NUREG-1412 

Foundation for the Adequacy 

of the Licensing Bases 

A Supplement to the Statement of Considerations 

for the Rule on Nuclear Power Plant 

License Renewal (10 CPR Part 54) 

Final Report 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 
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Foundation for the Adequacy 

of the Licensing Bases 

A Supplement to the Statement of Considerations 

for the Rule on Nuclear Power Plant 

License Renewal (10 CFR Part 54) 

Final Report 

Manuscript Completed: October 1991 

Date Published: December 1991 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555 

NUREG-1412 

ABSTRACT 

The objective of this report is to describe the regulatory processes that 

assures that any plant-specific licensing bases will provide reasonable 

assurance that the operation of nuclear power plants will not be inimical 

to the public health and safety to the end of the renewal period. It is on 

the adequacy of this process that the Commission has determined that a formal 

renewal licensing review against the full range of current safety requirements 

would not add significantly to safety and is not needed to assure that 

continued operation throughout the renewal term is not inimical to the public 

health and safety or common defense and security. 

This document illustrates in general terms how the regulatory process has 

evolved in major safety issue areas. It also provides examples illustrating 

why it is unnecessary to re-review an operating plant 1 s licensing basis, except 

for age-related degradation unique to license renewal, at the time of license 

renewal. 

This report is a supplement to the Statement of Considerations for the Nuclear 

Regulatory Commission 1 s rule (10 CFR Part 54) that establishes the criteria 

and standards governing nuclear power plant license renewal. 

iii 

TABLE OF CONTENTS 

Page 

ABSTRACT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 

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

1.1 Objective of This Analysis.................................. 1-1 

1.2 Scope of This Analysis...................................... 1-1 

1.3 Technical and Policy Overview............................... 1-1 

1.4 Conclusions................................................. 1-6 

2 SITE-RELATED ISSUES ............................................. . 2-1 

2.1 Geography, Demography, and Potential Site-Proximity Hazards. 2-1 

2. 2 Meteorology................................................. 2-3 

2.3 Hydrologic Engineering...................................... 2-5 

2.4 Geologic, Seismologic, and Geotechnical Engineering......... 2-7 

3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS ........ . 3-1 

3.1 Scope....................................................... 3-1 

3.2 Conformance With NRC General Design Criteria................ 3-1 

3.3 Classification of Structures, Components, and Systems....... 3-3 

3.4 Wind, Tornado, and Flood Protection......................... 3-4 

3.5 Missile Protection.......................................... 3-5 

3.6 Protection Against Dynamic Effects Associated With 

Po~tu~ated ~upture of Piping................................ 3-6 

3. 7 Se1sm1c Design.............................................. 3-7 

3.8 Design of Seismic Category I Structures..................... 3-9 

3.9 Mechanical Systems and Components........................... 3-10 

3.10 Seismic and Dynamic Qualification of Mechanical and 

Electrical Equipment........................................ 3-12 

3.11 Environmental Design of Mechanical and Electrical Equipment. 3-13 

4 REACTOR ......................................................... . 4-1 

4.1 Scope....................................................... 4-1 

4.2 Fuel System Design.......................................... 4-1 

4. 3 Nuclear Design.............................................. 4-2 

4.4 Thermal and Hydraulic Design................................ 4-3 

4.5 Reactor Materials........................................... 4-3 

4.6. Functional Design of Reactivity Control Systems............. 4-5 

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS .................... . 5-1 

5.1 Integrity of Reactor Coolant Pressure Boundary and Reactor 

Vessels..................................................... 5-1 

5.2 Component and Subsystem Design.............................. 5-4 

V 
TABLE OF CONTENTS (Continued) 

6 ENGINEERED SAFETY FEATURES ...................................... . 6-1 

6.1 Scope....................................................... 6-1 

6.2 Metallic and Organic Materials.............................. 6-1 

6. 3 Containment Systems......................................... 6-2 

6.4 Emergency Core Cooling Systems.............................. 6-5 

6.5 Habitability Systems........................................ 6-7 

6.6 Fission Product Removal and Control Systems................. 6-8 

7 INSTRUMENTATION AND CONTROL SYSTEMS ............................. . 7-1 

7 .1 Scope....................................................... 7-1 

7.2 Development of Regulatory Requirements...................... 7-1 

7. 3 Reactor Trip System......................................... 7-1 

7.4 Engineered Safety Features Actuation Systems................ 7-3 

7.5 Safety-Related Display Instrumentation...................... 7-4 

7.6 Safe Shutdown and All Other Systems Required for Safety..... 7-5 

7. 7 Control Systems ........................ , . . . . . . . . . . . . . . . . . . . . 7-6 

