Text

Requests Comments on Proposed Scope of SEP Phase III as Reflected in Encl Topic Definitions
	ML20027B595
Person / Time
Site:	Millstone, Dresden, Palisades, Oyster Creek, Haddam Neck, Ginna, San Onofre, Yankee Rowe, La Crosse, Big Rock Point
Issue date:	09/09/1982
From:	Eisenhut D; Office of Nuclear Reactor Regulation
To:	Denise Edwards; ATOMIC INDUSTRIAL FORUM
Shared Package
ML20027B596	List: ML20027B595; ML20027B597; ML20027D214;
References
	NUDOCS 8209280341
	Download: ML20027B595 (78)
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NUCLEAR REGULATORY COMMISSION WASHING TON, D. C. 20555 sn..t September 9, 1982 4

i Mr. D. Edwards, Chairman AIF Subcommittee on Backfitting Atomic Industrial Forum 7101 Wisconsin Avenue Washington, D.C.

20014

Dear Mr. Edwards:

SUBJECT:

PROPOSED SEP PHASE III TOPIC DEFINITIONS The staff has recently requested comments from the Systematic Evaluation Program (SEP) Phase II licensees and the Atomic Industrial Forum relative.

to the proposal the staff is developing for a continuation of SEP (Phase III). Most of the comments that we received expressed a concern regarding the proposed scope of the program.

Consequently, we would like your comments on the proposed scope of the l

program, as reflected in the enclosed topic definitions. The topic defini-tions also include the safety objective, NRC criteria for the review, review guidelines, and the safety basis including Phase II experience and the staff's perception of the risk significance of the issues involved based on the experience to date from risk assessments.

The enclosure also identifies the Phase II topics that have been deleted for Phase III, and the correspond-ing basis.

Any comments you might be able to offer on the basis for deleting topics would also be useful.

The Phase III topic list would be further reduced for a Phase III plant on a l

plant-specific basis where the licensee can demonstrate that specific issues F

are of lesser safety significance based on plant-specific considerations.

L The staff has conservatively assuned that the review effort for these topics would require no more than $2 million per plant, based on reported costs for Phase 11 ranging from $2.5 to $3.5 million and the reduction in the nunber of topics. We would also appreciate your views on this estimate.

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Mr. D. Edwards.

In order to maintain the current schedule for the development of a proposal for Phase III, we would appreciate any comments you can offer by September 30, 1982.

Sincerely, a

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" Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation

Enclosure:

c As stated cc w/ enclosure:

R. Kacich 4

T. Tipton i

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l TOPIC 1.1, SETTLEMENT OF FOUNDATIONS AND BURIED EQUIPMENT I.

INTRODUCTION The staff objective of this review is to assure that safety related structures, systems and components are adequately protected against excessive settlement. The scope includes the review of subsurface materials (soils or geologici and foundations to assess the potential static and seismically induced settlement of all safety related structures and buried equipment.

II.

REVIEW CRITERIA The acceptance criteria for the settlement of foundations and buried equipment are stated in General Design Criterion 2 (GDC 2) of Appendix A to 10 CFR Part 50 and Appendix A to 10 CFR Part 100. GDC 2, " Design Bases for Protection Against Natural Phenomena," requires that structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes, without loss of capability to perform their safety function. Appendix A to 10 CFR Part 100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants,"

describes the nature of the investigations required to obtain the geologic and seismic data necessary to determine site characteristics and to identify.

geologic and seismic factors required to be taken into account in the design of nuclear power plants.

I III. REVIEW GUIDELINES The review process is conducted in accordance with the procedures described in Standard Review Plan Section 2.5.4, " Stability of Subsurface Materials and Foundations," and 2.5.1, " Basic Geologic and Seismologic Information."

Regulatory Guides 1.132, " Site Investigations for Foundaitons of Nuclear Power Plants," and 1.138, " Laboratory Investigation of Soil for Engineering Analysis and Design of Nuclear Power Plants," provide information, recom-mendations, and guidance and describe a basis acceptable to the staff that may be used to implement the requirements of the criteria described in Section II above.

IV.

RELATED TOPICS 1.2 Stability of Slopes 1.3 Dam Integrity

1. 4 Ground Motion 1.5.1 Site Hydrologic Characteristics and Capability to Withstand Flooding

2. 2 Severe Weather Effects on Structures V.

SAFETY SIGNIFICANCE Excessive settlement or collapse of foundations and buried equipment for safety related structures, systems and components under either static or seismic loading could result in failure of structures; interconnecting piping, control systems or cables; or other equipments (tanks, etc.).

The consequences of these failures have not been considered in most

2-plant safety analyses. Therefore, assurance that excessive settlement of foundations and buried equipment will not occur must be provided.

Experience during construction and SEP Phase 11 revie'ws have identified safety concerns with respect to this issue.

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l TOPIC 1.2, STABILITY OF SLOPES 1.

INTRODUCTION Overstressing a slope (engineered and natural) may cause sudden failure with rapid displacement or shear strain which may damage power plant facilities.

The objective of this review topic is to assure that safety related structures, systems and components are adequately protected against failure of these slopes.

The scope of this topic includes the review of the stability of all earth and rock slopes, both natural and man-made (cuts, fills, embankments, etc.), whose failure, under any of the conditions to which they could be exposed during the life of the plant, could adversely affect the safety of the. pl ant.

II.

REVIEW CRITERIA The acceptance criteria for stability of slopes are stated in General Design Criterion 2 (GDC 2) of Appendix A to 10 CFR Part 50 and Appendix A to 10 CFR Part 100.

GDC 2, " Design Bases for Protection Against Natural Phenomena," requires that structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornados, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety function. Appendix A to 10 CFR Part 100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants,"

describes the nature of the investigations required to obtain the geologic and seismic data necessary to determine site characteristics and identify geologic and seismic factors required to be taken into account in the design of nuclear power plants.

,III.

REVIEW GUIDELINES The review process is conducted in accordance with the procedures described in Standard Review Plan Section 2.5.5, " Stability of Slopes."

Regulatory Guides 1.132, " Site Investigations for Foundation of Nuclear Power Plants," and 1.138, " Laboratory Investigation of Soil for Engineering Analysis and Design of Nuclear Power Plants," provide l

information, recommendations, and guidance and describe a basis

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acceptable to the staff that may be used to implement the requirements j

of the criteria described in Section II above.

IV.

RELATED TOPICS l.1 Settlement ~tf Foundations and Buried Equipment 1.3 Dam Integrity 1.4 Ground Motion

1. 5.1 Site Hydrologic Characteristics and Capability to Withstand Fl ooding 2.2 Severe Weather Effects on Structures 4

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SAFETY SIGNIFICANCE The safety significance of this review topic is that it provides the basis for assuring that safety related structures, systems and components are adequately protected against failure of on-site earth and rock slopes (both natural and engineered) or that the slopes are sufficiently stable so that their failure need not be considered as a design basis.

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TOPIC 1.3, DAM INTEGRITY I.

INTRODUCTION Dam integrity is the ability of a dam to safely perform its intended functions.

These functions would normally include remaining stable under all conditions of reservoir operation, controlling seepage to prevent excessive upliftng wa,ter pressures or erosion of soil materials-and providing sufficient freeboard and outlet capacity to prevent over-topping. The objective of this topic is to assure that adequate mar-gins of safety are available under all loading conditions and un-controlled releases of retained water are prevented.

II.

REVIEW CRITERIA The acceptance criteria for dam integrity are stated in General Design Criterion 2 (GDC 2) of Appendix A to 10 CFR Part 50 and, Appendix A to 10 CFR Part 100.. GDC 2, " Design Bases for Protection Against Natural Phenomena," requires that structures, systems and components important to safety shall be designed to withstand effects such as earthquakes, tornados, hurricanes, floods, tsunami, and seiches without loss of capa-bility to perform their safety function. Appendix A to 10 CFR Part 100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants,"

describes the nature of the investigations required to obtain the geologic and seismic data necessary to determine site suitability and identify geologic and seismic factors required to be taken into account in the siting and design of nuclear power plants.

III. REVIEW GUIDELINES The review process is conducted in accordance with the Standard Review Plan Sections 2.5.4, " Suitability of Subsurface Materials and Founda-tions," and 2.5.5, " Stability of Slopes." Additional information and guidance are presented in Regulatory Guides 1.27, " Ultimate Heat Sink for Nuclear Power Plants," 1.132, " Site Investigations for Foundations of Nuclear Power Plants," and 1.138, " Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants."

IV.

RELATED TOPICS 1.1 Settlement of Foundations and Buried Equipment 1.2 Stability of Slopes 1.4 Ground Motion 1.5.1 Site Hydrologic Characteristics and Capability to Withstand Fl ooding 2.2 Severe Weather Effects on Structures V.

SAFETY SIGNIFICANCE The safety significance of this topic is to provide the basis for assuring that all safety related structures, systems and components are adequately protected against flooding and an appropriate supply of cooling water during normal and emergency shutdown procedures.

TOPIC 1.4, GROUND MOTION I.

INTRODUCTION The objective of Topic 1.4 is to assure that the free field ground motion specified for the plant design adequately represents the l

vibratory ground motion associated with a postulated safe shutdown earthquake (SSE) at the plant.

The free field ground motion will be represented as the design spectra for the facility that will be utilized as the input to analyses to verify the design adequacy of structures, piping and equipment.

II.

REVIEW CRITERIA The acceptance criteria for the design bases for protection against natural phenomena are contained i'n Appendix A to 10 CFR 50, General Design Criterion 2.

It states, in part, "The design basts for these structuress systems and components shall reflect (1) appropriate consideration of the most severe of the natural phenomena that have been hfstorically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity and period of time in which the historical data have been accumulated..." Further acceptance criteria are described in 10 CFR Part 100, Appendix A.

" Seismic and Geologic Siting Criteria for Ncclear Power Plants."

These criteria describe the nature of the investigations required to cbtain the geologic and seismic data necessary to determine site suitability and identifies geologic and seismic factors required to be taken into account in the siting and design of nuclear power pl ants.

III. REVIEW GUIDELINES j

The acceptance guidelines are stated in Appendix A to 10 CFR Part 100 as implemented thru:

1 A.

Standard Review Plan Section 2.5.2, " Vibratory Ground Motion",

and/or B.

" Seismic Hazard Analysis," NUREG/CR-1582, August 1980.

IV.

RELATED TOPICS 2.5 Seismic Design of Structures, Systems and Components V.

SAFETY SIGNIFICANCE In order to ensure that a facility could withstand a very large earthquake without the loss of capability of systems important to safety and the structures that house them, it is necessary to l'

2-verify that the free field ground motion has been properly defined.

The effects from postulated external phenomena on a nuclear plant are particularly significant in that the entire faciljty is af fected.

