steam gen-erators, respectively. These tasks were later designated as unresolved safety issues in the Commissioner's 1978 Annual Report (NUREG-0516), pursuant to Sec-tion 210 of the Energy Reorganization Act of 1974.

The TAPs integrated technical studies in the areas of systems analyses, inserv-ice inspection, and tube integrity to establish improved criteria for ensuring adequate tube integrity and safe steam generator operation. A detailed descrip-tion of all the evaluations and calculations performed, the resulting require-ments and criteria, and the strategy for implementation are presented in a draft report entitled "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity" (NUREG-0844).

Members of the staff held a meeting on July 6, 1983, with the Steam Generators Owners Group to discuss the proposed resolution. The staff presented an over-view of the program development and scope. The program evolved from the origi-nal USI effort on steam generators and was expanded to include consideration of issues stemming from the staff review of previous steam generator tube rupture events (NUREG-0651 and NUREG-0909). The program package (NUREG-0844) has pro-posed a number of new requirements for industry and issues warranting further staff evaluation that are subject to approval by the Committee for the Review of Generic Requirements and the Commission. A summary of the July 6 meeting was issued by letter dated July 13, 1983.

Corrosion resulting in steam generator tube wall thinning (wastage) has been observed in, several Westinghouse and Combustion Engineering plants for a number of, years. .Major, changes in their secondary water treatment process ('from phos-phate treatment to all-volatile treatment) essentially eliminat'ed'thi's form of degradation., Another major corrosion-related phenomenon resulting from a of'upport plate corrosion products in the annulus between the tubes 'uildup

. the. support plates. also has, been .observed in a number of plants in 'rece'nt 'nd years. This buildup eventually causes a diametral reduction of'he tubes, called "denting,". and deformation-of the tube. support plates.. This phenomenon has led to other problems, including stress-corrosion cracking, leaks at the tube/sUpport plate intersections, and U-bend-section cracking of tubes that were-highly. stressed because of support'plate-;deformation.'I II Ginna SER C"'9 r E

As discussed in Section 5.4.2, RG8E replaced the original phosphate secondary-side water chemistry treatment with an all-volatile treatment in Nov'ember 1974 and added full-flow condensate polishing demineralizers in 1978. At present, less than 5X of the tubes in the A steam generator and slightly more than 5X of the tubes in the B steam generator have been plugged. In addition, 99 tubes in the B steam generator and 4 tubes in the A steam generator have been sleeved.

The primary reasons for tube repair have been wastage and crevice intergranular attack. As a result of the change in chemistry, wastage no longer appears to be occurring.

RGRE's present program of steam generator tube inspections provides for eddy current tests of the tubes, tube sheet water lancing, and crevice cleaning if necessary. Further, RGSE has developed a sleeving program to install sleeves as a preventive measure on those steam generator tubes considered most sus-ceptible to crevice intergranular attack.

Pending completion of Task A-3, the measures taken at Ginna should minimize the steam generator tube problems encountered. Further, the inservice inspec-tion and Technical Specification requirements will ensure that the licensee and the NRC staff are alerted to tube degrations should they occur. Appropriate actions such as tube plugging, increased and more frequent inspections, and power derating could be taken if necessary. Because the improvements that will result from Task A-3 are expected to be procedural, that is, improved inspection at the steam generators, they can be implemented by the licensee after resolution of this issue.

On the basis of the foregoing, the staff has concluded that the Ginna plant can continue to be operated before final resolution of this generic issue without undue risk to the health and safety of the public.

A-9 Antici ated Transients Wi,thout Scram Nuclear plants have safety and control systems to limit the consequences of temporarily abnormal operating conditions or "anticipated transients." Some deviations from normal operating conditions may be minor; others, occurring less frequently, may impose significant demands on plant equipment. In some anticipated transients, rapidly shutting down the nuclear reaction (initiating a "scram"), and thus rapidly reducing the generation of heat in the reactor core, is an important safety measure. If there were a potentially severe "anticipated transient" and the reactor shutdown system did not "scram" as desired, then an "anticipated transient without scram," or ATWS, would occur.

WASH-1270, "Technical Report on Anticipated, Transients Without Scram for Water-Cooled Power Reactors," discussed the. probability. of an ATWS event and an appro-priate safety objective for these events. Westinghouse Topical Report WCAP 8404, "ATWS Analysis for Westinghouse PWR's With 44 Series Steam Generators," was released in September 1974. Following review of this report, as'well as the many other vendor reports describing the analysis models and results, the NRC staff published, in late 1975, its status report on, each vendor analysis includ-ing detailed guidelines on analysis models and ATWS safety objectives (USNRC, Dec. 9, 1975).

f Since the publication of the 1975 status reports,'dditional information rele-vant to ATWS has been developed by the industry and the Reactor Safety Study Ginna SER C-10

Group. On the basis of its review of these reports and discussions with vendors, NRC published "Anticipated Transients Mithout Scram for Light-Water Reactors,"

NUREG-0460, Volumes 1 and 2, in April 1978. Since the issuance of Volumes 1 and 2, additional safety and cost information and new insights have been devel-oped on the general subject of quantitative risk assessment. On the bases of these considerations, the NRC staff issued a new report, Volume 3 to NUREG-0460, dated December 1978. In Volume 3 various alternative plant modifications for ATWS ranging from none to those needed to satisfy the proposed licensing crite-ria for new plants in NUREG-0460, Volumes 1 and 2, were considered. The staff assessed the corresponding degrees of assurance of safety achieved by these alternative modifications. In Volume 3, the staff also suggested plant modifi" cations on the basis of the plant design and age. To confirm its judgment on the adequacy of these designs, the staff issued requests for'industry to supply the necessary generic analyses. Genetic Westinghouse responses, applicable to Ginna, were presented to the NRC by reports ("Anticipated Transients Without Scram for Mestinghouse Plants" ) dated June 8, 1979, and December 30, 1979. In NUREG-0460, Volume 4, issued in March 1980 for public comment, the staff re-viewed the industry responses 'and concluded that the necessary verification of the adequacy of the proposed design changes had not been provided. The staft, therefore, proposed that early improvements in safety should be provided, and any additional requirements should be considered under the staff's recommended rulemaking. The staff has reviewed the comments of industry and the ACRS in Volume 4 and has published a proposed rule for resolution of ATMS in the Federal

~Re ister (45 FR 73000). Ginna will be required te meet the rule when it is issued in final form.

As a result of the Salem ATWS events, the Commission published NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant." In a letter dated July 8, 1983, the staff identified the actions all 1icensees needed to take based on NUREG-1000. These actions address issues related to reactor trip system reliability and general management capability. RG8E is required to provide in November 1983 the status of conformance with the posi-tions contained in the July 8 letter, and plans and schedules for any needed improvements to demonstrate conformance with the positions. The schedules for implementation will be negotiated between the NRC Project Manager and RG8E consistent with the staff's goal of integrating new requirements and consider-ing the unique status of each plant and the relative safety importance of the improvements, combined with all other plant programs. The resolution of-these actions will be reported separately.

On the basis of the above considerations and subject to satisfactory resolution of the open items identified, the staff has concluded that there is reasonable assurance that Ginna can continue to operate before the ultimate resolution ot this generic issue without endangering the health and safety of the public.

A-11 Reactor Vessel Materials Tou hness Resistance to brittle fracture, a rapidly propagating catastrophic failure mode for a component containing flaws, is described quantitatively by a material property 'generally denoted as "fracture toughness." Fracture tough-ness has different values and characteristics depending on the material being considered. For steels used in a nuclear reactor pressure vessel, three con-siderations are important; first, fracture toughness increases with increasing Ginna SER C"11

temperature; second,,fracture toughness decreases with increasing load .rates;.

and third, fracture toughness decreases with neutron irradiation. In recogni-tion of these considerations, power reactors are operated within, restrictions imposed by the Technical Specifications on the pressure during heatup and cooldown operations. These restrictions ensure that the reactor vessel will not be subjected to a combination of pressure and temperature that could cause brittle fracture of the, vessel if. there were significant flaws in the vessel materials. The effect of neutron radiation on the fracture toughness of the vessel material is accounted for in developing .and>>revising these Technical Specification l.imitations.