7.8 General Conclusions......................................... 7-6 

8 ELECTRIC POWER .................................................. . 8-1 

• 

8.1 Scope....................................................... 8-1 

8.2 Safety Issues and Regulatory Requirements................... 8-1 

8.3 Evolution of Current Licensing Basis........................ 8-1 

8.4 Conclusions................................................. 8-4 

9 AUX! LI ARY SYSTEMS ............................................... . 9-1 

9.1 Scope....................................................... 9-1 

9.2 Safety Issues and Regulatory Requirements................... 9-1 

9.3 Evolution of Current Licensing Basis........................ 9-1 

9.4 Conclusions................................................. 9-4 

10 STEAM AND POWER CONVERSION SYSTEMS............................... 10-1 

10.1 Scope....................................................... 10-1 

10.2 Safety Issues and Regulatory Requirements................... 10-1 

10.3 Evolution of Current Licensing Basis......................... 10-1 

10.4 Conclusions................................................. 10-3 

11 RADIOACTIVE WASTE MANAGEMENT SYSTEMS............................. 11-1 

11.1 Scope....................................................... 11-1 

11.2 Safety Issues and Regulatory Requirements................... 11-1 

11.3 Evolution of Current Licensing Basis........................ 11-1 

11.4 Conclusions................................................. 11-2 

vi 
TABLE OF CONTENTS (Continued) 

12 RADIATION PROTECTION............................................. 12-1 

12.1 Scope....................................................... 12-1 

12.2 Safety Issues and Regulatory Requirements................... 12-1 

12.3 Evolution of Current Licensing Basis........................ 12-2 

12.4 Conclusions................................................. 12-4 

13 CONDUCT OF OPERATIONS............................................ 13-1 

13.1 Management, Operations, and Technical Support Organizations. 13-1 

13. 2 Training.................................................... 13-2 

13.3 Emergency Planning.......................................... 13-4 

13.4 Review and Audit............................................ 13-6 

13. 5 Pl ant Procedures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-8 

13. 6 Phys i ca 1 Security. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-10 

14 INITIAL TEST PROGRAM............................................. 14-1 

14.1 Safety Issues and Regulatory Requirements................... 14-1 

14. 2 Evo 1 ut ion of Current Licensing Basis. . . . . . . . . . . . . . . . . . . . . . . . 14-1 

14.3 Conclusions................................................. 14-1 

15 ACCIDENT ANALYSES................................................ 15-1 

15.1 Scope....................................................... 15-1 

15.2 Safety Issues and Regulatory Requirements................... 15-1 

15.3 Evolution of Current Licensing Basis........................ 15-1 

15. 4 Conclusions... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-2 

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

16.1 Scope....................................................... 16-1 

16.2 Safety Issues and Regulatory Requirements................... 16-1 

16.3 Evolution of Current Licensing Basis........................ 16-1 

16.4 Conclusions................................................. 16-2 

17 QUALITY ASSURANCE................................................ 17-1 

17 .1 Scope....................................................... 17-1 

17.2 Safety Issues and Regulatory Requirements................... 17-1 

17. 3 Evo 1 ut ion of Current Licensing Basis. . . . . . . . . . . . . . . . . . . . . . . . 17-1 

17.4 Conclusions................................................. 17-2 

18 HUMAN FACTORS ENGINEERING........................................ 18-1 

18.1 Scope....................................................... 18-1 

18. 2 Control Room................................................ 18-1 

18. 3 Safety Parameter Display System............................. 18-3 

vii 
19 RESOLUTION OF SAFETY ISSUES: TECHNICAL AND 

IMPLEMENTATION STATUS......... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19-1 

19.1 Scope....................................................... 19-1 

19.2 Safety Issues and Regulatory Requirements................... 19-1 

19.3 Regulatory Process and Implementation Status................ 19-1 

19.4 Conclusions ............................ ·..................... 19-3 

APPENDIX A Unresolved Safety Issues for which A Final T~chnical 

Resolution Has been Achieved 

viii 
1 INTRODUCTION 

1.1 Objective of This Analysis 

The objective of this analysis is to describe the regulatory processes that 

assures that the plant-specific licensing bases provide reasonable assurance 

that the operation of nuclear power plants will not be inimical to the public health and safety. It is because of the adequacy of this process that the 

Commission has determined that a formal renewal licensing review against the 

full range of current safety requirements would not add significantly to safety and is not needed to assure that continued operation throughout the renewal 

term is not inimical to the public healt~ and safety. 