In SEP Phase II, the free field seisnic ground motion was found to be conservatively specified in some cases and not in

others, Results of seismic analyses using the free field ground motion motion defined as a part of SEP Phase II have identified the need for hardware modifications to some plant systems and structures.

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TOPIC 1.5.1, SITE HYDROLOGIC CHARACTERISTICS AND CAPABILITY TO WITHSTAND FLOODING 1.

INTRODUCTION Hydrologic considerations are the interface of the plant with the hydrosphere, the identification of hydrologic casual mechanisms that may require special plant design or operating limitations with regard to floods and water supply requirements.

The scope of this topic includes identifying the site hydrologic characteristics, the capability of structures important to safety to withstand flooding, the detennination of the adequacy of the cool-ing water supply and the inservice inspection of water control structures.

Specific issues reviewed in this topic are:

A.

Hydrologic Description - To assure that plant desig6 reflects appropriate hydrologic conditions.

B.

Flooding Potential and Protection - To assure that the plant is adequatley protccted against floods.

C.

Ultimate Heat Sink - To assure an appropriate supply of cooling water during nonnal and emergency shutdowns.

D.

Inservice Inspection of Water Control Structures - Assure adequate inspection program is in place to prevent water control structure deterioration or failure, which could result in flooding or loss of ultimate heat sink.

II.

REVIEW CRITERIA The acceptance criteria for this topic are stated in the General l

Design Criteria (GDC) of Appendix A to 10 CFR Part 50, in 10 CFR l

Part 100 and Appendix A to 10 CFR Part 100.

GDC 2, " Design Bases l

for Protection Against Natural Phenomena," requires that structures, systems and components important to safety be designed to withstand I

f the effects of natural phenomena, such as flooding.

10 CFR Part 100, " Reactor Site Criteria," as it relates to identifying and I

evaluating hydrologic features of a site. 10 CFR Part 100, Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants,"

l as it relates to establishing the design basis flood.

l III.

REVIEW GUIDELINES The current guideines used to detemine if plant design meets the I

topic acceptance criteria are those provided in Standard Review Plan Sections 2.4, the site hydrology review plans, 3.4.1, " Flood Protection," and 9.2.5, " Ultimate Heat Sink." Additional infonnation and guidance are provided in Regulatory Guides 1.27, " Ultimate Heat i

- Sink for Nuclear Power Plants," 1.59, " Design Basis Floods for Nuclear Power Plants," 1.102, " Flood Protection for Nuclear Power Plants," and 1.127, " Inspection of Water Control Structures Associated with Nuclear Power. Plants."

IV.

RELATED TOPICS 1.1 Settlement of Foundations and Buried-Equipment 1.2 Stability of Sicpes 1.3 Dam Integrity,

2. 2 Severe Weather Effects on Structures 1.6.3 Internally Generated Missiles 4.2.13. Shutdown Systems 1.5.2* Site Severe Weather Characteristics SdhETYSIGNIFICANCE V.

Plant designs incorporating appropriate site hydrologic characteristics and flooding protection significantly reduces the radiological release hazards resulting from flooding accidents.

In SEP Phase II several plants were found to not be adequately protected for external flooding.

Licensees h' ave implemented plant modifications to upgrade structures to withstand flood levels and hydrostatic loads.

Emergency procedures have been revised and ISI programs developed or modified for some.

facilities, upgrading to meet current licensing criteria (i.e., the Probable Maximum Flood) could not be practically accomplished.

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TOPIC 1.5.2, SITE SEVERE WEATHER CHARACTERISTICS e

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INTRODUCTION Safety-related structures, systems and components should be designed to _ function under all severe weather conditions to which they may be exposed.

Meteorological phepomena to be considered include straight x

wind loads, tornados, snow and ice loads, and other phenomena judged to be significant for a particular site.

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The objective of this topic is to identif thosel meteorological condi-tions which should be considered in the SEP structural reviews.

.s II.

REVIEW CRITERIA General Design Criterion 2, " Design l Bases for Protection Against Natural Phenomena," of 10 CFR Part 50, Appendix A requires that structures, systems and components important to safety shall be designed to withstand the effects of natural phenonena without loss of capability to perfom their safety functions. The design,bacis shall reflect consideration s

of the most severe of the phenomana' that has been historically reported 1

for the site and surrounding, area, appropriate combinations of the effects of normal and accident conditions with the natural phenomena and importance.of the safety functio to be performed.

III. REVIEW GUIDELINES

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The review is conducted in accordance with the criteria given in Standard Review Plan Section 2.3.1, " Regional Climatology."

IV.

RELATED TOPICS

2. 2 Severe Weather Effects on Structures

2. 3 Design Codes, Criteria and Load Combinations for Structures 1.6.2 Tornado Missiles V.

SAFETY SIGNIFICANCE The findings of this topic will be used as input to the related topics listed above and, thus, the safety significance of this topic will be determined in the structural evaluation of the plant. During SEP Phase II, site specific evaluations of wind speed versus recurrence interval, s

including confidence limits, were developed to assess the probacility of extreme wind speeds and loadings to assist making backfit decisions.

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TOPIC 1.6.1, INDUSTRIAL HAZARDS I'

INTRODUCTION The objective of this topic is to ensure that the integrity of the safety-related structures, systems and components would not be jeopardized due to the potential for hazards originating at nearby facilities.

Such hazards include shock waves from a nearby explosion, transport of explosive gases or chemicals resulting in fires or -

explosions, aircraft impact and missiles resulting from nearby explosions.

Examples of activities which could result in these hazards are industrial, transportation via nearby trucking routes or shipping lanes, local, federal and military aviation and military training activities.

II.

REVIEW CRITERIA General Design Criterion 4 " Environmental and Missile Design Basis,"

of Appendix A to 10 CFR Part 50, requires that nuclear power plant structures, systems and components important to safety be appropriately protected against events and conditions that may occur outside the nuclear power plant.

III. REVIEW GUIDELINES The review will be conducted.in accordance with the guidance given in Standard Review Plan Sections 2.2.1-2.2.2, " Identification of Potential Hazards in Site Vicinity;" 2.2.3, " Evaluation of Potential Accidents;" 3.5.1.5, " Site Proximity Mi ssiles (except Aircraft);" and 3.5.1.6, " Ai rcraf t Hazards."

IV.

RELATED TOPICS None.

V.

SAFETY SIGNIFICANCE Industrial hazards in the vicinity of the plant site could affect safety-related structures, systems or components necessary to achieve a safe shutdown condition or to mitigate the consequences-of a related accident.

The potential need for protection from nearby industrial hazards-which were not considered in the original design basis has been identified on some SEP Phase 11 plants.

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TOPIC 1.6.2, TORNADO MISSILES I.

INTRODUCTION Plants designed after 1972 have been consistently reviewed for adequate protection against tornados.

The concern exists, however, that plants reviewed prior to 1972 may not be adequately protected, in particular those reviewed before 1968 when AEC criteria on tornado protection were developed.

The scope of this topic is to assure that safety structures, systems and components can withstand the impact of an appropriate postulated spectrum of tornado generated missiles.

An assessment of the ability of a plant to withstand the impact of tornado missiles would include:

1.

Determination of the capability of the exposed systihs, components and structures to withstand key missiles (including small missiles with penetrating characteristics and larger missiles which result in an overall structural impact), and 2.

Determination of whether any areas of the plant require additional protection.

II.

REVIEW CRITERIA The criteria governing the design for tornado missiles are given in the General Design Criteria of Appendix A,10 CFR Part 50.

GDC 2

" Design Bases for Protection Against Natural Phenomena," requires that structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as tornados without loss of capability to perform their safety functions.

III. REVIEW GUIDELINES Acceptance guidelines are described in NRC Standard Review Plan 3.5.1.4,

" Missiles Generated by Natural Phenomena," and Regulatory Guide 1.117 Tornado Design Classification."

IV.

RELATED TOPICS

1. 5.1 Site Hydrologic Characteristics and Capability to Withstand Fl ooding 1.5.2 Site Severe Weather Characteristics V.

SAFETY SIGNIFICANCE The safety significance of this topic is to insure that structures, systems and components be adequately designed to resist the external loads.

In many instances in SEP Phase II, the potential flood, wind and tornado loads have been found to be greater.in magnitude than that considered in the original design, resulting in the need for plant modifications.

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TOPIC 2.3, DESIGN CODES, CRITERIA AND LOAD COMBINATIONS FOR STRUCTURES 1.

INTRODUCTION SEP plants were generally designed and constructed during a time span for which the design codes and criteria differ from those currently 4

used for evaluating new plarlts.

The objective of this topic is to provide assurance that plant Category I structures will withstand the appropriate design conditions without impairment of structural integrity or the perfonnance of required safety functions. The design codes, criteria and load combinations for all Category I structures (i.e., containment, structures inside containment and structures outside containment) are reviewed.

1 II.

REVIEW CRITERIA 10 CFR Part 50, Appendix A, General Design Criterion 1, " Quality Standards and Records," requires, in part, that structures important to safety be designed to quality standards commensurate with the importance of the safety function to be performed.

III.

REVIEW GUIDELINES The review guidelines are presented in Standard Review Plan Sections 3.8.1, " Concrete Containment," 3.8.2, " Steel Containment," 3.8.3,

" Concrete and Steel Internal Structures of Concrete or Steel Containments,"

3.8.4, "Other Seismic Category I Structures," and 3.8.5, " Foundations."

IV.

RELATED TOPICS 2.1 Classification of Structures

2. 2 Severe Weather Effects on Structures l

2. 5 Seismic Design of Structures, Systems and Components V.

SAFETY SIGNIFICANCE Code, load and loading combination requirements have evolved such that older plants may not have a sufficiently conservative design as compared to current requirements.

In Phase II of SEP, detailed comparisons of earlier versions to current American Concrete Institute and American Institute of Steel Construction and American Society of Mechanical Engineers ( ASME) standards identified a number of potentially safety significant changes which resulted in the need for some plant modifica-tions. This topic was coordinated with other structural topics (e.g.,

seismic, wind and tornado loading) during SEP Phase II where modifica-tions were involved.

TOPIC 2.4, CONTAINMENT DESIGN AND INSPECTION I.

INTRODUCTION The purpose of this topic is to review the inspection program of pre-stressed concrete containments. The program should include liftoff testing and acceptance criteria, testing of prestressing tendons and possible deterioration of pr~estressed containments. The likelihood of delamination occurring in the shell walls or dome is also reviewed.

II.

REVIEW CRITERIA 10 CFR 50, Appendix J and General Design Criterion 50 discuss contain-ment-design and inspection. GDC 50, " Containment Design Basis,"

requires that the containment: structure be designed to accommodate the calculated pressure and temperature conditions.