For the service times and operating conditions typical, of current operating plants, reactor vessel fracture toughness for most plants provides adequate margins. of safety against vessel failure under operating, testing, maintenance, and anticipated transient conditions and accident conditions over the life of the plant. However, results from a reactor vessel surveillance program and, analyses performed for,up to 20 older operating pressurized-water reactors, (PWRs) and those for some more recent vintage plants show, that they will have marginal toughness, relative to required margins, at normal full power after a comparatively short period of operation. In addition, results from analyses performed by manufacturers of PWRs,indicate that the integrity of some reactor vessels may not be maintained if a main steam-line break or a loss-of-coolant accident should occur after approximately 20 years of operation. The principal objective of Task A-ll was to develop an improved engineering method and safety criteria to allow a more precise assessment of the safety margins that are available during normal operation and transients in older reactor vessels with marginal fracture toughness and, of the safety margins available during accident conditions for all plants.

Appendices G and H of 10 CFR 50 require that compliance with minimum fracture toughness requirements be demonstrated and that a materials surveillance to monitor changes in the fracture toughness properties of'ferritic pro-'ram materials in the reactor vessel beltline region be maintained. This issue was discussed during the Systematic Evaluation Program (SEP) review of SEP Topic V-6, "Reactor Vessel Integrity," in NUREG-0569, ."Evaluation of the Integrity of SEP Reactor Vessels." On the basis of the recommendations in that report, RG8E committed to provide for staff review an evaluation of the next surveillance capsule, including a complete chemical analysis of the capsule. The capsule was removed from the reactor in 1980 and was shipped to RG&E's contractor, Westinghouse, in 1981.

The licensee submitted the material test results of the third capsule of the reactor pressure vessel material surveillance program for Ginna in a letter dated December 8, 1982. The staff has reviewed the test results and has de-termined that the licensee's proposed pressure-temperature limits are accept-able until the intermediate-to-lower shell, weld accumulates a neutron fluence of 1.5 x 10~e n/cmz (E > 1 MeV). The staff's review of the licensee's radia-tion analysis and neutron dosimetry indicated that a number of items need to be addressed further by the licensee before the staff can determine the number of EFPYs that the proposed pressure-temperature limits may be used. RGBE has agreed to provide the additional information.

Ginna SER C-12

On the basis of the above capsule evaluation results, the staff's May 3, 1982 evaluation of. SEP Topic V-6 as related to reactor vessel integrity and Supple-.

ment 1 to the IPSAR, the staff concludes that continued operation will not pose an undue risk to the health and safety of the public pending the ultimate resolution of this issue.

A-12 Potential for Low Fracture Tou hness and Lamellar Tearin .on PWR Steam Generator and Reactor Coolant Pum Su orts During the cour se of the licensing action for North Anna Power Station, Units 1 and 2, a number of questions were raised concerning the potential for lamellar tearing,and low fracture toughness of the steam generator and reactor coolant pump support materials for those facilities. Two different American Society for Testing and Materials steel specifications (ASTM A36 and ASTM A572) covered most of the material used, for these supports. Toughness tests, not originally specified and not in the relevant ASTM specifications, were made on those heats of steel for which excess material was available. The toughness of the A36 steel was found to be adequate, but the toughness of the A572 steel was rela-tively poor at a temperature of .80 F.

Because similar materials and designs have been used on other nuclear plants, the concerns regarding the supports for the North Anna facilities are appli-cable to other PWR plants. It was therefore considered necessary to reassess the fracture toughness of the steam generator and reactor coolant pump support materials for all operating PWR plants and those undergoing construction and operating license .reviews.

NUREG-0577, "Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," was issued for comment in October 1979. This report summarized work performed by the NRC staff and its contractor, Sandia Laboratories, to resolve the generic activity. The report describes the technical issues, the technical studies performed by Sandia Labo-ratories, the NRC staff's technical positions based on these studies, and a pro-posed plan for implementing the technical positions. As a part of initiating the implementation of the findings in this report, letters were sent to all applicants and licensees on May 19 and 20, 1980. In these letters, a revised proposed implementation plan was presented and specific criteria for material qualification were defined.

With regard to the lamellar tearing issue, the results of an extensive litera-ture survey by Sandia revealed that although lamellar tearing is a common occur-rence in structural steel construction, virtually no documentation exists des-cribing inservice fai lures as a result of lamellar tearing. Nonetheless, additional research was recommended to provide a more definitive and complete evaluation of, the importance of lamellar tearing to the structural integrity of nuclear power plant support systems.

Many comments on both the draft of NUREG-0577 and the letters of May 19 and 20 were received by the NRC staff. After completing its review and analysis of the comments provided, the staff prepared NUREG-0577, Revision 1, which includes a full discussion and resolution of the comments and a considerably revised plan for implementation.

Ginna SER C-13

The materials of construction in the Ginna steam generator and reactor coolant pump supports have been determined to be different from those used at the North Anna Station. RG&E's submittal relative to the support material was provided in a report transmitted by letter dated June 26, 1978. RG&E determined that adequate fracture toughness exists for the supports at Ginna. This report also reviewed the parameters that affect the potential for lamellar tearing in weld-ments--steel quality, steel fabrication practice, weld joint design, grade of filler material, weld dimensions, and postweld heat treatment. RG&E deter-mined that lamellar tearing would not be a problem for the Ginna design and installation.

The staff has concluded in NUREG-0577, Revision 1, that modifications such as installation of heaters or structural repairs to existing supports are not justified on the basis of arguments presented and supported by the value-impact analysis. Further, the staff has concluded that 'because repairs or other cor-rective measures are not justified, no actions are required to verify material properties beyond what is already accomplished. The staff concludes that this issue has been adequately addressed for Ginna and that operation can continue without undue risk to the health and safety of %he public.

A-17 S stems Interaction"in Nuclear Power Plants The staff's systems interaction program was initiated in May 1978 with'he definition of USI A-17 (" Systems Interaction in Nuclear Power Pla'nts") and was intensified by TMI-2 Action Plan (NUREG-0660) Item II.C.3 (" Systems Interaction" ).

The concern arises because the design, analysis, and installation of systems are frequently the responsibility of teams of engineers with functional specialties such as civil, electrical, mechanical, or nuclear. Experience at operating plants has led to questions of whether the work of these functional specialists is sufficiently integrated to enable them to minimize adverse interactions among systems. Some adverse events that occurred in the past 'might have been prevented if the teams had ensured that there was necessary independence of safety systems under all conditions of operation.'he NRC staff's current procedures assign primary responsibility for review of various technical areas to specific organizational units and assign secondary responsibi li'ty to other units where there is a functional interface. Designers follow somewhat similar procedures and provide the analyses of systems and interface reviews. Under Task A-17 methods are being developed that will be able to identify adverse systems interactions that were not considered by cur-rent review procedures.- The first phase of this study began in May 1978 and was completed in February 1980 by Sandia Laboratories under contract to the NRC staff (letter dated February 22, 1980).

The Phase I investigation was structured to identify areas where interactions are possible between systems and that have the potential for negating or se-riously degrading the performance of safety functions. The study concentrated on commonly caused failures among systems that would violate a safety function.

The investigation was to then identify where NRC review procedures may not have properly accounted for these interactions.

Sandia Laboratories used fault-tree analysis on the selected design to identify component failure combinations (cut-sets) that could result in a loss of a Ginna SER

safety function. The cut-sets were further reduced by incorporating six link-ing failures into the analysis. The results of the Sandia effort indicated "a few potentially adverse. systems'nteractions within the limited scope of the study. The staff reviewed the interactions for safety significance and generic implications. The staff'concluded that no corrective measures needed to be implemented immediately, except for the potential interaction between the"power-operated relief valve and its block v'alve. This interaction had been separately identified by the evaluations of the THI-2 accident while Sandia was performing the study. Because corrective measures were already being implemented, no separate measures were needed under USI A-17'. H W

lt NUREG-0660 provides for a systems interaction follow-on study in Section II. C. 3, "Systems Interactions." Since April 1980, NRC has intensified the effort both by broadening the study of*'methods to identify potential systems interactions and by preparing guidance for audit reviews of selected plants for systems interactions. Recent experience pr'ovides a basis for the development'by the staff of a more efficient review process for potential systems interactions.

The process will provide for a resolution of USI A-17, assimilate operating reactor experience, and rank identified systems interactions by their relative importance to safety.

It is expected that the development of systematic ways to identify, rank, and evaluate systems interactions will further reduce the likelihood of intersystem failures that result in the loss of plant safety functions. A comprehensive program is expected to employ analytical methods, visual inspections, experi-ence- feedback, and simulator deperidencies experiments. The light-water-reactor industry s current experience with systems interaction reviews is fragmented.