This report supplements the Statement of Considerations for the Nuclear Regulatory Commission's rule (10 CFR Part 54) that establishes the criteria and 

standards governing the renewal of a nuclear power plant's operating license. 

1.2 Scope of This Analysis 

This document discusses in general terms how the regulatory process has evolved 

in areas of major safety concerns. It also provides examples illustrating why it is unnecessary to re-review an operating plant's licensing basis, except for 

age-related degradation unique to license renewal. Plant-specific details of 

how the regulatory processes have been implemented for specific technical areas 

can be found in the records of each plant's license applications and licenses. 

These are maintained in the docket files. 

The statement of considerations for the license renewal rule includes an 

overview of the basis for the Commission conclusion that the regulatory processes provide reasonable assurance that the current licensing basis (CLB) provides an acceptable level of safety and that licensees comply with their 

CLB. The elements of the regulatory process relied on are identified and 

explained. Among these elements are the original licensing basis, the 

workings of the Commission's backfit policy, and the roles of operating event 

monitoring and safety issue resolution. In the present document, that overview 

is restated and is also supported by the addition of a substantial detailed 

examination of the CLB adequacy basis for the full range of specific areas of 

major safety issues. In the examinations for each of these areas, the safety issues involved are described, the key features of the regulatory requirements 

are noted, the evolution of the current licensing bases is explained, and conclusions are presented, stating once more the main foundations for the continued 

acceptability of the CLB for older as well as newer plants. 

1.3 Technical and Policy Overview 

1.3.1 Principles of the License Renewal 

The license renewal rule rests on two key principles. The first principle is 

that, with the exception of age-related degradation unique to license renewal, 

the regulatory process provides reasonable assurance that the licensing bases 

1-1 
of all currently operating plants provide and maintain an acceptable level of 

safety for operation during any renewal period. The second and equally 

important principle is that each plant 1 s current licensing basis must be maintained throughout the renewal period, in part through a program of age-degradation management for systems, structures, and components that are important for 

license renewal. This report relates to the first of these principles that 

the regulatory process provides reasonable assurance of adequate protection to 

the public health and safety. 

1.3.2 Current Licensing Basis 

The current licensing basis (CLB) is the set of NRC requirements applicable 

to a specific plant and a licensee 1 s written commitments for assuring compliance 

with and operation within applicable NRC requirements and the plant-specific 

design basis (including all modifications and additions to such commitments 

over the life of the license) that are docketed and in effect. The CLB 

includes the NRC regulations contained in 10 CFR Parts 2, 19, 20, 21, 30, 

40, 50, 51, 54, 55, 70, 72, 73, 100, and appendices thereto; orders; license 

conditions, exemptions; and technical specifications. It also includes the 

plant-specific design-basis information defined in 10 CFR 50.2 as documented 

in the most recent final safety analysis report (FSAR) as required by 10 CFR 50.71 

and the licensee 1 s commitments remaining in effect that were made in such 

docketed licensing correspondence as such licensee responses to NRC bulletins, 

generic letters, and enforcement actions, as well as licensee commitments 

documented in NRC safety evaluations, or as described in licensee event reports. 

The CLB differs among plants. These differences arise because plants are licensed at different times, at different sites, with different designs and individual operating experience. This document describes and discusses the regulatory processes designed to ensure that notwithstanding these differences the 

CLBs of all plants remain acceptable throughout plant life, including any renewal period. This document also notes the role of the backfit process in 

incorporating newly evolving requirements in previously licensed plants and the 

contribution made to safety by staff and industry guidance. Analyses in this 

document show in specific detail how these evolutionary processes have worked 

in the various safety-issue areas to provide reasonable assurance that the NRClicensed nuclear power plants provides and maintains an acceptable level of 

safety. 