III. REVIEW GUIDELINES Regulatory Guides 1.35, " Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures," and 1.40, " Inservice Inspec-tion of Prestressed Concrete Containment Structures with Grouted Tendons,"

and Standard Review Plan Section 3.8.1, " Concrete Containment."

IV.

RELATED TOPICS 2.1 Classification of Structures

2. 3 Design Codes, Criteria and Load Combinations for-Structures V.

SAFETY SIGNIFICANCE Should the tendons of a prestressed concrete containment vessel ex-perience faster than normal relaxation or corrosion, the ability of the containment to withstand the design loads may be compromised.

The two plants for which this topic was applicable in SEP Phase II did not have an adequate inspection program. The inspection programs were upgraded to meet the acceptance criteria.

TOPIC 2.5, SEISMIC DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS I.

INTRODUCTION The objective of this topic is to review and evaluate the original seismic design (seismic input, analysis methods, design criteria, seismic instru-mentation, seismic classification) of the safety related plant structures, systems and components to enspre the capability of the plant to withstand the effect of earthquake. This review and evaluation will address 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 includes three major areas: the integrity of the reactor coolant pressure boundary; the integrity of fluid and electrical distribution systems related to safe shutdown; and the integrity of mechanical and electrical equipment and engineered safety features systems (including containment). A detailed review of all safety related str'uctures, systems and components will not be, conducted; rather a sampling approach supported by a se_t_of confirmatory analyses will be. performed. The sample size and confirmatory analyses will be increased, if necessary.

II.

REVIEW CRITERIA The acceptance criteria for the seismic design consideration are stated in the General Design Criteria of Appendix A,10 CFR Part 50. Criterion 2, " Design Bases for Protection Against Natural Phenomena," requires that structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes, without loss of capability to perform their safety function.

III. REVIEW GUIDELINES The following review criteria and guidelines to be used for review are:

1.

NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuc19er Power Plants," by N. M. Newmark and W. J. Hall, May 1978.

2.

"SEP Guidelines for Soil-Structures Interaction Review," by SEP Senior Seismic Review Team, December 8,1980.

For the cases that are not covered by the criteria stated above,' the following Standard Review Plans and Regulatory Guides will be used:

1.

Standard Revi w Plan, Sections 2. 5, 3. 7, 3. 8, 3. 9, and 3.10.

2.

Regulatory Guides 1.26,1.29,.1.60,1.61,1.92,1.100, and 1.122; l

1 IV.

RELATED TOPICS 1.4 Ground Motion 2.1 Classification of Structures

2. 3 Design Codes, Criteria and Load Combinations for Structures V.

SAFETY SIGNIFICANCE.

Many nuclear power plant facilities received construction permits in the 1960's.

Seismic design criteria and procedures evolved significantly.

during and after this period. As a result of the seismic review of Palisades, Ginna, Oyster Creek, Millstone 1, and Dresdep_2 during SEP Phase II, four common design deficiencies of plant facilities were identified:

(1) structural integrity of anchorage and support systems of all safety related electrical equipment, (2) structural integrity of vertical pumps, (3) field erected tanks, and (4) support systems of safety related piping.

It is expected that similar design deficiencies would be found in the Phase III review. By recognizing the evolution of design criteria and procedures, and the experience obtained from SEP Phase II seismic review, a reassessment of the seismic design basis is necessary to ensure the capability of the plant facilities to withstand the effects of earthquakes.

TOPIC 3.1, RCPB LEAKAGE DETECTION I.

INTRODUCTION The safety objective of Topic 3.1 is to determine the reliability and sensitivity of leakage detection systems which monitor the reactor

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coolant pressure boundary. The leakage detection systems should monitor reactor coolant pressure boundary leakage to the containment and to interconnecting systems.

II.

REVIEW CRITERIA l

The acceptance criteria for the detection of leakage from the reactor coolant pressure boundary are stated in the General Design Criteria of Appendix A,10 CFR Part 50.

Criterion 30, " Quality of Reactor Coolant Pressure Boundary," requires that means shall be prwided for detecting and, to the extent practical, identifying ~the location of the sources of leakage in the reactor coolant pressure boundary.

III. REVIEW GUIDELINES The acceptance criteria are described in Standard Review Plan Section 5.2.5, " Reactor Coolant Pressure Boundary Leakage Detection."

Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," provides guidance on the measures acceptable to the staff to implement the review criteria abwe.

IV.

RELATED TOPICS 7.1. 2 Pipe Break Effects on Systems and Components V.

SAFETY SIGNIFICANCE The safety significance of a reliable and sensitive leakage detection system is that it provides the reactor operator with an adequate margin of time to initiate procedures to identify the source of a leak, and isolate and repair it before the leak can grow to a size that might result in loss of component or system function or a loss of coolant accident.

It was the experiencs of the SEP Phase II review that no plant met all of the topic acceptance criteria such that system and/or procedural modifications were required.

TOPIC 3.2, REACTOR CORE ISOLATION COOLING. SYSTEM (BWR) 1.

INTRODUCTION The reactor core isolation cooling system (RCIC) prwides core cooling in the event of reactor isolation with loss of feedwater flow.

For small loss of coolant accidents, the RCIC may be a backup for the high pressure coolant injection system (HPCI). However, the RCJC was not originally designed as a safety system so credit for its operation in safety analyses should not be given unless the system.

classification is upgraded.

II.

REVIEW CRITERIA General Design Criterion 35 of Appendix A to 10 CFR Part 50, " Emergency Core Cooling," states that a. system to provide abundant emergency core-cooling should be provided. The system should have suitable redundancy to Jmre that with just onsite or just offsite power, the system safety function can be accomplished assuming a single failure.

III.

REVIEW GJIDELINES The acceptance criteria for emergency core cooling systems are described in Standard Review Plan Section 6.3, " Emergency Core Cooling System."

The criteria for the RCIC are described in SRP 5.4.6, " Reactor Core Isolation Cooling Systcm (BWR)."

IV.

RELATED TOPICS 4.2.1 Shutdown Systems V.

SAFETY SIGNIFICANCE Since the RCIC is relied upon for emergency core cooling and decay heat removal, the system design should be such that there is reasonable assurance that the system will perfom its safety function.

For Phase II, none of the plants had an RCIC, so this topic was no't applicable.

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TOPIC 4.1, CLASSIFICATION AND DESIGN OF SYSTEMS AND COMPONENTS I.

INTRODUCTION Systems and components important to safety should be designed, fabricated, erected and tested to quality standards commensurate with the safety function to be performed.

Also, systems and components that are required to withstand the effects of a safe shutdown earthquake and remain functional should be classified as Seismic Category I.

Due to the evolutionary nature of design codes and standards, operating plants may have been designed to requirements that are not as conser-vative as those currently required. The objective of this topic is to determine the classifications of systems and components applicable at the construction permit stage and at present.

II.

REVIEW CRITERIA The acceptance criteria for classification of systems and components are stated in the General Design Criteria (GDC) of Appendix A to 10 CFR Part 50. GDC 1, " Quality Standards and Records," requires that systems and components be designed to generally recognized codes and standards acceptable to the NRC. GDC ?. " Design Bases for Protection Against Natural Phenomena," requires that systems and components i

important to safety be designed to withstand the effects of earthquakes.

l III. REVIEW GUIDELINES The review guidelines for this topic are contained in Standard Review Plan Sections 3.2.1, " Seismic Classification," and 3.2.2, " System Quality Group Classification." Additional guidance is found in the recommendations of Regulatory Guides 1.26, " Quality Group Classifica-tion and Standards," and 1.29, " Seismic Design Classification."

IV.

RELATED TOPICS 2.1 Classification of Structures The other Section 4, " Plant Systems," topics will incorporate the i

findings of this topic into their review.

V.

SAFETY SIGNIFICANCE Due to the evolutionary nature of component and system codes, it is possible that some aspects of the original plant. design are significantly less conservative than that required by current codes. This topic will serve as input to the other olant systems topics and, therefore, the, safety significance of the findings in this topic will be evaluated elsewhere.

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TOPIC 4.2.1, SHUTDOWN SYSTEMS I.

INTRODUCTION The safety objective of Topic 4.2.1 is to ensure reliable plant shutdown capability using safety-grade equipment. Systems needed to remove decay heat and reach safe shut,down should have sufficient redundancy to assure that system function can be accomplished with a loss of offsite power and a single failure.

Systems needed to shutdown must also remain functional following external events. This topic identifies the necessary systems and reviews plant operating procedures.

II.

REVIEW CRITERIA The acceptance criteria for systems provided to remove decay heat are stated in the General Design. Criteria (GDC) of Appendix A,10 CFR Part 50.

GDC 34, " Residual Heat / Removal," requires that sysfWh function be accomplished with a single failure and just onsite or just offsite power available. GDC 2, "Desfhn Bases for Protection Against Natural Phenomena,"

and GDC 4, " Environmental and Missile Design Bases," require that systems and components important to safety be designed to withstand the effects of earthquakes, tornados, hurricanes, floods and missiles without loss of safety function. GDC 19, " Control Room," requires that equipment be located outside of the control room with a design capability for prompt hot shutdown and potential capability for cold shutdown through the use of suitable procedures.

III.

REVIEW GUIDELINES Current licensing guidelines for the review of decay heat removal capability are contained in Standard Review Plan Section 5.4.7, " Residual Heat Removal (RHR) System," Branch Technical Position (BTP) RSB 5-1,

" Design Requirements of the Residual Heat Removal System," and Regulatory Guide 1.139, " Guidance for Residual Heat Removal."

i IV.

RELATED TOPICS 4.2.2 Shutdown Electrical Instrumentation and Control 4.1 Classification and Design of Systems and Components 1.4 Ground Motion

2. 5 Seismic Design of Structures, Systems and Components 1.5.1 Site Hydrologic Characteristics and Capability to Withstand Flooding Topics 1.6.1 through 1.6.4 7.1. 2 Pipe Break Effects on Systems and Components 3.2 Reactor Core Isolation Cooling System (BWR) 4.3 Service and Cooling Water Systems

2 Note: The Appendix R and 10 CFR Section 50.48 reviews evaluate the capability to withstand fires and reach safe shutdown. This includes shutdown capability from outside the control room.

Y.

SAFETY SIGNIFICANCE A reliable method for decay heat removal is needed to ensure capability to safely shutdown the reactor.- The plant must have the capability to safely shutdown following external events. This topic identifies the required functions and. systems. This infomation is used-as input to the related topics.

It was the experience in Phase II that some of the systems nomally used for safe shutdown were not adequately protected for external events. Some systems were seismically qualified, but not protected for floods or wind and tornado effects. Thus, no single shutdown method or set of systems met all of the review criteria. The SEP review evaluated operation of alternative systems that could be used in the event of failures, however, the operating procedures did not address use of these systems.

Several SEP Phase II licensee's have commented that this review was particularly useful and worthwhile.