Experience like that gaine'd by the 'Phase I study -is an essential ingredient in the staff's considerations of a comprehensive systems interaction program.

After the resolution of USI A-17, the staff will determine whether RG8E must perform further evaluations for adverse systems interaction.

Although RG8E has not described a comprehensive program that separately eval-uates all structures, systems, and components important to safety for the three categories of adverse systems interactions (spatially coupled, functionally coupled, and humanly coupled), there is assurance that Ginna can be operated without endangering the health and safety of the public. The common-mode ef-fects of various postulated external events, as well as inplant failure effects, on safety-related structures, systems, and components have been extensively studied for the Ginna plant to ensure safe shutdown capability. These studies have been made both as a result of the Systematic'Evaluation Program (SEP) and the TMI Action Plan items. Areas most recently studied include the effects of seismic events, pipe breaks, internal and external 'flooding, wind and tornado loadings, internal missiles, and site hazards. In addition, the RGLE fire pro-tection study, together with the staff's proposed course of action, provides substantial assurance that separation and independence of safety-related systems at Ginna are provided.

The plant has been evaluated against current licensing requir'ements that are founded on the principle of defense in depth. Adherence to this principle re-sults in requirements such as physical separation and independence of redundant safety systems and protection against hazards such as high-energy-line ruptures, missiles, high winds, flooding, seismic events, fires, human factors, and sabo-tage. These design provisions are subject to review against the General Design Ginna SER C"15

Criteria (GDC) (10 CFR 50, Appendix A) that address some types of potential systems interacti,ons such as the spatially coupled interactions associated .with fires, floods, and high-energy-line,.breaks. Also, the quality assurance pro-gram, which is followed during the operational phase for each plant, contributes to the prevention of introducing .adverse systems interactions. Thus, the li-censing requirements and procedures have provided an adequate degree of plant safety pending identification of any new systems interaction by this task.

On the basis of, the above consideration, the staff has concluded that there is reasonable assurance that Ginna can continue to be operated until ultimate resolution of this issue without endangering the health and safety of the publ i c.

A-24 Environmental uglification of Safet -Related Electrical E ui ment The evolutionary -process of developing environmental qualification requirements and a.case-by-case implementation has resulted in a diversity of equipment, in-stalled in nuclear plants,.and different levels of documentation of the extent to which equipment is environmentally qualified. In an effort to further stand-ardize the qualification methods and documentation, Generic Task A-24 was de-veloped. Issuance of NUREG-0588 by NRC in July 1981 completed this unresolved safety issue. For operating reactors such as the Ginna plant, the Division of Operating Reactors (DOR) Guidelines, transmitted to RG&E by letter dated Febru-ary 15, 1980, provide the basis. for environmental qualification requirements.

By letter dated September 19, 1980, NRC transmitted a Revised Order for Modifi-cation of License directing that information regarding the environmental quali-fication of safety-.related, electrical equipment be submitted to the staff by November,l, 1980.

Franklin Research Center (FRC), under contract to NRC, reviewed the RG&E re-sponses and provided an assessment in a draft Interim"Technical Evaluation Report (TER) (FRC TER C5257-178), dated August 20, 1980. RG&E provided addi-tional information in a response, to the FRC report, by letter and report dated 31, 1980. Review of the additional information by FRC'esulted in a 'ctober Safety Evaluation Report (SER) forwarded by letter dated June 1, 1981, with FRC TER C5257-178 attached (March 18, 1980). RG&E's responses .to this SER, dated September 4, 1981, November 6, 1981, and February 18, 1982, resulted in the staff issuing a third SER forwarded by letter dated December 13, 1982.

The staff concluded in the December 13, 1982, SER that continued operation until completion of the licensee's env'ironmental qualification program will not pre-sent undue risk to the public health and safety. Furthermore, the staff has continued to review the licensee s environ'mental qualification program. If any additional qualification deficiencies are identified during the course of this review, the licensee wi,ll be required to reverify the justification for continued operation.

On February 22, 1983, the final Environmental qualification (Eg) rule became effective. The Eg rule in Paragraph 50.49(g) of 10 CFR 50 requires each holder of anoperating l,icense issued before February 22, 1983, to identify to the Commission by May 20, 1983, the electrical equipment important to safety which is already qualified and submit a schedule for completing final equipment quali-fication for, the remaining electrical equipment important to safety (within the Ginna SER C-16

I scope of the rule). gualification is to be completed by the end of the second refueling outage after March 31,, 1982, or by March 31, 1985, whichever is earlier By letter dated May 19,'983, RG&E provided the information required by the rule. The licensee stated that he will strive to complete the environmental qualification by the end of the 1984 refueling outage., If any unavoidable delays occur, the licensee will promptly notify the NRC and request approval, of the revised schedule.

On the basis of RG&E's commitment to meet the regulatory criteria provided in.

the rule and the conclusions reached in the SER dated December 13, 1982, the staff concludes that operation of the Ginna plant can continue without undue risk to the, health and safety of the public until completion of RG&E's environ-mental qualification program.

A-26 Reactor Vessel Pressure Transient Protection a

Over the years there have been several reported incidents of pressure transients in pressurized-water reactors that have exceeded the pressure/temperature limits of the reactor vessels involved. Most events occurred while the plant was in a solid-water condition, normally during startup or shutdown operations and at relatively low reactor vessel temperatures.

The causes of these overpressurizations were grouped into the following general categories: personnel error, procedural deficiencies, component random failures, and spurious valve actuation. The pressure transient was the result .of either a mass input (charging pumps, safety injection pumps, and accumulators) or a thermal expansion of the primary fluid typically from heat input from the steam generator.

In light. of the frequency of these transients and the decreasing reactor vessel toughness with age (because of increased neutron fluence), the NRC adopted this task so that methods to prevent and minimize, the effects of reactor vessel overpressurization could be developed.

RG&E installed a reactor vessel low-temperature overpre'ssure. protection system during the 1978 r'efueling outage. The Technical Specification changes, and the SER accepting this'ystem, were forwarded by letter dated April 18, 1979.

The implementation of the, resolution of this unresolved safety issue is con-sidered complete for th'e Ginna facility.

II g ~

A-31 Residual Heat Removal Re uirements The safe shutdown of a .nuclear power plant following an accident not related to a LOCA,has been typically interpreted as achieving .a "hot standby" condition (i.e., the reactor is shut down, but system temperature and pressure are still at or near. normal operating values). Considerable emphasis has been placed on the hot standby condition of a power plant ig the event of an accident or other abnormal occurrences. A similar, emphasis has been placed on long-ter~ cooling, which is achieved by the residual heat removal (RHR) system. The RHR system starts to oper'ate when the reactor coolant pressure and temperature are sub-stanti'ally lower thanthe hot standby. condition values., However, the review of the transient conditions of getting from hot shutdown to cold shutdown condi-tions was limited.

Ginna SER C-17

USI A-31, as implemented by SRP Section 5.4.7 and BTP RSB 5-1, requires that the plant can be taken from normal operating conditions to cold shutdown using only safety-related systems (assuming a single failure) and either onsite or offsite power through the use of suitable procedures. Safe shutdown, including maintenance of hot standby, cooldown, and cold shutdown operation,"was reviewed during the evaluation of SEP Topics V-10.B,, V-11.B, and VII-3. The review of this capability at Ginna was documented in NRC's SER forwarded by letter dated September 29, 1981. The staff found the Ginna plant has safety-related plant systems capable of safe shutdown under these conditions; however, the plant =

operating procedures rely on other nonsafety-related systems and do not specify how the cooldown woul'd be accomplished by the operator systems failed.

if nonsafety-related The licensee uses safety-related as well as nonsafety-related systems to bring the plant to a cold shutdown. Nonsafety-related systems may not be available following either a seismic or loss-of-power event., Lack of sufficient informa-tion in operating procedures may adversely affect the time required to shut down the reactor plant; for instance, failure .of the nonsafety-related air system could result in inoperability of the steam atmospheric dump valves.

Manual opening of the steam atmospheric dump valves would be required backup nonsafety-related nitrogen system also failed.

if the The licensee has proposed to develop appropriate documented procedures for operation of safety-related systems and components to achieve cold shutdown nonsafety-related systems are unavailable. The staff finds this acceptable.

if The procedures will be developed and implemented in coordination with Item 7 of Supplement.l to NUREG-0737, "Upgrade Emergency Operating Procedures."