1.3.3 Acceptable Level of Safety 

The Atomic Energy Act of 1954, as amended, directs the Commission to ensure 

that nuclear power plants operation is not inimical to the health and safety of 

the public. However, this 11 not inimical 11 standard contemplated neither an 

absolute protection or zero risk and, therefore safety improvements beyond the 

minimum needed for adequate protection are possible. As new information is 

developed on technical subjects, the NRC reviews the potential safety concerns 

and then requires that plant designs be able to cope with the identified concern with sufficient safety margins and reliable systems. Should new information reveal an unforeseen significant safety concern or insufficient margins 

and backup capability, the new information is carefully evaluated and the Commission may, in light of the information, conclude that the existing regulations need to be changed in order to continue to assure an acceptable level of 

safety, or that some other regulatory action needs to be taken. Therefore, as 

the Commission identifies new issues or concerns, reasoned engineering deci1-2 
sions are made within the Commission about whether any additional measures must 

be taken at plants to resolve the issues. When specific actions are identified, 

the Commission, through its regulatory programs, can modify the licensing bases at operating plants at any time to resolve the new concern. The process of backfitting requirements to plants already licensed is currently guided by the provisions of the backfit rule (10 CFR 50.109). Before promulgation of the 

backfit rule, similar considerations were applied, although the backfit rule 

gave the process more structure. 

1.3.4 Regulatory Oversight 

The Commission 1 s regulatory oversight programs ensure that the plant 1 s licensing basis is modified as appropriate to reflect significant new information on technical topics, including the effects of age-related degradation affecting the 

design or operation of the licensed plant so that the licensing bases at operat- ing plants continue to provide an acceptable level of safety. These continuing activities in place during the initial license term would continue to the end of 

the renewal term as well. Examples of such programs include operating events 

assessment and generic issues programs, discussed in the subsections that follow, 

as well as the Commission's inspection program. Should the Commission find that 

additional protection is needed to ensure the public health and safety or if 

significant additional protection at a reasonable cost would substantially enhance plant safety, the Commission may require the backfit of a licensed plant, that is, the addition of, elimination of, or modification to plant systems, 

structures, or components. 

Operating Events Assessment 

The Commission has an aggressive program for reviewing operating events at 

nuclear power plants. As a requirement of the current licensing basis, and one 

which would continue to the end of the renewal term, each licensee must notify the Commission promptly of any plant event that meets or exceeds the threshold 

defined in 10 CFR 50.72 and must file written licensee event reports (LERs) for 

those events that meet or exceed the threshold defined in 10 CFR 50.73. The 

staff reviews this information daily and follows up on events that appear to be 

potentially risk significant or are Judged to be a possible precursor to a more 

severe event. Depending on the significance of the reported event, the staff 

may take further action to notify one or more classes of licensees or to impose additional requirements on some or all licensees. Industry groups, notably the 

Institute of Nuclear Power Operations (INPO), disseminate information in signi- ficant operating experience reports (SOERs). The total process offers a high degree of assurance that events which are potentially risk significant or are 

precursors to potentially significant events are being reviewed and resolved 

expeditiously. 

Generic Issue Programs 

As described in SECY-89-138, the Commission also maintains an active program for evaluating and resolving generic safety issues that may impact public health 

and safety. A generic safety issue (GS!) involves a safety concern that may affect the design, construction, or operation of all, several, or a class of 

reactors or facilities. Its resolution may have a potential for safety improvements and promulgation of new or revised requirements or guidance. The prioriti- zation process, as described in NUREG-0933, evaluates the safety significance of 

an issue and classifies the issues as high, medium, or low priority GSis. GSis 

1-3 
that are categorized as high priority are further evaluated to determined whether 

they involve questions regarding adequate protection of the public health and 

safety and therefore should be re-categorized as unresolved safety issues (USis). GSis are issues that involve enhancements to safety but do not call into question the adequacy of the current licensing basis. By contrast, USis are defined as 

issues that potentially involve adequate protection of the public health and 

safety. Thus, a US! may represent a matter where the adequacy of the current 

licensing basis has not been established Resolution of a US! may result in a 

determination that action is necessary to ensure adequate protection, or it 

may result in a conclusion that, in fact, there are no concerns as to adequate protection of the public health and safety and further action is not warranted. 