TOPIC 4.2.2, SHUTDOWN ELECTRICAL INSTRUMENTATION AND CONTROLS 1.

INTRODUCTION Review plant systems that are needed to achieve and maintain a safe shutdown condition of the plant, including the capability for prompt hot shutdown of the reactor from outside the control-room.

The t

l review also includes the capability and methods for bringing the plant from a high pressure condition to a low pressure cooling condition assuming the use of only safety equipment.

II.

REVIEW CRITERIA 10 CFR 50.55a(h) requires that new plants satisfy the requirements of IEEE Standard 279-1971.

III. REVIEW GUIDELINES The acceptance guidelines for the review of safe shutdown systems are presented in Standard Review Plan Sections 7.3, " Engineered Safety Features Systems," 7.4, " Safe Shutdown Systems," and 7.6,

" Interlock Systems Important to Safety."

IEEE Standard 279-1971 presents design criteria for assuring that systems required for safety will function even in the event of a single random failure.

i Required techniques include redundancy, independency and periodic testing.

Required design reviews include design basis events, response times, instrument accuracy and setpoints.

The plant design is reviewed to assure:

A.

The adequacy of the safe shutdown system to (i) initiate automatically the operation of appropriate systems, including the reactivity control systems, such that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences or postulated accidents and (ii) initiate the operation of systems and components required to bring the plant to a safe shutdown.

B.

That the required systems and equipment, including necessary instrumentation and controls to maintain the plant in a safe condition during hot shutdown are in an appropriate location outside the control room and have the capability for subsequent cold shutdown _of the reactor through the use of suitable pro-cedures.

C.

For PWRs, that safety grade equipment is available to bring the reactor coolant system fr'om a high pressure condition to a low pressure cooling condition.

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IV.

RELATED TOPICS 4.2.1 Shutdown Systems 5.2 RPS and ESF Testing V.

SAFETY SIGNIFICANCE A reliable method of = decay heat removal is necessary to ensure the capability to safely shutdown the reactor for all design basis conditions. This-topic identifies the instrumentation and controls necessary to accomplish this function and evaluates their quality with respect to I,EEE Standard 279-1971.

4 It was the experienc~e of the SEP Phase II review that most plants failed to satisfy the acceptance criteria and that system and technical specification modifications were required.

j TOPIC 4.3, SERVICE AND COOLING WATER SYSTEMS 1.

INTRODUCTION The objective of Topic 4.3 is to assure that the station service and cooling water systems have the capability, with adequate margin, to meet their design objective., To assure, in particular, that:

A.

Cooling water systems are capable of transferring heat from structures, systems and components important to safety to the ultimate heat sink.

B.

Systems are provided with adequate physical separation such that there are no adverse interactions among those systems under any mode of operation.

C.

Sufficient cooling water inventory has been provide ~'or that d

adequate provisions for makeup are available.

II.

REVIEW CRITERIA The acceptance criteria for the service and cooling water system are stated in the General Design Criteria (GDC) of Appendix A,10 CFR Part 50, GDC 44, " Cooling Water," GDC 45, " Inspection of Cooling Water System," and GDC 46, " Testing of Cooling Water System." These General Design Criteria require that a cooling water system be provided, inspected and tested and that the system be capable of transferring heat from structures, systems and components important to safety to the ultimate heat sink.

III. REVIEW GUIDELINES The acceptance criteria are described in Standard Review Plan Sections 9.2.1, " Station Service Water System," and 9.2.2, " Reactor Auxiliary Cooling Water Systems."

IV.

RELATED TOPICS 4.1 Classification and Design of Systems and Components 4.2.1 Shutdown Systems

4. 4 Ventilation Systems V.

SAFETY SIGNIFICANCE These systems are the mechanisms for removing decay heat from the core, under emergency conditions, to the ultimate heat sink. Loss of these systems could result in a core melt event.

In SEP Phase II, deviations from the acceptance criteria were found that resulted in modifications to plants and plant procedures.

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TOPIC 4.4, VENTILATION SYSTEMS 1.

INTRODUCTION To assure that the ventilation systems have the capability to provide a safe environment for plant personnel and for engineered safety 4

feature systems, it is necessary to review the design and operation of these systems.

For example, the function of the spent fuel p,ool area ventilation system is to provide ventilation in the spent fuel pool equipment areas, to permit personnel access, and to control air-borne radioactivity in the area during normal operation, anticipated operational transients, and following postulated fuel handling accidents. The function of the engineered safety feature ventilation systems is to provide-a suitable ~and-controlled environment for engineered. safety feature components following certain anticipated j

A transients ~ and design basis accidents.

4 II.

REVIEW CRITERIA The acceptance criteria for ventilation systems are stated in the General Design Criteria (GDC) of Appendix A,10 CFR Part 50. GDC

4. " Environmental and Missile Design Bases," requires that systems and components important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation and postulated accidents. Ventila -

tion systems maintain those conditions.

III. REVIEW GUIDELINES The guidelines used to determine if the plant systems meet the topic safety objective are those provided in Standard Review Plan Sections 9.4.1, " Control Room Area Ventilation Systems," 9.4.2,

" Spent Fuel Pool Area Ventilation Systems," 9.4.3, " Auxiliary and Radwaste Area Ventilation System," 9.4.4, " Turbine Area Ventilation System" and 9.4.5, " Engineered Safety Feature Ventilation System."

IV.

RELATED TOPICS

5. 2 RPS and ESF Testing.

V.

SAFETY SIGNIFICANCE The failure of systems used~ to change-from injection to recirculation or failure of the operator to complete the switchover in a timely _

l manner can result in the. loss of emergency-core -cooling and/or residual heat. removal capability.

In SEP Phase II, 50% of the plants failed to satisfy"the acceptance criteria and system and procedural modifications were required.

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TOPIC 4.7.2, RECIRCULATION LOOP ISOLATION (NON-JET PUMP BWR)

I.

INTRODUCTION All BWR facilities which have completed the Low Pressure Coolant Injection (LPCI) system modification (removal of the LPCI loop selection logic) are now required to have the rec,irculation pump discharge valves and bypass valves close upon initiation of LPCI. The closure of these valves is necessary for non-jet pump BWRs to prevent the loss of coolant water by reverse flow through the pump or its bypass line and out the pipe break.

The failure of the recirculation pump discharge or bypass valve to close can adversely affect core cooling in a manner similar to the failure of a LPCI valve to open, t

l This topic has been limited to non-jet pump BWRs because risk assessments have shown that a LOCA in a recirculation line in a jet pump BWR is not a

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significant contributor to risk.

II.

REVIEW CRITERIA The acceptance criteria for this topic are contained in General Design Criterion 35, " Emergency Core Cooling," of 10 CFR Part 50.

I III.

REVIEW GUIDELINES The review is conducted using the recommendations of NUREG-0123, " Standard Technical Specifications for General Electric Boiling Water Reactors," for emergency core cooling system actuation instrumentation.

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IV.

RELATED TOPICS

5. 2 RPS and ESF Testing.

V.

SAFETY SIGNIFICANCE The lack of isolation devices in the event of non-safety system failures could result in the propagation of faults to safety systems and common cause failures may result. During SEP Phae II, it was detemined that some RPS systems had no isolation from non-safety systems (e.g., plant computer) that multiple channels of the RPS were involved and that system modifications were necessary.

TOPIC 5.2, RPS AND ESF TESTING 1.

INTRODUCTION The plant design must assure that all Emergency Core Cooling System (ECCS) components, including the purnps and valves, are included in the component and system test, that the frequency and scope of the periodic testing is identified, and that the test program will provide adequate assurance that the system will function when needed.

II.

REVIEW CRITERIA 10 CFR 50.55a(h) requires that new plants satisfy the requirements of IEEE Standard 279-1971. General Design Criteria (GDC) 21, " Protection System Reliability and Testability," and GDC 37, " Testing of Emergency Core Cooling System," also apply as they relate to perto_dic testing requirements of the RPS and ESF.

III. REVIEW GUIDELINES The acceptance guidelines for the review of safety systems is presented in Standard Review Plan Section 7.2, " Reactor Trip System," and Section 7.3, " Engineered Safety Features Systems." IEEE Standard 279-1971 presents design criteria for assuring that systems required for safety will function even in the event of a single random failure. Required techniques include redundancy, independency, periodic testing, and instrument accuracy and setpoints.

The plant design information presented in the Safety Analysis Report, technical specifications and drawings are reviewed to assure that all ECCS components (e.g., valves and pumps) are included in the component and system test, that the frequency and scope of the periodic testing are adequate and meet the requirements of GDC 37.

Also, the review should verify that the operability of the RPS and ESF (on a periodic basis), and that the test program demonstrates a high degree of avail-ability of the systems.

IV.

RELATED TOPICS 4.2.2 Shutdown Electrical Instrumentation and Controls 4.6 Isolation of High and Low Pressure Systems 4.7.1 Automatic ECCS Switchover 4.8.1 Emergency AC Power Sysems 4.8.2 Emergency DC Power Systems 4.8.3 Swing Bus Design (BWR-4) 4,9 Shared Systems 5.'

Reactor Protection System and Engineered Safety Feature Systems Isolation

V.

SAEFTY SIGNIFICANCE The lack of an adequate test program may invalidate the design of the safety systems because only random failures are considered in single failure analyses.

If failures exist that cannot be detected by test the system function may be defeated. A periodic test program which identifies. pre-existing. failures.will significantly improve system availability.

In SEP Phase-II, several plants ~ failed to satisfy the -IEEE 279-1971 testing criteria-resulting in the need for system and procedure modifications. Response time testing was not within the, scope of' the review. Also, it was determined that instrument functional testing and calibration, when performed in accordance with the IEEE 279-1971 guidelines, provided assurance of system operability.

Al so,

important mechanical systems and components were subjected to response time testing (i.e., diesel generators, critical ESF pumps and valves, and control rod drive systems). Therefore, because mechanical system response times are much greater than those of instrumentation, the additional response time testing of instrumentation would not signifi-cantly improve system reliability.

s

TOPIC 6.1, ORGANIC MATERIALS 1.

INTRODUCTION The design basis for the selection of paints and other organic materials is not documented for most operating reactors. The plant design must assure that organic materials, such as organic paints, coatings and insulation materials, used inside containment do not adversely affect the operation of the engineered safety feature equipment inside contain-ment during accidents when they may be exposed to high temperatures, steam environments, high radiation fields, and containment spray systems.

The scope of this review will include an evaluation of qualification tests and licensee inspection and repair programs to assure that the i

organic materials will maintain their integrity and remain in a serviceable condition after exposure to the extreme entironmental conditions of a design basis accident.

II.