A-36 Control of Heav Loads Near S ent Fuel Overhead handling systems (cranes) are used to lift heavy objects in the vicin-ity of spent fuel in pressurized-water reactors and boiling-water reactors. If a heavy object, e.g., a spent fuel shipping cask or shielding block, were to fall or tip onto spent fuel in the storage pool or the reactor core and damage the fuel, radioactivity could be released to the environment and radiation over-exposure of inplant personnel could occur. If many fuel assemblies were dam-aged, and the damaged fuel contained a large amount of undecayed fission pro-ducts, radiation releases to the environment could exceed 10 CFR 100 guidelines.

Additionally, a heavy object could fall on safety-related equipment and prevent it from performing its intended function. If equipment from redundant shutdown paths were damaged, safe shutdown capability might be defeated.

The NRC staff requested, by letter dated December 22, 1980, that licensees make a determination of the, extent to which the guidelines of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," are met. RG8E responded to this request by letter dated February 1, 1982. The auxiliary building crane movement over the spent fuel pool is limited by a system of electrical inter-locks except for a small portion of the southeast corner and a.narrow strip on the north side. Admini.strative procedures limit travel in these areas of the pool. A review by Franklin Research Center, draft TER C5257-444, transmitted by NRC letter of August 19; 1982,,currently is being evaluated by RG8E. In that report, FRC recommends additional administrative clarification of load-handling procedures and more explicit marking of load paths.

Ginna SER C"18

On the basis of the present, controls placed on movement of heavy loads at the Ginna plant, including the area in the vicinity of spent fuel, and the addi-tional'effort to be made in clarifying load paths and procedures, the staff concludes that this issue is being adequately addressed for the Ginna plant and that operation can continue without undue risk to the health and safety of the public.

A-40 Seismic Desi n Criteria - Short-Term Pro ram NRC regulations require that nuclear power structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes. Detailed requirements and guidance regarding the seismic design of nuclear plants are provided in the NRC regulations and in regulatory guides (RGs). However, a number of plants have construction permits and operat-ing licenses issued before NRC's current regulations and regulatory guidance were in place. For this reason, rereviews of the seismic design of various plants are being undertaken to ensure that these plants do not present an undue risk to the public. Task A-40 is, in effect, a compendium of short-term efforts to support such reevaluation efforts of the NRC staff, especially those related to older operating plants. In addition, some revisions to sections of the Standard Review Plan and to RGs to bring them more in line with the state-of-the-art will result.

The primary objective of the SEP seismic review was to make an overall safety assessment of the seismic capability of the existing plant and, if necessary, to modify the design to ensure the ability to shut down safely in the event of an earthquake. Current review criteria, as defined in the Standard Review Plan (SRP), and the criteria and guidelines developed for seismic review of older plants were used to assess safety margins. Conformance with the SRP would imply acceptability; however, significant differences in analysis methods and criteria were expected because these plants were originally designed to the criteria developed 10 to 15 years ago. As a result, the staff developed a more reasonable and realistic approach for reanalyses, including the use of ductil-ity reduction methods, nonlinear analysis methods, higher damping, and other factors identified in NUREG/CR-0098. The reanalyses performed as described would ensure an adequate seismic design.

The SEP seismic review addressed the safe shutdown earthquake (SSE) only, since it represents the most severe event that must be considered in the plant design.

The scope of the review included three major areas: the integrity of the reac-tor coolant pressure boundary, the integrity of fluid and electrical distribu-tion systems related to safe shutdown and engineering safety features, and the integrity and operability of mechanical and electrical equipment and engineer-ing safety feature systems (including containment). A detailed review of the facilities was not performed; rather the review relied on the sampling of representative structures, systems, and components. Confirmatory analyses using a conservative seismic input were performed for the sampled structures, systems, and components. The site-specific spectra supplied to all SEP owners by letter dated June 8, 1981, and more sophisticated analysis techniques were used if the conservative sample result indicated overstresses. The results of these analyses were reviewed by the NRC seismic review team. The results of that review for Ginna were published in NUREG/CR-1821.

Ginna SER C" 19

C The NRC seismic review team confirms that this issue is adequately addressed for the G'irina'plant.'h'e staff does not expect the results of Task A-40 to affect thes'e concl,u'si'ons because the techniques under consideration are essen-tially 'similar to 'those used in the seismic rereview of the Ginna plant as part of the"SEP pr'ogram." Should the resolution of Task A-40 indicate a change is needed in licensing requirements, all operating reactors, including Ginna, will be reevaluated on a case-by-case basis. Accordingly', the staff has con-cluded that this facility can continue to 'be operated until the ultimate reso-

'ution of this generic issue without endangering the health and safety of. the public.'

A-'43'" Conta'inment Eme'r enc Sum Reliabilit

'ollowiiig a postulated LOCA - a break in the react'or coolant system piping - the water flowing from th'e break would be collected in the containment emergency sump "'for"long-term recirculation through the reactor system by the emergency

'core co'oling (ECC), pumps to maintain core cooling. In addition, this water would be'recirculated 'through'he containment spray system to remove heat and "fiss'ion products"from within the containment. Loss of the ability to draw .

water 'from'the'containment emergency sump could significantly degrade long-term coolin'g 'and p'otentially disable the ECC systems.

Two major safety concerns have been postulated:

(1) adver'se post-LOCA'hydraulic conditions in the sump (e.g., high air inges-

-'ti'on, break 'impingement effects, and vortex formation) that could lead to

'los's'of"re's'i"dual hea't removal pump net positive, suction head (NPSH) t (2) sey'ere'sump "screen blockages resulting, from LOCA-generated debris and lead-

""'ng to 'loss of NPSH margin.

Full-seal'e 'sdmp 'hydraulic experiments (conducted as part of the USI A-43 resolu-tion" activit'ie*s); have shown that air ingestion levels are generally low (i. e.,

<2X):an'd th'at observation's of vortex'ormation are an unreliable indication of sump hydr'aulic'pe'r'formance. Similar generic 'studies and experiments addressing the debris blockage issues, on the other hand, have shown that the current 50K sump blockage criteria in RG 1.82, "Sumps for Emergency Core Cooling and Con-tainment"Spray-Systems" do not correctly (or adequately) address the blockage question'. 'n'addition, debris transport experiments have shown that transport of 'highly fi agmented '(or destroyed) insulation materials occurs at very low velo'cities (i'.e.," 0.2-0.5 ft/sec): Finally, assessment of debris blockage effects* has 'been'hown to be highly plant dependent because insulation types (and 'quantities), containment .layout, sump design, recirculation flow require-ments, And'"pump NPSH;requirements all inter relate to determine NPSH impact.

i'l k. <~

The staff's'technical evaluations regarding the 'above matter are preserited in NUREG-0897 'the"re'vised RG 1.82 is in NUREG-0869. Both these NUREG reports wer'e i'ssued "fo'r comment" in'April 1983. (The proposed resolution for A-43 rec'ommends"'issuance of a generic letter to all operating plants and operating license 'applicants -requesting an assessment of sump design adequacy and confir-matio'n of"NPSH margins'sing the methodology set forth in RG 1.82, Revision 1.)

The licensee has been provided the reports noted above, along with related NUREG/CR reports.

Ginna SER C-20"'

The staff has also performed a probabilistic assessment, of recirculation sump, blockage resulting from a LOCA (NUREG/CR-3394). This study'ndicated that on1y pipe breaks more 'than about 10 i~. in diameter generated; sufficient'.debri's block the containment. The Commission currently "is'valuating to','otentially whether to require backfitting in operating licenses of, the proposed require-ments in RG 1.82 because of the 'relatively low probability of 1arge-break'LOCAs and sump blockage. Consequently, the staff has concluded that there is'eason-able assurance that Ginna can continue operation until this generic issue, has .

been ultimately resolved without endangering the health and safety'f;the public.