The licensing basis of individual plants includes changes that have resulted 

from resolution of generic issues determined to be applicable and will include 

applicable generic-issue-derived changes in the future. It should be noted, 

however, that, as discussed immediately below, all currently unresolved GS!s 

address only enhancements of safety. This conclusion was determined during the 

initial evaluation of the generic concern, which assessed whether any aspect of 

the generic concern might have a significant impact on the protection of the 

public health and safety such that immediate action would be warranted. Since 

GSis involve only enhancements to safety, implementation of the resolution of 

any GSI is not necessary to ensure public health and safety during the renewal 

term. 

A special group of 22 GSis deemed to be of sufficient significance to warrant 

both a high-priority resolution effort and special attention in tracking were 

designated as Unresolved Safety Issues (USis). All USis have been resolved. 

Most of the USI resolutions have been implemented; the remainder are being implemented on a satisfactory schedule. In one case, USI A-46, "Seismic Qualifi- cation of Equipment in the Operating Plants, 11 the NRC and the utility groups 

are negotiating the implementation schedule in accordance with the NRC policy 

on integrated Schedule for Plant Modifications, Generic Letter 83-20, dated 

May 9, 1983. 

This process for ensuring implementation of these remaining USis is the same 

process used by the NRC in the past to ensure resolution and implementation of 

USis. Furthermore, this process will be used in the future if the NRC identifies new issues that meet the definition of a USI. 

The generic issues program is described and discussed more fully in Chapter 19. 

1.3.5 Evolution of NRC Requirements 

In the late 1960s and early 1970s, the U.S. Atomic Energy Commission 1 s (now Nuclear Regulatory Commission) scope of review of proposed power reactor designs 

was evolving. In the same way that the scope of review evolved, the licensing bases for early plants have evolved as appropriate actions were identified 

through the NRC 1 s continuing oversight activities and regulatory processes. In 

1967, the Commission published for comment and interim use a proposed "General 

Design Criteria (GDC) for Nuclear Power Plants" that established minimum requirements for the principal design standards. The GDC were formally adopted in 1971 

and have been used since that time as guidance in reviewing new plant applica- tions. The general design criteria are contained in Appendix A to 10 CFR Part 50. 

They establish minimum broad requirements for the principal criteria for the 

materials, design, fabrication, testing, inspection, and certification of all 

structures, components, equipment, and systems in nuclear power plants that are 

1-4 
important to safety. The staff's plant-specific reviews with respect to the 

various areas of safety (discussed in Chapters 2 through 19) must arrive at a 

conclusion that the overall plant design satisfies the intent of the GDC requirements and that the plant can safely be operated. 

Safety guides issued in 1970 became part of the regulatory guide series in 1972. 

These guides described methods acceptable to the staff for implementing specific portions of the regulations, including certain GDC, and formalized staff techniques for reviewing a facilitX. In 1972, the Commission distributed for information and comment a proposed 'Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, 11 now Regulatory Guide 1.70. It gave applicants 

a standard format for these reports and identified the principal information 

needed by the staff for its review. The Standard Review Plan (SRP), NUREG-75/087, 

was published in December 1975 and revised in July 1981 (NUREG-0800) to give the staff more guidance for improving the quality and uniformity of its reviews. 

This guidance consisted of acceptance criteria and review procedures necessary 

to provide the staff with the basis for concluding that applicable GDC have 

been satisfied. For the most part, the detailed acceptance criteria prescribed in the SRP were not new; rather, they were methods of review that, in many cases, -- had not been previously published in any regulatory document. 

1.3.6 Systematic Evaluation Program 

In 1977 the NRC initiated the Systematic Evaluation Program (SEP) to review the 

designs of 10 of the oldest operating nuclear power plants and thereby confirm 

and document their safety. The reviews were organized into approximately 90 

review topics (reduced by consolidations from 137 originally identified). 

The SEP effort highlighted a group of 27 regulatory topics for which corrective 

action was generally found to be necessary for the initial SEP plants and. for 

which safety improvements for other operating plants of the same vintage could 

be expected. The topics on this smaller list are referred to as the SEP lessons 

learned, and the Commission expects that these topics would be generally applic- able to operating plants that received their construction permits in the late 

1960s or early 1970s. 

Of the 27 regulatory topics highlighted in the SEP effort, four have been completely resolved and one is of such low safety significance as to require no 

regulatory action. Of the 22 issues remaining open, the Commission has determined that none require immediate action to protect the public health and safety. 

The Commission has incorporated the 22 issues into the established regulatory 

process for determining the safety importance of GSis and has determined that 

none require immediate action as part of a license renewal application. As 

with the case for GSis and USis, the existing prioritization process to be used 

during the review and prioritization of the SEP lessons learned issues should 

prove to be adequate in the future to resolve these issues. 