REVIEW CRITERIA The plant design will be reviewed with regard to General Design Criterion 1, " Quality Standards and Records," of Appendix A to 10 CFR Part 50, " General Design Criteria for Nuclear Power Plants,"

l which requires that structures and systems important to safety be designed and tested to quality standards commensurate with the importance of the safety function to be perfonned. Also, Appendix B to 10 CFR 50, " Quality Assurance Criteria for Nuclear Power Plants and fuel Reprocessing Plants," describes an acceptable method of -

complying with the Commissions quality assurance requirements with regard to protective coatings.

III. REVIEW GUIDELINES Current guidance for the review of organic materials in containment is provided in Standard Review Plan Sections 6.1.1, " Engineered Safety Features Materials," and 6.1.2, " Protective Coating Systems (Paints) - Organic Materials," and in Regulatory Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants."

IV.

RELATED TOPICS None.

V.

SAFETY SIGNIFICANCE The containment isolation system provides the primary barrier to the release of radioactivity to the environment. The isolation provisions must be capable of performing this function for a spectrum of accident and transient events with a single failure.

In SEP Phase II, signif-icant deviations from the acceptance criteria specified in NRC's regulations resulted in the need for both procedural and hardware modifications.

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TOPIC 7.3.1, RCS SPECIFIC ACTIVITY LIMITS I.

INTRODUCTION The coolant activity levels have a proportionate effect on those accidents involving primary coolant release (without core damage).

Implementation of Standard Technical Specifications (STS) limits are usually adequate to alleviate the concerns regarding resultant offsite doses.

The scope of the topic will be to examine the plant technical specifications to determine if they comply with the appropriate STS.

If not, an evaluation will be performed to detemine the adequacy of the existing plant technical specification limits in restricting offsite dose. The review will cover those accidents whose primary dose contribution is from reactor coolant leakage to the atmosphere (i.e., main steam line break outside W contain-ment, steam generator tube rupture and small line breaks outside containment).

A detemination of the plant specific atmospheric dispersion factor will be required to evaluate the offsite dose consequences.

II.

REVIEW CRITERIA Section 50.36, " Technical Specifications," of 10 CFR Part 50 requires that each license authorizing operation of a nuclear power reactor include technical specifications derived from the analysis.and evaluation included in the safety analysis report. The technical

(

specifications cre to include limiting conditions for operation (LCO) i and surveillance requirements. LCO's provide the lowest perfomance level required for safe operation of the plant.

III. REVIEW GUIDELINES The initial review will compare the existing plant technical specifica-tions with those recommended in the STS associated with the plant design.

The evaluation of offsite dose consequences will be done in accordance with the guidance provided in Standard Review Plans (SRP) 15.1.5, Appendix A, " Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR," 15.6.2, " Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment,"

15.6.3, " Radiological Consequences of Steam Generator Tube Failure (PWR),"

and 15.6.4, " Radiological Consequences of Main Steam Line Failure Outside Containment."

Atmospheric dispersion factors will be calculated in accordance with SRP 2.3.4, "Short-Term Diffusion Estimates for Accidental Atmospheric Rel ease. "

I l

IV.

RELATED TOPICS None.

V.

SAFETY SIGNIFICANCE The primary coolant activity is a measure of the core integrity during nomal plant operation. Excessireactivity would suggest an accelerated degradation of the fuel and could leak to offsite doses in excess of that which would reasonably be expected for a "non-accident" event.

In SEP Phase-II, several plants were found to have-high primary cocTant activity limits.

9 4

TOPIC 7.3.2, MSIV LEAKAGE (BWR) 1.

INTRODUCTION Operating experience has indicated that there is a relatively high failure rate and variety of failure modes for the main steam isolation valves in operating BWRs.

The objective is to assure that MSIV leakage rate limits are not routinely exceeded and the resulting calculated offsite doses do not exceed 10 CFR Part 100 guidelines.

II.

REVIEW CRITERIA The acceptance criteria for the main steam isolation valve leakage review is stated in the General Design Criteria of Appendix A,10 CFR Part 50.

GDC 54, " Piping Systems Penetrating Contalnment,"

requires leak detection capabilities for piping penetrating con-tainment.

III. REVIEW GUIDELINES The review for this topic is conducted in accordance with Standard Review Plan Sections 15.6.5, " Radiological Consequences of a Design Basis Loss of Coolant Accident Including Containment Leakage Contribution," and 6.7, " Main Steam Isolation Valve Leakage Control System (BWR)." Additional guidelines are presented in Regulatory Guide 1.96, " Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants."

IV.

RELATED TOPICS None.

Y.

SAFETY SIGNIFICANCE f

Operating experience has shown that about 30% of the non-service MSIV leakage tests are ur. successful in demonstrating compliance with plant license conditions. Many of these failures to meet technical specifi-cation leakage limits are by orders of magnitude, and some by such a large factor that the design basis pressure could not be attained in order to perform the measurement.

Such large leakages can greatly increase ~ the consequences of severe accidents, becoming the dominant pathway of releases to the environmen.

In addition to increasing the consequences of low probability events, excessive MSIV leakage can also increase the probability that accidents within the design basis will propagate into higher consequence sequen-Some measures MSIV leakages were so high that, had these plants ces.

suffered a steam line break accident, leakage would have been great enough to constitute a small LOCA, with a simultaneous loss of con-tainment due to the leak.

Since the MSIV leaks are a direct pathway to uncontrolled areas and then possibly to the environment, it can be a significant contributor to offsite releases during accident conditions.

~

f RISK IMPORTANCE OF SELECTED SEP III TOPICS To provide some insight into the' risk significance of the proposed SEP Phase III i

topics, the Reliability and Risk Assessment Branch evaluated selected topics.

The issues which were evaluated were" those where current probabilistic risk l

assessment methodology is mature enough to model the various elements which com-prise the issue.

Generally, external events have not been modeled in PRA studies, and were excluded from this review.

Due to the very limited time available it was not possible to perform a detailed Rather, a panel of risk assessment of each of the topics being considered.

experts was established which was knowledgeable of previous PRA studies, as well as commercial power plant design and operation. The input provided from this panel was judgmental in nature to provide results in the necessary time frame.

j Because of this approximate effort and because of the perceived variability in plant systems design, the panel's analysis may be biased in a conservative direction.

In evaluating each topic the panel considered whether that topic related to issues Specifically, each which previous risk studies had shown to be of significance.

SEP topic was first examined to determine if it was encompassed by present state-of-the-art risk assessment techniques.

If so, a determination was made on those Based plant systems affected by compliance with the SEP topic differences.

then on results of past risk assessments, it was determined what the importance of these systems were in a risk sense.

That is, in previous risk studies were failures of these systems shown to be of high or medium importance as leading causes to core melt? Finally, the impact of compliance with the SEP topic dif-ference upon these system's unavailability was crudely estimated.

In some cases, In other compliance with the SEP t9pic could greatly improve system reliability.

instances, compliance with the licensing criteria had little actual improvement on system reliability.

In the last step all these elements; system impacted, importance of system, issue impact on system, were considered to arrive at a panel consensus as to the general l

risk contribution potential for each of the reviewed SEP topics. These results j

are discussed individually below, and are summarized in Table 1.

It should be noted that only a limited number of detailed probabilistic risk studies have been concluded to date.

Thus, a number of topics were identified as of risk signifi-cance even though they have not been identified as such in previous risk. studies.

This occurred because, in the opinion of the review panel, the topic had the potential of being of risk significance, depending upon the design and operational details of the plant being reviewed.

I Since a large. number of SEP III topics have been identified as having pot'ential risk significance, it is believed that a plant specific risk study (performed for each Phase III plant) should provide important insight as to what issues are significant Addi-for a spe'cific plant, and to suggest the most effective backfit program.

tionally, these plant specific assessments may well identify many issues which are of no risk significance, and thereby preclude corrective actions on unnecessary topics.

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2-3.1 RCPB Leakage Detection l

The presence of an adequate leakage detection system can serve to reduce the frequency of loss of coolant accidents.

In a number of PRA studies to data, LOCA's have shown up as measurable contributers to risk, the actual risk reduction patential from an adequate leakage detection system-is very plant specific.

Because LOCA frequency. reduction would.have some effect"on risk for all plants, we assess this issue,as medium importance.

3.2 Reactor Core Isolation Cooling System (BWR)

Previous risk assessments of BWRs with RCIC systems (e.g., Browns Ferry IREP) 1 have shown that accident sequences involving high RCS pressure are dominant contributers to risk. RCIC is one of only two systems capable of' injecting water into the core at high pressure (the other is HPCI), although it is possible to blow down and use-low pressure systems for' injection. However, station blackout has been shown to be very important to risk, thus systems capable of operating without AC power prove to be very important.

RCIC is one such system, therefore we assess the importance of this issue to be high.

4.2.1 Shutdown Systems & 4.2.2 Shutdown Electrical Instrumentation & Controls Risk Assessments such as Browns Ferry IREP have shown that failure to maintain decay heat removal capability is a dominate contributer to risk. Due to the high importance of these systems, a comprehensive review of their capability and available means of alternate shutdown approaches has the potential for significant risk reduction. Therefore, we conclude the imprortance of these issues as high.

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, 4.3 Service and Cooling Water Systems and 4.4 Ventilation Systems.

Various risk studies have shown that the realiability 'of certain support systems is of high importance since they of ten maintain the operability of numerous front lina safety systems. Thus, the failure of one of these support systems, such as service water, could result in the loss of all auxiliary feedwater and emergency AC power in addition to main feedwater.

Due to the high consequences of failures in these support systems and potential for common cause failures of multiple safety systems, we rank these issues as high.

C 4.6 Isolation of High and Low Pressure Systems Tha failure to provide adequate isolation between high and low pressure piping could result in LOCAs outside the containment, which inevitably lead to core melt. The importance of this is very plant specific based on the plant design and testing procedures.

In some previous studies (e.g., Reactor Safety Study - Surry) these i

events are dominant risk contributors.

It is important, then, to evaluate this scenario since it has been shown to have high risk contribution in some plants.

We therefore rank the issue as high.

.4.7.1 Automatic ECCS Switchover Previous risk assessments have shown that operator error in ECCS switchover to l

recirculation to be dominant contributors to risk (e.g. - Crystal River IREP).

This has been due to such things as poor operating procudures or poor system

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design. Since the automatic switchover may affect these dominant contributors to l

a large degree, we assess the imporance of this issue to be high.

4.7.2 Recirculation Loop Isolation (Non-jet Pump BWRs)

In The purpose of this system is to isolate a large LOCA in a recirculation line.

previous risk studies perfonned, large LOCAs in this line have not been dominate l

contributors to risk. The reason for this is because a break in this line will only render one low pressure injection subsystem incapable of providing core cooling.

l

4

'HowevIr, there are multiple backup subsystems capable of providing this function.

The probability of additionally failing these backup systems has always been found low in prsvicus risk studies.

Therefore, we believe the impact from resolution of this issue to be small and conclude the importance of this issue to be low.