A-44 Station Blackout Electrical power fo'r safety systems at nuclear power plants must be, supplied by at least two redundant and independent divisions. The systems 'used to remove decay heat to cool the reactor core following a reactor shutdown are included among the safety systems that must meet these requirements.'ach electrical division for safety systems includes an offsite power'connecti'on, a standby emergency diesel generator ac power supply, and dc sources'.

k Task A-44 involves a study of whether or not nuclear power jl,ants'ho'uld be designed to accommodate a complete loss of all ac power (i.e.",'o'ss of both offsite and the emergency diesel generator ac 'power suppli'e's). This issu'e 'he arose because of operating experience regarding the reliability of ac power supplies. A number of operating plants have e'xperienced a total loss of off-electrical power, and more occurrences are expected.'" In'almost 'every 'ite one of these loss-of-offsite-power events,'the onsite emergency ac power sup-plies were available immediately to supply the power needed by"-v'ital 'safety equipment. However, in some instances, one of the redundant "emergency'power supplies was unavailable. In' few cases there has been a complete l.oss of ac power, but during these events, ac power was restored in less th'an a"few minutes without any serious consequences. In addition, there have been numei ous reports of emergency diesel generators failing to start and run in'"operatijg"plants" dur-ing periodic surveillance tests.

A loss of all ac power was not a design-basis ev'ent for Ginna.'onethe'less,'

combination of design, operation, and testing requirements that have been im-posed at the plant will ensure that this unit will have substantial'resi'stan'ce to a loss of all alternating current and that, even if a loss'f "all alternat-ing current should occur, there is reasonable assurance that decay heat will be removed by the steam-driven auxiliary feedwater system and the'core wi-ll be

'.'ooled.

3 The issue of ac power dependence of the auxiliary feedwater '(AFW) 'system'as "

considered both as a TNI item and in SEP Topic X. The Ginna design includes' 200K steam-driven AFW pump as well as four lOOX"motor-dr'iven AFW'pumps'. 'The staff has concluded that, on the basis of system design and te'sting,'he'team-driven AFW pump'ould perform its safety function without rely'ing on ac" power.'he issue of onsite and offsite'power reliability was also exterisively r'eviewed during the conduct of the SEP. It was concluded during the'review of SEP Topics VIII-1.A, "Potential Equipment Failure Associated With a Degraded Grid Voltage,", and VIII-2, "Onsite Emergency Power Systems - Diesel'e'n'erators';"

'the Ginna onsite and offsite ac power systems meet;current criteria." A'Iso, "'hat Ginna SER C-21

during the review of SEP Topic VIII-3.A, "Station Battery Capacity Test Require-ments," it was concluded that if both the onsite and offsite ac power simultaneously, the 8-hour capacity of the onsite batteries's sufficient systems'ail to ensure adeq'uate dc power to the station until" ac power can be restored.

Emergency Procedure E-" 4. 3, "Loss of A. C. Power," has been'eveloped to detail the required actions.

The staff, therefore, concludes that the issue of station blackout 'is being adequately addressed for the Ginna plant and that=operation can continue with-out undue risk to the health and safety of the public until ultimate resolution of this issue.

A-45 Shutdown Deca Heat Removal Re uirements Under normal operating conditions, power generated within a reactor is removed as steam to produce 'electricity by means of' turbine generator. Following a reactor shutdown, a reactor produces insufficient power to operate the turbine; however, the radioactive decay of fission products continues to produce'heat (so-called "decay heat"). Therefore, when reactor shutdown occurs, other meas-ures must be available to remove decay heat from the reactor. All light-water reactors share two common decay heat removal functional requirements 'to (1) provide a means of transferring decay heat from the reactor coolant system to an ultimate heat sink (2) maintain sufficient water inventory inside the reactor vessel to ensure

'adequate cooling of the reactor fuel The reliability of a particular power plant to perform these functions depends on the frequency of initiating events that require or jeopardize decay heat removal operations'nd the. probability that required systems will respond to remove the decay heat.

The principal means for removing the decay heat in a pressurized-water reactor (PMR) under normal conditions immediately following reactor shutdown is through

~

the steam generators by using the auxiliary feedwater (AFM) system. Following the TMI-2 accident, the NRC staff required plants to make improvements to the AFW systems. The staff also believes that providing an alternative means of decay heat removal could substantially increase the plant's capability to deal with a broader spectrum of transients and accidents and, therefore, could potentially significantly reduce the overall risk to the public. Consequently, alternative means of decay heat removal in PWR plants, including but not limited to the use of existing equipment where possible, will be investigated under this unresolved safety issue. This study will consist of a generic systems evaluation and wi 11 result in recommendations regarding (1) the adequacy of existing shutdown de'cay heat removal requirements and (2) the desirability of and possible design requirements for an alternative decay heat removal method, that is, a method other than that normally associated with the steam generator and secondary system.

The design and qualification of the AFM system were reviewed both as part of the TMI review and as part of the review of SEP Topics X, "Auxiliary Feedwater Systems," and V-10.B, "RHR Reliability." The present AFW system consists of Ginna SER C"22

two 100K motor-driven AFW pumps, a 200K steam-driven AFW pump (independent of ac power), and two 100K motor-driven standby AFW pumps. The motor-driven

'pumps normally:.take suction from onsite condensate storage tanks, but also get water from the service water system (Lake Ontario). Furthermore, a modifica-tion made, during the SEP review of,Ginna provided for connections that allow the use of the yard fire hydrant system (independent of onsite or offsite power) as a source of water for the motor-driven and steam-driven pumps. A similar modification was made for the standby AFW pumps. It is apparent that many di-verse means of water supply to and from the AFW systems are available at Ginna.

During the course of the Appendix R Fire Protection reviews, RG&E identified a means of going from hot shutdown to cold shutdown, conditions that could include filling of the steam generators and steamlines with water and using them, as water-to-water heat exchangers. This method has been accepted as viable'y the staff as discussed in Section 9.6.

Other means of removing decay heat have also been investigated. They are described in NRC's "Safe Shutdown Evaluation" for the Ginna SEP (letter dated September 29, 1981). These include use of the chemical and volume control system, RHR system, and steam generator blowdown systems and potentially the "bleed-and-feed" method, which uses the pressurizer power-operated relief valves and the,.

safety injection pumps.

Given'he extensive diversity and capacity of the Ginna AFM systems and the, other methods available for alternative decay heat removal, the staff concludes that until final resolution,.of this generic issue, the Ginna plant can continue operation without undue risk to the health and safety of the. public.

A A-46 Seismic uglification of E ui ment in 0 eratin Plants.

The design criteria, and methods for the seismic qualification of mechanical and electrical equipment in nuclear power plants have undergone significant, changes during the course of the, commercial nuclear power program. Conse-quently, the margins of safety provided in existing equipment to, resist seis-mically induced loads and perform intended safety functions may vary consider-ably among plants licensed in different time frames. The staff has determined that the seismic qualification of the equipment in operating plants should be reassessed to ensure the ability of the equipment to perform its designed safety functions during and/or after a seismic event. The objective of USI A-46 is to establish explicit guidelines that can be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at all operating plants instead of attempting to backfit current design criteria. This guidance will concern equipment, required to perform a safety function, as well as equip-ment that is not required to perform a safety function, but;whose failure could result in adverse conditions that might impair the safety functions of other equipment or systems'he Systematic Evaluation Program, through the Senior Seismic Review Team, performed an audit of the Ginna safety-related structures, systems, and compo-nents. The results of the audit are provided in NUREG/CR-1821. Most equip-ment was found to be capable of withstanding the Ginna safe shutdown earthquake.

In certain areas sufficient documentation was not available. Reanalysis and, in some cases, redesign or resupport are being conducted-. The status of these items is provided jn NUREG-0821 (November 1982).

Ginna SER C-23

S In addition, the anchorage of major equipment was addressed. Experience from major earthquakes has shown 'that 'almost all seismically induced equipment failures in quality industrial facilities have occurred because the components were not adequately anchored to their foundations and that few equipment 'fail-ures have occurred in equipment that was anchored. As a result of the review of 'electrical equipment anchorage, modifications to upgrade the anchorages of a number of safety-related electrical components at Ginna were made.

ll RGRE has also been a participant in a Seismic qualification Utility Group (SHRUG) that has completed a pilot program to explore a'n alternative method for seis-

, mically qualifying selected nuclear plant components based on experience with the equipment during earthquakes. Additional work on this program now underway, is expected to provide for (1) the developing of qualification methodology for installed equipment at operating plants, (2) the screening and assigning priorities for more efficient use of NRC and industry resources, of'ualification and (3) possibly the qualifying of certain 'classes of equipment on a generic basis without specific testing or analyses of the components.

On the'asis of the above discussion, the staff concludes that this issue is being adequately addressed and that operation of the Ginna plant can continue without undue risk to the health and safety of the public.