1.3.7 Review of Changes in NRC Requirements and Guidance 

The NRC has arranged for systematic review to help ensure the effectiveness, 

efficiency, and coherence of changes in requirements and guidance. The Committee To Review Generic Requirements (CRGR) has the responsibility to review 

and recommend to the Executive Director for Operations (EDO) approval or disapproval of requirements or staff positions to be imposed by the NRC staff on 

1-5 
one or more classes of power reactors. This review applies to staff proposals of requirements or positions that reduce existing requirements or positions and 

proposals that increase or change requirements. This responsibility is imple- mented in such a manner as to ensure that the provisions of 10 CFR 2.204, 10 CFR 

50.109, and 10 CFR 50.54(f) pertaining to generic requirements and staff posi- tions are carried out by the staff. The CRGR process aims to eliminate or remove any unnecessary burdens placed on licensees, reduce the exposure of workers 

to radiation in adhering to some of these requirements, and conserve NRC 

resources -- all to be done without impairing the adequate protection of the 

public health and safety and with furthering the review of new, cost-effective 

requirements and staff positions. The CRGR and the associated staff procedures 

are intended to ensure NRC staff implementation of 10 CFR 50.54(f) and 50.109 

for generic backfit matters. By having the committee submit recommendations 

directly to the EDO, only a single agency-wide point of control is transmitted. 

For those rare instances where it is judged that an immediately effective action 

is needed to ensure that facilities pose no undue risk to the health and safety of the public (10 CFR 50.109(a)(4)(ii)), no prior review by the CRGR is necessary. However, before or after any such action, the staff conducts a documented 

evaluation that includes a statement of the objectives of and reasons for the 

actions and the basis for invoking the exception. 

In earlier years, the function of the CRGR to determine the need for backfit 

was performed by the Regulatory Requirements Review Committee; the CRGR process introduced enhanced discipline and documentation to the determination. 

Throughout the process of evolution of requirements and staff guidance, the 

Commission has had the benefit of advice on significant safety issues from the 

Advisory Committee on Reactor Safeguards (ACRS). 

1.3.8 Detailed Analyses 

Chapters 2 through 19 of this report describe and discuss, in more specific 

terms, how the previously stated regulatory programs and processes have worked 

in major administrative, technical, and procedural areas, thereby detailing the 

Commission 1 s reasons for considering it unnecessary to review an operating plant 1 s licensing basis, except for age-related degradation unique to license 

renwal, at the time of license renewal. These discussions are indicative of 

how the regulatory process will continue to ensure that, despite the variations 

in plant design criteria, an operating reactor will continue to provide an 

acceptable level of safety. 

1.4 Conclusions 

The processes outlined above and discussed more substantially in the specific 

areas on major safety issues supports as the justification for focusing the NRC 

review to only age-related degradation unique to license renewal. The key elements include the original licensing basis and the Commission rules, regula- tions, requirements, and reviews underlying it; the Gommission 1 s backfit 

policy, which has historically resulted in backfit of new requirements when 

required for adequate protection of public health and safety, compliance with 

the plant licensing basis, or when cost-justified as safety enhancements; and 

the Commission 1 s programs of inspection, of monitoring operational events, and 

for resolution of plant-specific and generic safety issues. It is through these processes and programs that the staff ensures the continued acceptability of the licensing bases of all plants. 

1-6 
2 SITE-RELATED ISSUES 

The following section discusses a number of site-related topics that include 

general site geography and demography, potential site-vicinity hazards, and 

potential accidents from natural phenomena. The natural phenomena discussed in 

this section include regional climatology and local meteorology, site hydrology and potential flooding issues, as well as geology and related seismic considerations. The nature of the licensing basis for each of these topics will be 

discussed in greater detail below. 

2.1 Geography, Demography, and Potential Site-Proximity Hazards 

2.1.1 Scope 

Site geography includes the consideration of site location and description, exclusion area authority and control, and broad land use patterns. Demography 

covers the population distribution in the vicinity of the site. With respect 

to both, geography and demography, changes over the licensing term are addressed by making conservative projections for land use and population changes. 