However, for non-jet pump plants, this issue is judged to be high due to potential uncovery of the core due to isolation of all loops i.e., Oyster Creek triple low level event of May 1979.

4 '. 8.1 Emergency AC Power Systems Lcss of offiste power transients have contributed significantly to almost every risk assessment previously perfonned.

In one such study (Millstone-1 IREP) it centributed over 85% of the core melt risk.

The emergency AC power system is always sh:wn to be the key system in mitigating these loss of offsite transients. This being the case, we assess the importance of this issue to be high.

4.8.2 Emergency DC Power Systems Loss of offsite power transients have contributed significantly to almost every risk assessment study previously perfonned.

Regardless'of the availability of emergency AC power, no system can operate in the absence of DC power. Failures of emergency DC power have been shown in these risk studies to contribute significantly to comon cause failures of multiple systems.

Therefore, we conclude this issue to bn of high importance.

4.8.3 Swing Bus Design (BWR4)

Propagation of electrical faults can contribute to the frequency of station blackout.

Station blackout has shown to be a significant contributor to plant'ris'k in a numbar of studies, e.g., Millstone I IREP.

However, circuit protection devices associated with the swing buses result in a relatively low probability of propagation of bus faults as comprared to the probability of station blackout from other sources.

Since the potential for some risk reduction exists, we have assessed the importance of this issue to be medium.

.S.

4.8.4 Electrical Penetration Design Pravious risk assessments have shown that containment leakage as the primary con-teinment failure mode following core melt,is not significant to total risk.

This is

b2cause there are more serious containment failure modes with much higher prob-abilities.

Additionally, evaluations of this particular mode of containment leakage (penstration failure due to overcurrent) has been shown not to contribute significantly to the total containment leakage probability (0yster Creek SEP Integrated Assessment).

~~

Therefore, we conicude the importance of this issue to be low.

4.9 Shared Systems Failures in shared systems could result in multiple failures which would be expected to have significant impact on safety. The limited reviews of SEP Phase II issues at Dresd:n 2 indicates that this topic importance is high.

l 5.1 Reactor Protection System and Engineered Safety Feature System Isolation In previous risk assessments it has been shown that the failure to generate initiating signals has not contributed to risk.

Additionally, in evaluation of 1 systems where isolation was not present for certain signals (Oyster Creek SEP) it was shown that due to the presence'of backup signals the lack of isolation did not contribute to the failure rate of initiation systems.

The reason for this is that multiple isolation failures would be required to render the initiation system totally inoperable.

However, this issue would be of high importance if a common fault could disable the protection system function.

Therefore, we believe some review of this topic is prudent, and we rate the issue conservatively as high.

l

5.2 RPS' and ESF Testing Failure rates of components are related to the time between tests, since some failures are only detectable during testing.

Some studies have shown that certain

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plant components are sometimes not tested for long periods of time, or due to inadequate procedures are'not tested at all.

The higher failure rates associated with these components have been shown in. some PRAs- (Millstone IREP) to have some m:asurable. impact on dominate risk sequences.

Thus, improved test procedures for these components would provide some risk reduction. Th'erefore ye' coh,clude this issue to be of medium importance.

g 7.1. 2 Pipe Break Effects on Systems and Components There has never been a detailed assessment of the pipe break effects on systems and components since PRA's studies, to date have not modeled the potential locations.

where breaks could occur and.the. resulting impact.on adjacent components.

However, due to the potential for multiple system failures due to high energy pipe breaks, we believe this issue should be reviewed on each plant and conservatively rank the issue as high.

7.2 Containment Isolation Previous risk assessments have shown that containment leakage as the primary con-tainment failure mode following core melt is not significant. to total risk. This is b:cause there are more serious containment ~ failures modes with much higher prob-abilities. However, these -conclusions are based on the presence of reasonably reliable containment isolation systems for. each plant studied.

If a plant was dis-covered with a very unreliable isolation system, containment leakage could con-ceivably show up as an important risk contributor. Therefore, we conservatively rate this issue as medium.

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7.3.2 MSIV Leakage (BWR)

The possiblity of MSIV leakage contributing to the probability of LOCA outside csntainment may have some impact on tot ~al risk.

Since a LOCA outside containment will result in core melt, this issue has potential significance from that perspective.

Howaver, studies performed to date have indicated that the dominate cause of LOCAS outside containment is from interfacing systems LOCA, see 4.6.

Therefore, since it is a serious failure mode, but not the dominate failure mode.,,we assess this to be of medium importance.-

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6 TABLE I SEP III TOPIC Title Risk Contribution 3.1 RCPB Leakage Detection" M

3.2 Reactor Core Isolation Cooling H

System (BWR) 4.2.1 Shutdown Systems H

4.2.2 Shutdown Electrical Instrumentation H

and Controls 4.3 Service and Cooling Water Systems H

4.4 Ventilation Systems H

4.6 Isolation of High and Low Pressure H

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Systems 4.7.1 Automatic ECCS Switchover H

4.7.2 Recirculation Loop Isolation (BWR)

H for non-jet pump plants only 4.8.1 Emergency AC Power System H

4.8.2 Emergency DC Power System H

4.8.3 Swing Bus Design,(BWR4)

M 4.8.4 Electrical Penetration Design L

4.9 Shared Systems H

5.1 Reactor Protection System and H

Engineered Safety Features Isolation 5.2 RPS and ESF Testing M

7.1. 2 Pipe Break Effects on Systems and H

Components 7.2 Containment Isolation M

7.3.2 MSIV Leakage (BWR)

M i

l l

1 t

l s

e TOPICS DELETED FOR PHASE III The topics described in the following sections, which evolved from SEP Phase I, would be excluded from SEP Phase III for the reasons identified.

Generic Activities I

The topics listed in-Table 1 were' deleted from SEP Phase II, and would similarly be deleted from SEP Phase III, because they either have been_or will be addressed as part of the implementation of the Three Mile Island (TMI) Action Plan or the Unresolved Safety Issues (USIs).

The results of these generic programs will be considered in the Integrated Assessments for Phase III when the imple-mentation requirements are available on a schedule consistent with the SEP review.

Topic V-4 was-deleted early in Phase II because of the generic activities associated with TAP A-26.

During the course of the generic review we found that a limited number of BWRs had furnace-sensitized stainless steel safe-ends.

Of those, only three plants (Nine Mile Point, Dresden 2, and Big Rock' Point),

did not fully comply with the requirements of NUREG-0313, Revision 1.

Two of these plants were in Phase II.

Nine Mile Point recently discovered extensive cracking of the safe-en'ds and is replacing the safe-ends and attached loop piping. This aspect of Topic V-4 hras been reopened for SEP II to examine the furnace-sensitized stainless steel safe-ends at Dresden and Big Rock Point.

Other aspects of this topic are being completed under USI A-42.

Phase II Experience During the Phase II SEP reviews, the current regulatory requirements associated with the topics listed in Tahle 2 were satisfied by all of the applicable plants or were essentially equivalent to current licensing criteria without the need for any plant modifications, such that there was not a safety issue involved.

Based on this experience, we expect that the newer plants in Phase III and begnd would similarly meet the current regulatory requirements.

SEP III TOPICS 4

Table 1.

Topics Deleted Based on Generic Issues SEP TMI or~U51 TNI, U51, or SEP title Topic No.

SEP title Instrumentation for Monitoring Accident Conditions 11-2.B Onsite Meteorological Measurements TMI II.F.3 TMI 111.4.1 Improve Licensee Emergency Preparedness - Short Te Program Instrumentation for Monitoring Accident Conditions Il 2.D Availabilit'y of Meteorological TMI II.F.3 Data in the Control Room TMI III.A.1 Improve Licensee Emergercy Preparedness - She'rt Te TMI 1.D.1 Control Room Design Reviews

!!! 8.0 Core Supports and Fuel Integrity USI A-2 Asymetric Blowdown Loads on Reactor Primary Coolant System Fracture Toughness of Steam Generator and Reactor 111-9 Support Integrity USI A-22 Coolant Pump Supports Mark 1 Containment Long Term Program USI A-7 Environmental Qualification of Safety-Related Equi.

USI A-24 Seismic Qualification of Equipment in Operating Pi USI A 46 Seismic Design Considerations Compliance With Codes and Standards (10 CFR Part 5 Section 50.55a)

Seismic Qualification of Equipment in Operating P'

!!!-11 C~omponent Integrity U51 A-46 USI A 2 Asymetric B1Adown Loads on Reactor Primary Coola Seismic Design Considerations

!!!-12 Environmental Qualification of USI A-24 Qualification of Safety-Related Equipment Safety-Related Equipment Reactor vessel Pressure Transient Protection' V-3 Overpressurization Protection USI A-26 V-4 Piping and Safe-End Integrity USI A 42 Pipe Cracks in Boiling Water Reactors Westinghouse, Combustion Engineering, and V8 Steam Generator Integrity USI A-3.

Babcock and Wilcox Steam Generator T6be Integrit)

A-4, A-5 USI A-1 Waterhammer V-13 Waterhammer VI-2.A Pressure Suppression-Type E=R USI A-7 Mark 1 Containment Long-Term Program Containments Asymetric Blowdown Loads on Reactor Primary Cool.

VI-2.B Subcompartment Analysis U51 A-2 System VI-5 Combustible Gas Control TMI II.B.7 Analysis of Hydrogen Control USI A 48 Hydrogen Contrcl Measures and Effects of Hydroge-Burns on Safety Equipment VI-7.F Emergency Core. Cooling System Sump USI A 43 Containment Emergency Sump Reliability Design and Test for Recirculation Mode Effectiveness VI-8 Control Room Habitability THI III.D.3.4 Control Room Habitability Requiremeats Vll-4 Effects of Failure in Nonsafety-USI A-47 Safety Implications of Control System Related Systems on Selected USI A-17 Systems Interactions in Nuclear Power Plants Engineered Safety Features Additional Accident Mcnitoring Instrumentation VII-5 Instruments for Monitoring Radia-TMI II.F.1 Identification of and Recovery From Conditions tion and Process Variables During TMI II.F.2 Leading to Inadequate Core Cooling Accidents Instruments for Monitoring Accident Conditions TMI II.F.3 Control of Heavy Loads Near. Spent Fuel Pool IX 2 Overhead Handling Systems (Cranes) USI A-36 X

Auxiliary Feeo.ater System TMI II.E.1.1 Auxiliary Feed.ater System Evaluation Procedures for Verification of Correct Performa-XIll-1 Conduct of Operations TMI I.C.6 of Operating Activities THI I!!.A.1 Jeprove Licensee Emergency Preparee ess - Shcrt-TMI !!!.A.2 Improving Licersee Emergency Prepa caness - Lor Control of Heavy Loads Near Spent Fuel Pool USI A-36 XV-21 Spent Fuel Cask Drop Accident Ar.ticipated Transients Without Scram Anticipated Transients Without Scram USI A-9 xv-22 bestinghouse, Coebustion Engineerf. ;, and Babcc XV-23 Multiple Tube Failures in Steam USI A-3, A-4, A-5 ars Wilcon Sten Generator Tube Irtegrity Generators A-ticipated Transients Witt' cut Scra-USI A-9 USI A-44 Station Blackout of All AC Power

I i

w Therefore, we have recommended that these topics be-deleted for future SEP reviews.