A-47 Safet Im lications "of Control S stems This issue concerns the potential for transients or accidents being made more severe as a result of control system failures or malfunctions. These failures or malfunctions may occur independently or as a result of the accident or tran-sient under consideration. One concern is the potential that a single failure such as a loss of a power supply, short circuit, open circuit, or sensor fai lure could cause simultaneous malfunction of several control features. Such an oc-currence could conceivably result in a transient more severe than those tran-sients analyzed as anticipated operational occurrences. A second concern is that a postulated accident could cause control system failures that could make the accident more severe than analyzed. Accidents could conceivably cause con-trol system failures by creating a harsh environment in the area of the control equipment or by physically damaging the control equipment. -It is generally believed by the staff that such control system failures would not lead to se-rious events or result in conditions that safety systems cannot safely handle.

Systematic evaluations on all nonsafety systems, h'owever, have not been rigor-ously performed to verify this belief. The potential for an accident that could affect a p'articular control system, and effects of the control system failures, may differ from plant to plant. Therefore, it is not possible to develop ge-neric answers to these concerns, but rather plant-specific evaluations are required. The purpose of this USI is to'erify the adequacy of the existing criteria for control systems and, if necessary, develop and propose additional criteria or guidelines to improve system reliability and enhance safety.

Ginna's control and safety systems have been designed to ensure that control system failures will not prevent automatic or manual initiation and operation of any safety-system equipment required for accident mitigation and/or to main-tain the plant in a safe shutdown condition following any anticipated opera-tional occurrence or "accident." This has been accomplished by providing inde-pendence between safety-system trains and between safety and nonsafety systems.

Ginna SER C-24

For the latter, as a minimum, isolation devices were provided;" These devices preclude the propagation of nonsafety-related equipment faults'o the protection systems. Also, to ensure that the operation of safety-'relat'ed equipment is not impaired, the single-failure criterion has been applied in the plant design of the protection systems. SEP Topics VI-7.A.3, VI-7.C.2, VII-1.A, VII-2, and VII-3 address elements of this issue.

A systematic evaluation of the control system design, as contemplated for this unresolved safety issue, has not been performed to determine whether postulated accidents could cause significant control system failures that would make the accident consequences more severe than currently analyzed. However, a wide range of bounding transients and accidents is being analyzed currently to ensure that the postulated events, such as steam generator overfill'and overcooling events, would be adequately mitigated by the safety systems. In addition, re-views of safety systems were performed w'ith the goal of ens'uring that control system failures will not defeat safety-system action.

Additional studies probing the interaction of safety and nonsafety systems were performed during Ginna's fire protection reviews in'esponse to 10 CFR 50, Appendix R. Within designated fire zones, it'as assumed that damage to any equipment '(or its control cables, if affected) could cause failure of any type.

The dedicated shutdown system proposed by RG&E as a result of the fire protec-tion study will incorporate the required separation of safety and nonsafety systems.

Also, the licensee has been required (NRC Information Notice 79-22) to review the possibility of consequential control system failures'hat exacerbate- the effects of high energy line breaks (HELBs) and adopt new operator procedures, where needed, to ensure that the postulated events would be mitigated. RG&E performed an evaluation of those potential harsh environment effects. By let-=

ter dated October 5, 1979, RG&E concluded that none of the scenarios identified in the information notice constituted potential failure modes that could com-promise a safe shutdown of the Ginna plant.

The staff is also evaluating the qualification program to ensure that equipment that may be exposed to HELB environments has been adequately qualified or an adequate basis has been provided for not qualifying the equipment to the limit-ing hostile environment. The status of this review is contained in the discus-sion of USI A-24.

In addition, IE Bulletin 79-27 was issued to the licensee requesting that eval-uations be performed to ensure the adequacy of plant procedures for accomplish-ing shutdown on loss of power to any electrical bus supplying power for instru-ments and control. The licensee responded to this IE bulletin by letter dated February 26, 1980. The staff reviewed RG&E's submittal and concluded that the response and design were acceptable (memorandum dated June 22, 1982).

Another potential control and safety-system interaction addressed in RG&E's response to NRC's September 16, 1980 letter was the loss of dc sources and inverters. RG&E's instrumentation bus and power supply arrangement is such that loss of any dc source or inverter would not result in the loss of any instrument buses because the instrument buses would be automatically transferred to and powered from Class 1E constant voltage transformers. This is decribed in RG&E's letter of October 9, 1981.

Ginna SER C-25

On the basis of above considerations, and subject to satisfactory resolution of the Ginna equipment qualification program, the staff has concluded that there is reasonable assurance that Ginna can continue to be operated until the ulti-mate resolution of this generic issue without endangering the health and safety of the public.-

A-49 Pressurized Thermal Shock The issue of pressurized thermal shock (PTS) arises because in PWRs transients and accidents can occur that result in severe overcooling (thermal shock) of the reactor pressure vessel, concurrent with or followed by repressurization.

In these PTS events, rapid cooling of the reactor vessel internal surface results in thermal stress with a maximum tensile stress at the inside surface of the vessel. The magnitude of the thermal stress depends on the temperature profile across the reactor vessel wall as a function of time. The effects of this thermal stress are compounded by pressure stresses.

Severe reactor system overcooling events simultaneous with or followed by pres-surization of the reactor vessel (PTS events) can result from a variety of causes. These include system transients, some of which are initiated by in-strumentation and control systems malfunctions (including stuck-open valves in either the primary or secondary system), and postulated accidents such as small-break LOCAs, main steam line breaks (MSLBs), and feedwater line breaks.

The PTS issue is a concern for PWRs only after the reactor vessel has lost its fracture, toughness properties and is embrittled by neutron irradiation. The standards and regulatory requirements to which the Ginna reactor vessel was designed and fabricated are described in Section 5.3 of this SER.

As long as the fracture, resistance of the reactor vessel material is relatively high, overcooling events are not expected to cause vessel failure. However, the fracture resistance of reactor vessel materials decreases with exposure to fast neutrons during the life of a nuclear power plant. The rate of decrease is dependent on the metallurgical composition of the vessel walls and welds.

If the fracture resistance of the vessel has been reduced sufficiently by neu-tron irradiation, severe overcooling events could cause propagation of small flaws that might exist near the inner surface. The assumed initial flaw might be enlarged into a crack through the vessel wall of sufficient extent to threaten vessel integrity and, therefore, core cooling capability.

For the reactor pressure vessel to fail and constitute a risk to public health and safety, a number of contributing factors must be present. These factors are (1) a reactor vessel flaw of sufficient size to initiate and propagate; (2) a level of irradiation (fluence) and material properties and composition sufficient to cause significant embrittlement (the exact fluence depends on materials present; that is, high copper content causes embrittlement to occur more rapidly); (3) a severe overcooling transient with pressurization; and (4) the crack resulting from the propagation of initials cracks of such size and location that the vessel fails.

As a result of the evaluation of the PTS issue, the staff recommended to the Commission in SECY-82-465 (November 23, 1982) actions to prevent PTS events in operating reactors. The Commission accepted the staff recommendations and Ginna SER C-26

h has directed the staff to develop a Notice of Proposed Rulemaking that would establish an RTNDT, screening criterion (below which PTS risk is considered acceptable), require licensees to submit present and projected values of RTNDT,

'equire early analysis and implementation of such flux reduction programs. as are reasonably practicable to avoid reaching the screening criterion, and require plant-specific PTS safety analysis before plants are within three calendar'"years of, reaching the screening cri'terion, including analyses, of pro-posed alternatives to minimize'the PTS problem.

Such a rule is now being developed by the staff. On the basis of its review of material properties at the Ginna facility, the staff believes that the Ginna plant could meet the requirements of such a rule.

On the basis of the above consideration, the staff 'concludes that the Ginna facility can continue to be operated before complete resolution of this issue and completion of the proposed rulemaking without undue risk to the health and safety of the public.

C.5 References Atomic Safety and Licensing Appeal Board, ALAB-444, 6'RC 760, "Gulf States Utilities Co., River Bend, Units '1 and 2," Nov. 23, 1977.

,---, ALAB 491, 8 NRC 245, Virginia Electric and Power Company (North Anna Nuclear Power Station, Unit Nos. 1 and 2)," Aug. 25, 1978.

Code of Federal Re ulations, Title 10 (10 CFR), "Energy," U.S. Government Printing Office, Washington, D.C. (includes General Design Criteria).

Federal Re ister, 45 FR 65474, "Severe Accident Rule," NRC, Oct. 2, 1980.

---, 45 FR 73080, "Electric Utilities; Filing of Petition for Rulemaking," NRC, Nov. 4, 1980.

---, 46 FR 62281, "Hydrogen Control for Mark III/Ice Condenser Containment and Dry Containment Analysis for Operating Plant and Plants Undergoing OL Review,"

NRC, Dec. 23, 1981.

Franklin Research Center (FRC), FRC TER C5257-178, "Equipment.,Environmental gualification," C. Crane et al., Mar. 1980.

---, FRC TER C5257-178, "Draft Interim Technical Evaluation Report," C. Crane, et al., Aug 20, 1980.,

---, FRC Draft TER C5257-444, ".Control of Heavy Loads," T. Hofkin, Aug. 10, 1982.

Letter, Dec. 30, 1977, from J. Hendrie (NRC) to Congress,

Subject:

NRC Progr am for the Resolution of Generic 'Issues Related to Nuclear Power Plants.

--", June 26, 1978, from L. White (RG&E) to D. Ziemann (NRC),

Subject:

, Steam Generator and Reactor Coolant Pump Support Material.

Ginna SER C-27

---, Apr. 18, 1979, from D. Ziemann (NRC) to L. White (RGBE),

Subject:

Reactor Cool ant System Over press uri zati on Protecti on.

---, Oct..5, 1979, from K. Amish (RG8E) to V. Stello (NRC),

Subject:

Potential Unreviewed Safety question on Interaction Between Non-Safety Grade Systems and Safety Grade Systems..

---, Dec. 20, 1979, from D. Ziemann (NRC) to L. White (RG8E),

Subject:

Steam Generator Technical Evaluation, Ginna Power Station.

---, Feb. 15, 1980, from D. Ziemann (NRC) to L. White (RG8E),

Subject:

Electrical Equipment Environmental qualification.

---, Feb. 22, 1980, from J. Angelo (NRC) to S. Hanauer (NRC),

Subject:

Summary of 2/20/80 Meeting with Sandia Labs and ACRS re Phase 1 Work on System Interaction.

---, Feb. 26, 1980, from L. White (RG8E) to B. Grier (NRC),

Subject:

IE Bulletin No. 79-27, Loss of Non-Class lE Instrumentation and Control Power System Bus During Operation.

---, Mar. 6, 1980, from D. Ziemann (NRC) to L. White (RG8E),

Subject:

Environ-mental qualification of Electrical Equipment.

---, Mar. 28, 1980, from D. Ziemann (NRC) to L. White (RG8E),

Subject:

Environ-mental qualification of Electrical Equipment.

---, May 19, 1980, from D. Eisenhut (NRC) to All Applicants and Licensees, Sub-ject: Requests Comments on NUREG-0577.

---, May 20, 1980, from D. Eisenhut (NRC) to All Applicants and Licensees, Sub-ject: Requests Comments on NUREG-0577.

---, Sept. 16, 1980, from D. Crutchfield (NRC) to J. Maier (RG8E),

Subject:

Request for Additional Information on SEP Topics VI-7.B, YI-7.C, YI-10, YII-3, VIII-2, VIII-3 (R. E. Ginna Nuclear Power Plant).

---, Sept. 19, 1980,'from D. Eisenhut (NRC) to L. White (RG8E),

Subject:

En-vironmental qualification of Electrical Equipment.

---, Oct. 31, 1980 from J. Maier (RG8E) to D. Eisenhut (NRC),

Subject:

Environ-mental qualification of Electrical Equipment.

---, Dec. 22, 1980, from D. Eisenhut (NRC) to All Licensees and Applicants,

Subject:

Control of Heavy Loads.

---, June 1, 1981', 'from D.'rutch'field (NRC) to J. Maier (RG5E),

Subject:

En-vironmental qualification of Safety-Related Electrical Equipment.

-'"-, June 8, 1981', from'0.'.Crutchfield (NRC) to.All SEP Owners,

Subject:

Site Specific Ground 'Response Spectra for SEP Plants 'Located in the Eastern United States.

Ginna SER C-28

---, Sept. 4, 1981, from J. Maier (RG8E) to D. Crutchfield (NRC);

Subject:

Environmental Qualification of Safety-Related, Equipment.

---, Sept. 29, 1981, from D. Crutchfield (NRC) to J. Maier (RG8E),

Subject:

Ginna-SEP Topics V-lO.B, RHR Systems Reliability,, V-11.B, RHR Interlock Require-ments, and VII-3,, Systems Required for Safe Shutdown (Safe Shutdown Systems Report).

---, Oct. 6, 1981, from J. Maier (RG8E) to D. Crutchfield (NRC),

Subject:

Preliminary Results of Charpy Impact Test on Specimens From Ginna's Reactor Vessel Surveillance Capsule T.

---, Oct. 9, 1981, from J. Maier (RG8E) to D. Crutchfield (NRC),

Subject:

SEP Topics VI-7.B; VI-7.C, VI-10, VII-3, VIII-2, VIII-3 (R. E. Ginna Nuclear Power Plant).

---, Nov. 6, 1981, from J. Maier (RG8E) to D. Crutchfield (NRC),

Subject:

En-vironmental Qualification of Electrical Equipment.

---, Feb. 1, 1982, from J. Maier (RG8E) to D. Eisenhut (NRC),

Subject:

Control of Heavy Loads.

---, Feb. 18, 1982 from J. Maier (RGLE) to D. Crutchfield (NRC),

Subject:

Environmental Qualification of Electrical Equipment.

---, May 3, 1982, from D. Crutchfield (NRC) to J. Maier (RG8E),

Subject:

SEP Topic V-6, Reactor Vessel Integrity, R. E. Ginna.

---, June 10, 1982, from J. Maier (RG8E) to D. Crutchfield (NRC),

Subject:

Environmental Qualification of Safety-Related Electrical Equipment.

---, Aug. 19, 1982, from D. Crutchfield (NRC) to J. Maier (RG8E),

Subject:

Control of Heavy Loads - NUREG-0612 Ginna.

---, Oct. 19, 1982, from J. Maier (RG8E) to D. Crutchfield (NRC),

Subject:

Unresolved Safety Issue Status R. E. Ginna Nuclear Power Plant Docket No. 50-244.

---, Dec. 8, 1982, from J. Maier (RG8E) to D. Crutchfield '(NRC),

Subject:

Analysis of Reactor Vessel Surveillance Capsule T.

---, Dec. 13, 1982, from D. Cr utchfield (NRC) to J. Maier (RGLE),

Subject:

Safety Evaluation Report For Environmenta) Qualification of Safety-Related Electrical Equipment."

---, May 19, 1983, from J. Maier (RGBE) to D. Crutchfield (NRC),

Subject:

10 CFR '50;49; Environmental Qual'ification of Electrical .Equipment.

---, July 8, 1983, from D. G. Eisenhut,(NRC) to. All Licensees of,.Operating

'eactors,-Applicants'foi Oper at'ing Licen'ses', and Holdersof Construction Permits,

Subject:

Required Actions '"Based on Generi'c Implications of Salem ATQS Events (Generic Letter 83-28).

Ginna SER ,C-29 P4

---, July 13, 1983, F. Miraglia (NRC) to D. Eisenhut (NRC),

Subject:

Summary of Meeting With Steam Generator Owners Group (SGOG) on July 6, 1983.

Memorandum, June 22, 1982, from D. Eisenhut (NRC) to Regional Administrators (NRC),

Subject:

Completion of NRR Review of Responses to IE Bulletin 79-,27, "Loss of Non-Class 1E Instrumentation and Control Power System Bus During, Operation (MPA B-61)."

Public Law (PL)95-209, "NRC Authorizations for FY 1978" (91 Stat. 1481).

U.S. Atomic Energy Commission, MASH-1270, "Technical Report on Anticipated Transients Without Scram for Water-Cooled Power Reactors," Sept. 1973.

U.S; Nuclear Regulatory Commission, NUREG-0410, "NRC Program for the Resolution of Safety Issues Related to Nuclear Power Plants,",Jan. 1, 1978.

'--, NUREG-0460, "Anticipated Transients Without Scram for Light-Water Reactors,"

Vols. 1 and 2, Apr. 1978; Vol. 3, Dec.'978; Vol. 4, Mar. 1980.