Included in this group are those topics which are reviewed periodically under separate ongoing NRC reviews (e.g., QA) and those topics that only served as inpute to other topics, such that a separate SEP evaluation was not necessary.

In addition, we have judged that the topics identified in Table 2 will either be considered by the staff' explicitly within the defined scope of the NREP analyses or will be considered in a limited fashion through a coupling of the NREP analyses results with a. standardized consequence analysis.

The topics listed in Table 3 would be excluded principally on.the basis of simi-lar Phase II experience, however, these topics will not be considered in th'e NREP analyses.

The staff has concluded that conformance with current licensing criteria or equivalence for all of the Phase II plants is a sufficient basis t'o exclude these topics for furture SEP reviews.

Miscellaneous Deletion of the following topics has been recommended for a variety of reasons, as detailed below for each topic.

Some of these topics will be covered by the NREP evaluations; however, the principal basis for the. staff's recommendation to exclude these topics is as noted.

III-8.A Loose Parts Mo'nitoring and Core Barrel Vibration Monitoring The staff is currently considering implementation of Regulatory Guide 1.133 on a generic basis.

The implementation plan for operating reactors, if any, will resolve this issue independently of SEP.

III-8.B Control Rod Drive Mechanism Integrity The staff evaluated this issqe on a generic basis and reported the results of that review in NUREG-0479, " Report on BWR Control Rod Drive Failures." This report fulfills the intent of the safety objective associated with this topic.

SEP III TOPICS 6

Table 2 Topics Excluded Based on Phase II Experience (Addressed by NREP)

Topic No.

SEP Title IV-1.A Operation with Less Than All Loops In-Service IV-2 Reactivity Control Systems Including Functional Design and Protection Against Single Failures IV-3 BWR Jet Pumps Operating In,dications V-7 Reactor Coolant Pump Overspeed VI-7.A.1 ECCS Reevaluation to Account for Increased Vessel Head Temperature 4 j

VI-7.A.2 Upper Plenum Injection VI-7.A.4 Core Spray Nozzle Effectiveness VI-7.C ECCS Single Failure Criterion and Requirements for Locking Out Power to Valves Including Independence of Interlocks on ECCS Valves VI-7.C.2 Failure Mode Analysis - ECCS VI-7.C.3 Effect of PWR Loop Isolation Valve Closure During a LOCA on ECCS Performance VI-7.D Long Term Cooling - Passive Failures (e.g., Flooding of Redundant Components)

VI-7.F Accumulator Isolation Valves-Power and Control System Design VII-1.B Trip Uncertainty and Setpoint Analysis Review of Operating Data Base' IX-4 Boron Addition System XV-4 Loss of Non-Emergency 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-8 Control Rod Misoperation.(System Malfunction or Operator Error)

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 the Boron XV-11 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position XV-12 Spe:trum of Rod Ejection Accidents (PWR)

XV-13 Spectrum of Rod Drop Accidents (BWR)

XV-14 Inadvertent Operation of ECCS 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-19 Loss-of-Coolant Accidents Resulting,from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary XVII Operational QA Program SEP III TOPICS 7

Table 3 t

Topic.s Excluded Based on Ph'ase II Experience (Not Addressed by NREP),

Topic No.

5 P Title II-1.A Exclusion Area Authority and Control.

Population Distribution I1-1.B Containment-Structural Integrity Tests;j III-7.D III-8.C Irradiation Damage, Use of Sensitized 5,tainless Steel and Fatigue Resistance-XV-20 Radiological Consequences of Fuel Damaging Accidents ~(Inside.-

and Outside Containment)

~~

XIII-2 Safegurds/ Industrial Security I

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SEP III TOPICS 8

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IIII-10.A Thermal Overload Protection for Motors of Motor-operated Valves The purpose of this topic.was to review the thermal overload protection devices '

on engineered safety feature valves to assure that they meet the appropriate IEEE standard.

In this manner, proper functioning of the valve could be assured in the event of an accident, because bypass circuits could be used to override the pro-An AEOD tection devices to drive the valve to its proper post-accident position.

.epor.t dated May 1982, " Survey of Value Operator-Related Events Occurring During 1978, 1979 and 1980" (AEOD/C203), concludes that the " recommendation or guidance to bypass thermal overload protective devices associated with safety-related valve motor operators should be reassessed," because failures of the protective cir-cuits do not contribute significantly to valve operator failures, while operator motor burn-out does. On this basis we concl.ude that this topic should be deleted for future SEP reviews.

III-10.B Pump Flywheel Integrity The pump flywheel integrity of the PWR plants in Phase II was reviewed to deter-mine whether the intent of Regulatory Guide 1.14 was met.

Four of the five PWR facilities in Phase II have coolant pumps with flywheels; Yankee Rowe has a canned-motor type pump.

(1) materir The topic review was mainly concerned with the following three areas:

and fabrication, (2) design, and (3) inservice inspection.

The general tough-ness of the flywheel materials evaluated were found to be' adequate, the margins against ductile rupture at normal operating speed and design overspeed were found to. exceed the RG 1.14 requirements, and the plant technical specifications generally has inservice inspection programs whic.h could be readily made to satisfy the P.G 1.14 inservice inspection requirements for pump flywheels.

The experience of motor-flywheel combinations similar to the Palisddes design has been very favorable.

Moreover, there have been no flywheel failures or Therefore, associated flywheel problems experienced with this type of design.

we recommend that this topic be deleted for future SEP reviews.

1 9

SEP III TOPICS

'ompliance with Codes and Standards (10 CFR 50.55a)

C V-1 I

The purpose of this topic review was to determine whether the inservice inspec-

. tion and testing programs complied with the,ASME Code, Sections III and XI, such that the integrity of vessels, piping, and comp 6nents would be maintained throughout the service life.

In the fall of 1979, 10 CFR 50.55a was changed to require that inservice inspection and inservice testing programs be updated every 120 months.

The basis for this change was that changes between the original l

and updated programs were-found to be minor.

These programs are routinely l

reviewed by the NRC'and experience has.shown that only minor hardware modifica-l tions, if any, result from these reviews.

Thesereviewsareindep'ndhntofSEP e

and, therefore, we recommend that this topic be excluded for futur,e S'EP reviews.

l V-2 Applicable Code Cases The objective of this topic was to assure that only those Code Cases acceptable to the NRC are used in the design, fabrication, or-repair of'the plant, as reflected in Regulatory Gudies 1.84 and 1.85.

This topic only l

effectively relates to future plant modifications because design considerations are covered by Topic III-1.

The NRC reviewes proposed plant modifications for all operating reactors on an ongoing basis, including-applicable Code, Cases referenced for the design.

Therefore, we recommend that this topic be excluded for future SEP reviews.

V-4 Safe-End Integrity PWR and BWR facilities may have components in the reactor coolant pressure that have been fabricated from sensitized stainless. steel. The scope of this topic ir the Phase II review was to investigate the safety aspects that affect PWR and BWi piping and safe-end integrity for compliance with 10 CFR Part 50,. including fracture toughness, flaw evaluation, stress corrosion cracking, and control of materials and welding.

SEP III TOPICS 10

1 Topic V-4 Ws originally deleted from SEP on May 7,1981 because USI A-42, " Pipe Cracks in Boiling Water Reactors," and NRR Generic Activity C-7, "PWR System Piping," addressed all of.the safety concerns of the topic.

NUREG-0313, "Techni.

cal Report on Material Selection and Processing Guidelines for BWR Coolant Pres-sure Boundary Piping," and NtfREG-0691, " Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors," detailed the technical reso-lution of the two generic issues, respectively.

Implementation of the recom-mendations of the NUREG reports was sufficient to satisfy the safety concerns of Topic V-4.

l In light of the continued occurrence of stress corrosiert cracking in the sensi-tized stainless steel safe-ends, Topic V-4 was reopened to accelerate the imple-mentation schedule.

However, the SEP review.was limited only to the issue.of stress corrosion cracking of sensitized stainless steel safe-ends. The other concerns in Topic V-4 will be resolved in the generic implementation program for NUREG-0313 (NUREG-0691 contained essentially the same technical resolution as did NUREG-0313).

l The revised Topic V-4 should not be part of SEP Phase III because all BWRs that still had sensitized stainless steel safe-ends were included in SEP Phase II.

Therefore, we recommend that this topic be deleted for future SEP reviews.

I V-6 Reactor Vessel Integrity

.The quality level initially built into a nui: lear reactor pressure vessel is an important factor affecting the integrity of.the vessel throughout its service f

life. This quality level is directly related to the materials and construction

. practices used in the manufacture of the vessel. The scope of this topic is the review of the safety aspects that affect BWR an'd PWR reactor vessel integrity throughout service life. These aspects include fracture toughness, neutron irradiation, evaluation of material surveillance programs, pressure-temperature limitations, inservice. inspections, flaw evaluation and transient analyses.

All of these safety aspects are routinely reviewed by the staff during the license application review process and periodically during the service life of SEP III TOPICS 11

.z.- - -- -

e o facility.

This review process is covered by the requirements of Appendices G and H to 10 CFR Part 50 with the acceptance criteria detailed in Sections III and XI of the ASME Boiler and Pressure Vessel Code, in Regulatory Guide 1.99 and i

in ASTM Standard E185-73.

Additionally, the safety concern of pressurized therma shock is currently being inwestigated in USI A-49.

These reviews are independent of SEP and, therefore, we recommend that this topic be deleted for future SEP reviews.

1 V-10.A Residual Heat Removal Heat Exchanger Tube Failures RHR heat exchangers are designed to remove residual and decay heat so that the reactor can be placed in a safe cold shutdown condition and to maintain core cooling following a postulated loss-of-coolant accident. A leak in an RHR heat exchanger tube will, with normal pressure differentials, cause an inflow of cooling water into the primary system introducing impurities, such as i

chlorides, into the primary coolant water.

The additional failure of the RHR pressure control system, for a BWR, would allow leakage of radioactive fluids directly to the environment.

The scope of this topic was proposed for SEP Phase-II to include all sources of intersystem leakage.