~ ---, NUREG-0510, "Identification of Unresolved Safety Issues Relating to Nuclear Power Plants - A Report to Congress," Jan. 1979.

---, NUREG-0516, "U.S. Nuclear Regulatory Commission, 'Annual Report 1978,"

Feb. 1978.

- , NUREG-0569, "Evaluation of the Integrity of SEP Reactor Vessels," Dec. 1979.

---, NUREG-0577, "Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports," Oct. 1979; Rev. 1, Oct. 1983.

---, NUREG-0582, "Waterhammer in Nuclear Power Plants," July 1979.

---, NUREG-0588, "Interim Staff Position on Environmental, Qualification of Safety-Related Electrical Equipment," July 1981.

---, NUREG-0606, "Office of Nuclear Reactor Regulation Unresolved Safety Issues Summary," Aqua Book, issued quarterly.

---, NUREG-0609, "Asymmetric Blowdown Loads on PWR Primary Systems, Resolution of Generic Task Action Plan A-2,". Jan. 1981.

---, NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants - Resolution of Generic Technical Activity A-36," July 1980.

---, C NUREG-0649, "Task'Action Plan for i6 l Unr'esolved Safety Issues Related to Nuclear Power Plant's," Feb. 1980.

---, NUREG-0651, "Evaluation of Steam Generator Tube Rupture Events," Mar. 1980.

---, NUREG-0660,'NRC Action Plan Developed as a Result of; the TMI-2,Accident,"

Vol. I, May 1980.

Ginna SER C".30

---, NUREG-0705, "Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants, Special Report to Congress," Mar. 1981.

---, NUREG-0737, "Clarification of TMI Action Plan Requirements: Requirements for Emergency Response Capability," Suppl. 1, Jan. 1983.

---, NUREG-0800 (formerly NUREG-75/087), "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," July 1981 (includes Branch Technical Positions).

---, NUREG-0821, "Integrated Plant Safety Assessment - Systematic Evaluation Program - R. E. Ginna Nuclear Power Plant," draft, May 1982; final, Nov. 1982; Suppl. 1, Aug. 1983.

---, NUREG-0844, "Resolution of Unresolved Safety Issues A-3; A-4, and A-5 Regarding Steam Generator Tube Integrity," Dec. 1982.

I

---, NUREG-0869, "USI A-43 Resolution Positions," Apr. 1983.

---, NUREG-0897, "Cont'ainment Emergency Sump Performance. Technical Findings Related to Unresolved Safety Issue A-43," Apr. 1983.

---, NUREG-0909, "NRC Report on the January 15, 1982 Steam Generator Tube Rup-ture at R. E. Ginna Nuclear Power Plant," Apr. 1982.

---, NUREG-0916, "Safety Evaluation Report Related to the Restart of R. E. Ginna Nuclear Power Plant," May 1982.

---, NUREG-0927, "Evaluation of Water Hammer Experience in Nuclear Power Plants," May 1983.

---, NUREG-1000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant," Vol. 1, Apr. 1983; Vol. 2, Aug. 1983.

---, NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected Nuclear Power Plants," June 1978.

---, NUREG/CR-1717, "Soil-Structure Interaction Methods," Vol. I, Dec.,1980.

---, NUREG/CR-1821, "Seismic Review of the Robert E. Ginna Nuclear Power Plant as Part of the Systematic Evalua'tion Program,"" Dec. 1980.

---, NUREG/CR-2403, "Survey of Insulation Used in Nuclear Power Plants and 'the Potential for Debris Generation," Suppl. 1, May 1982.

---, NUREG/CR-2569, "Response of Zion and Indian Point Containment Buildings to Severe Accident Pressures," May 1982.

- , NUREG/CR-2792, "An Assessment of Residual Heat Removal and Containment Sump Performance Under Air and Debris Ingestion Conditions," Sept. 1982.

---, NUREG/CR-3394, "Probabilistic Assessment of Recirculation Sump Blockage Due to Loss-of-Coolant Accidents," Vols. 1 and 2, July 1983.

Ginna SER C-31

---, Regulatory Guide (RG) 1.7, "Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant Accident."

---, RG 1.61, "Damping Values for Seismic Design of Nuclear Power Plants."

---, RG 1.82, "Sumps for Emergency Core Cooling and Containment Spray Systems."

---, RG 1.82, "Water Sources for Long-Term Recirculation Cooling Following LOCA," Draft Rev. 1, Aug. 1983.

---, RG 1.83, "Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," Rev. 1.

---, RG 1.99, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," Rev. 1.

---, RG l. 121, "Basis for Plugging Degraded Pressurized Water Reactor Steam Generator Tubes," for comment, Aug. 1976.

---, SECY-82-465," Pressurized Thermal Shock (PTS)," Nov. 23, 1982.

---, "Status Report for Anticipated Transients Without Scram for Westinghouse Reactors," Dec. 9, 1975.

U.S. Nuclear Regulatory Commission, Office of Inspection and Enforcement (IE),

Bulletin 79-27, "Loss of Non-Class lE Instrumentation and Control Power Systems During Operation," Nov. 30, 1979.

Information Notice 79-22, "Potential Unreviewed Safety questions on Inter-action Between Non-Safety-Grade Systems and Safety-Grade Systems," Feb. 19, 1982.

Westinghouse, "Anticipated Transients Without Scram for Westinghouse Plants,"

NS-TMA-209b, June 8, 1979 and Dec. 30, 1979.

---, Topical Report, WCAP 8404, "ATWS Analysis for Westinghouse PWR's With 44 Series Steam Generators," Sept. 1974.

---, WCAP 9558, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Con-taining Postulated Circumferential Throughwall Crack," through Rev. 2, May 1981.

WCAP 9570, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing Postulated Circumferential Throughwall Crack," Oct'. 1979 and June 1980.

---, WCAP 9628, "Westinghouse Owners Group Asymmetric LOCA Loads Evaluation,"

Nov. 1979.

'I WCAP 9662, "Westinghouse Class 3, Westinghouse Owners Group Asymmetric LOCA Loads Evaluation," Jan. and Feb. 1980.

---, WCAP 9748', "Westinghouse Owners Group Asy'mmetric LOCA Load Evaluation, Phase C," June 1980.

Ginna SER C-32

- , WCAP 9749, "Westinghouse Owners Group Asymmetric LOCA Load Evaluation, Phase C," June 1980.

---, WCAP 9787, "Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation," through Rev. 1, May 1981.

Ginna SER C-33

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NRC FORM 335 1. REPORTNUMBER (Assignedby DDCJ I7 771 U.S. NUCLEAR REGULATORY COMMISSION BIBLIOGRAPHIC DATA SHEET NUREG-0944

4. TITLE AND SUBTITLE (Add Volume No., ifappropriaieJ 2. (Leave blank/

Safety Evaluation Report related to a Full-Term Operating License for the R.E. Ginna Nuclear 3. RECIPIENT'S ACCESSION NO.

Power Plant

7. AUTHOR(s) 6. DATE REPORT COMPLETED MON TH YEAR October 1983

9. PERFORMING ORGANIZATION NAME AND MAILINGADDRESS (Include Zip Codei DATE REPORT ISSUED MONTH YEAR Division of Licensing October 1983 Office of Nuclear Reactor Regulation 6. (Leave blankJ U.S. Nuclear Regulatory Commission Washington, DC 20555 B. (Leave blankJ

12. SPONSORING ORGANIZATION NAME ANO MAILING ADDRESS (include Zip Codei

10. PROJECT/TASK/WORK UNIT NO.

Same as 9 above 11. CONTRACT NO.

13. TYPE OF REPORT pERloo covEREo (Inclusive dares(

15. SUPPLEMENTARY NOTES 14. (Leave olankl Do ket No..50-244

16. ABSTRACT f200 words or lessi The Safety Evaluation Report for the full-term operating license application filed by Rochester Gas and Electric Corporation for the R.E. Ginna Nuclear Power Plant has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is located in Wayne County, Rochester, New York. Subject to favorable resolution of the items discussed in this report, the staff concludes that the facility can continue to be operated without endangering the health and safety of the public.

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS 17b. IOENTIFIERS/OPEN ENDED TERMS IB. AVAILABILITY STATEMENT 19. SECURI TY CLASS (This reporsl 21, NO. OF PAGES uncl assi fied Unlimited 20. SECURITY CLASS JThis pagei

'unc"I ass1h ea 22. PRICE S

NRC FORM 335 17 77)