I Leakage of radioactive primary coolant water to interconnecting fluid' systems, i

l l

except from PWR steam generators, is covered in detail in SEP Phase III Topic 3.1 "RCPB Leakage Detection." Leakage of primary coolant through failed PWR steam generator tubes is addressed in USIs A-3, " Westinghouse Steam Generator Tube l

Integrity," A-4, " Combustion Engineering Steam GeneratorTube Integrity," and A-5, " Babcock & Wilcox Steam Generator Tube Integrity." Leakage of radio-active primary water to secondary or service water systems is unlikely since the primary water is at a lower pressure than the interconnecting systems.

This is true in BWRs, except for pressure control system failures, and in PWRs, except for steam generator tube leakage.

Also, if leakage of radio-active primary coolant were to occur, each operating plant has the capability to limit the release to the environment in accordance with 10 CFR Part 50, Appendix 1, " Numerical Guides for Design Objectives and L,imiting Conditions for Operation to Meet the Criterion ' As Low As Is Reasonable Achievable' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents."

SEP III TOPICS 12

o j

Leakage of contaminants (such as chlorides) into BWR primary cooling water is addressed by SEP Phase III Topic 6.3, "RCS Water Purity (BWR)." Leakage of contaminants into.PWR primary cooling water is only likely when the primary '

system is depressurized such as during refueling.

Before the system is returned to power, the primary water is routinely monitored.

(

Several probabilistic risks assessments have been performed on this fssue to compare its relative importance to safety.

In a' memorandum from S. H. Hanauer to H. R. Denton, " Preliminary Ranking of NRR Generic Safety Issues," dated March 26, 1982, this topic was rated low on a risk / cost scale.

Since this topic is addressed by other issues and since the relative risk is low, we, therefore, recommend that this topic be deleted for future SEP reviews.

t VI-2.C. Ice Condenser Containment The objective of this topic was to assure the functional performance of the ice condenser containment system based on operating experience and avail -

ability of an independent analysis capability by the staff. Since this topic f

would only affect one plant in future SEP reviews (D. C. Cook) and the results of surveillance testing from.the plant have been the. subject of an ongoing NRC review, we recommend that this topic be excluded for future SEP reviews.

VI-2.D Mass and Energy Release for Possible Pipe Breaks Inside Containment t

l VI-3 Containment Pressure and Heat Removal Capability l

The objective of these topics was to assure that the containment design conditions are adequate in consideration o'f improvements of the containment response analysis techniques.

There are related ongoing NRC reviews associated with environmental qualification (e.g., multi plant action B-69) w'hich address the containment response to steam line breaks.

For loss-of-coolant accidents, the Phase II experience has shown that the differences in analysis methods and assumptions have not resulted in a significant difference in the containment i

SEP III T'0 PICS 13

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response nor has this difference resulted in an issue of safety significance.

Therefore, we recommend that these topics be excluded for future SEP reviews.,,

VI-6 Containment Leak Testing VIII-1.A Potential Equipment Failures Associated wi~th a Degraded Grid Voltage IX-6 Fire. Protection The issues associated with these-topics are multi plant action items which are'being reviewed by the staff for all plants on a+ generic basis.

In order to avoid duplication of effort, we recommend that these issues by excluded for future SEP reviews.

However, any backfit requirements-not yet implemented for these generic actions will be considered during the integrated assessments to coordinate Phase III requirements and implementation schedules.

VII-6 Frequency Decay The objective of this topic was to investigate the effects of an offsite power system frequency decay which could, depending on~the rate of decay, provide an electrical brake on the coolant pumps that would slow the pumps faster than the assumed flywheel coastoown flow rates normally used in the loss-of-flow accident analysis.

This issue is limited to PWRs and certain grid characteristics.

During the Phase II~ reviews there were not any plants th.at were affected by this issue. However, in consideration of the likelihood of occurence of a frequency decay excessive enough to go beyond the transient analysis and the atmospheric release'that might result from that portion of the gap activity that would be released to the primary coolant, we conclude that such an event would represent a negligible portion of the overall risk l

predicted in the Reactor Safety Study (WASH-1400). Although the consequence of an under-frequency transient may be shown to be more severe than the conse-quences for a normal loss-of-flow event, for some plants, no undue _ risk to the health and safety of the public is involved. Ccnsequently, we recommend that this topic be excluded for future SEP reviews.

SEP III TOPICS 14

~

VIII-3.A Station Battery Capacity Test Requirements During the SEP Phase II reviews, we found that half of the plants did not satisfy the requirements for battery testing.

In addition to other reliability concerns, the high importance.of the battery capacity to risk reduction is described in NUREG-0666, "A Probabilistic Safety Analysis of DC Power Supply Requirements for Nuclear Power Plants." In response to the concerns expressed in NUREG-0666 we are presently developing generic guidance to improve the reliability of DC systems at operating reactors which will include battery capacity testing.

This guidance will be presented to the CRGR.

Therefore, we recommend that this topic be handled generically and be.. excluded for future SEP reviews.

VIII-4 Electrical Penetrations of Reactor Containment Currently the staff requires Electrical Penetration Assembly Designs to incor-porate redundant class 1E overcurrent protection to maintain containment integrity during worst case short-circuit current conditions and to maintain mechanical integrity after single random failures of circuit overload protection devices.

i Previous risk assessments have shown that containment leakage (i.e., small holes less than 4" in diameter), as the primary mode of containment failure following This is l

a postulated core melt accident, is not significant to the total risk.

due to the existence of higher consequence modes of containment failure which also Con-

.have much higher probabilities of occurrence than this topic issue does.

tainment failure due to penetration overcurrent.has been shown to be a low probability event (low probability of high current combined with high reliability of overcurrent protection devices, including those devices that are not Class 1E).

Also, an electrical penetration failure would only constitute a small leakage Consequ'ently, this issue does not contribute significantl'y to the total pathway.

containment leakage probability.

l Therefore, we recomment that this topic be deleted from future SEP. reviews.

i 1

SEP III TOPICS 15

XI-I Appendix I XI-2 Radiological (Effluent and Process) Monitoring Systems These topics are part of ongoing NRC generic activities (NRR A-02, B-35, and B-67).

For this reason, these topics were deleted'during the Phase II reviews.

B-67 was subsequently removed from the list of generic issues.

These activities have resulted primarily in Technical Specification changes.

Consequently these generic' activities are not expected to affect the integrated assessments for SEP plants. We recommend that these topics.be excluded for future SEP reviews; XV-1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow A number of transients which are expected to occur with moderate frequency, and which involve an unplanned increase in heat removal by the secondary system, are covered by this topic.

Excessive heat removal, i.e., a heat removal rate in excess of the. heat generation rate in the core, causes a decrease in moderator temperature which increases core reactivity and can lead to a power level increa and a decrease in shutdown margin. The power level increase will lead to a reac tor trip.

Any unplanned power level increase may result in fuel damage or exces sive reactor system pressure.

In Phase II reviews, operation of the turbine bypass system was assumed for the feedwater controller failure event for BWR's, although it is not classified as a safety system.

Feedwater controller failure is one of the design basis events routinely reevaluated during core reload reviews.

In future reloads, if r;redit is given for turbine bypass operability, limiting conditions for operation and surveillance requirements will be implemented in the Technical Specifications.

Feedwater controller failure without bypass could result in a larger, reduction i critical power ratio and a higher increase in pressure than the event with bypas With the margins provided in the plant design, the magnitude of the increases is not expected to lead to a significant de' gradation in safety.

SEP III' TOPICS 16

Another issue raised in the Phase II reviews was a continued increase in feedwater flow.

This was the only issue of safety significance identified for the PWR's in 1

Phase II.

For many plants the, flow increase is terminated by turbine protection systems or operator action in a relatively short time.

The effects of' delayed operator action or a. single failure in the control systems have not been eval-uated.

If the overfeeding continued long enough, water ingress into the steam line space could occur, resulting in stresses that the system may not be designed to withsta.nd.

The issue of steam generator (PWR) or reactor. vessel (BWR) overfill and the interface with control systems is subtask 1 of Unresolved Safety Issue A-47,

" Safety Implications of Control Systems." Therefore, we recommend that this topic be excluded for future SEP reviews.

XV-2 Spectrum of Steam System Piping Failures Inside and Outside of Containment (PWR)

The objective of this topic was to assure that the primary system and contain-ment response to a spectrum of steam line breaks is within the design bases and the resultant ~offsite doses are a small fraction of the 10 CFR 100 guide-lines.

These issues are being reviewed in conjunction with multi plant action B-69 (see topics VI-2.D and VI-3) for all plants. This generic evalua-tion is considering core return to power and single failures, with special attention to main feedwater isolation..

In the Phase II reviews, no.significant recriticality or differences in containment response due to continuous feedwater addition were identified.

In addition, NRR generic activity A-22 and pressurized thermal shock address aspects of this issue.

Topics III-5.A and III-5.8 address l

the effects on migigating systems, structures, and components.

Since fuel fail-ures are typically not predicted for steam line breaks, the resultant doses are

~

usually not controlling (i.e., less than a steam generator tube rupture).

There-fore, we recommend that this topic be excluded for future SEP reviews.

l nclosure 3 SEP III TOPICS 17

-7

l XV-3 Loss'of External Load, Turbine

Trip, Loss of Condenser Vacuum, Closure of Main Steam Isolation Valve (BWR), and Steam Pressure Regulatory Failure (closed).

XV-7 Reactor' Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break i

~I The. issues associated with these topics will be addressed by the NREP analyses.

During the Phase II. reviews, differences were only noted for one plant for each j

topic.

For IV-3, the difference will be resolved in conjunction with XV-1, relating to the feedwater transients. Those issues will carry over for logical consequences for an event that'does not rupture either the primary or secondary boundaries.

However, the resultant doses are net controlling.

Therefore, we recommend that these topics be exclused for future SEP reviews.

l XV-16 Radiological Consequences of Failure of Small lines Carrying Primary i

Coolant Outside Containment i

XV-18 Radiological Consequences of Main Steam Line Failure Outside Containment (BWR) i During the Phase II reviews, a. number of facilities could not meet the, ai:ceptance criteria associated with the current regulatory requirements for i

these topics.

These criteria require that the offsite doses be a.small fraction of the 10 CFR 100 guidelines.

In each case, the issue was, acceptably resolved by implementation of the Standard Technical Specification (STS) l'imits for the primary coolant activity.

For Phase III, we recommend that a limiting event be used to determine whether the STS primary coolant activity should be imposed.

{

XVI Standard Technical Specifications Those aspects of the Standard Technical Specifications (STS) that deal with

~

organization, staff, procedures, and emergency plans are being covered by.

TMI-related activities for al.1 plants.

Other aspects of the STS are being i

dealt with in conjunction with routine license amehdments, reload applications, or. generic problems as they arise.

Therefore, recommend that this topic be excluded for future SEP reviews.

SEP III TOPICS 18

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ML20027B595

Contents

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Dear Mr. Edwards:

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ML20027B595

Text

Dear Mr. Edwards:

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Enclosure:

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