Nearby industrial, transportation, and military facilities can pose a threat to 

the safe operation of a nuclear power plant. Site acceptability depends, in 

part, on the adequacy of the licensee 1 s assessment of and protection against the hazards posed by nearby man-made activities. Although these activities 

present a wide range of external events, the hazards associated with them can 

be grouped into four broad categories: (1) missiles, (2) explosions, (3) fires, and (4) toxic gases. 

2.1.2 Safety Issues and Regulatory Requirements 

The Commission regulations require that site evaluation factors be considered 

in the review of a license application, including those relating to site location, exclusion area, low population zone, and population center distance. In 

addition, the regulations require that plant systems, structures, and components important to safety be appropriately protected against the dynamic effects from events and conditions outside the nuclear power unit. 

2.1.3 Evolution of Current Licensing Basis 

Prior to the issuance of the Standard Review Plan (SRP), the staff reviewed 

site exclusion area control and demography on a case-by-case basis, with empha- sis on independent verification of site-specific characteristics such as property ownership, mineral rights, and nearby population distributions. It was 

recognized that land use and local and State regulations can and do change with 

time. Hence, for example, staff review of the licensee 1 s ability to exercise 

appropriate authority and control of the activities within the exclusion area 

focused on those aspects that were deemed to be time-dependent. 

With respect to demography, efforts were made to obtain reasonable projections of population distributions to the end of the licensing term. This was done in 

2-1 
recognition that population changes, driven by such factors as local socioeconomic conditions, can be potentially significant with passage of time. 

Similarly, potential external hazards in the vicinity of the site that could 

affect plant safety were reviewed on a case-by-case basis, with emphasis on 

independent verification of individual hazards for each site. Review of numerous license applications in the early and middle 1970s led to a two-part ap- proach in addressing site hazards. First, the site and its surroundings were 

examined with the intent of identifying all existing hazards that had a poten- tially significant impact on the safe operation of the plant. Once the hazards 

were identified, they were evaluated in terms of their severity and likelihood. 

For hazardous industrial or military activities, conservative projections were 

made concerning transportation traffic or accident rates. Where projections 

were not feasible, conservatism commensurate with the potential changes to the 

hazardous materials operations was applied to the current conditions. 

The SRP was published in the early 1970s to improve the quality and uniformity of staff review in the particular subject areas as well as to specify acceptance criteria of the staff for concluding that the applicable regulations had 

been satisfied. The detailed acceptance criteria contained in the SRP were not 

new, but rather standardized the acceptance practices which had been established in previous reviews. 

As mentioned earlier, the regulations require that reactors be protected against the dynamic effects of events and conditions outside the nuclear plant. For external hazards such as toxic gases or airplane hazards, the licensees 

frequently provided information, during the initial licensing review, concerning the frequency of shipments or the amount of airplane traffic and then pro- posed plant protection features or determined that no protection was necessary based on a determination or analysis of the hazard. Because the regulations remain in effect for the term of the license, licensees have the responsibility 

to ensure that the plant remains appropriately protected from any site-related 

hazards, new or existing at the time the plant was licensed. The staff recog- nizes that licensees cannot control development around the site and the regula- tions do not require licensees to do so. However, the regulations clearly place the responsibility on licensees for ensuring the protection of the 

reactor. 

The Commission inspection activities also provide another source of information 

concerning changes that are occurring around the reactor site. The resident 

inspector, who typically resides in the area of the plant, has direct knowledge and access to the local media and therefore can be informed of potential developments in the surrounding environment that can potentially affect plant safety. In addition, regional and headquarters-based inspectors routinely visit 

sites, thereby affording further opportunities for observations of potential changes of the surrounding environment. As new issues are identified, these 

issues are raised to both the licensee and the staff for resolution. The staff 

is undertaking revisions to selected inspection procedures to require routine 

documentation of potential changes in the general environs of the facility. 

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2.1.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the plant siting met the staff acceptance criteria and the intent 

of all applicable regulations. The staff has and will continue to obtain new 

information related to this subject area through a variety of sources such as 

updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance 

of adequate protection of the public health and safety. In addition, the staff 

will continue to review new information in this subject area and if the staff 

determines that new or different requirements are needed, the staff has the 

capability within the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and 

safety of the public. In conclusion, the Commission concludes that the current 

regulatory process has and will continue to provide reasonable assurance that 

the licensing basis of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

2.2 Meteorology 

2.2.1 Scope 

This section discusses the regulations and licensing requirements used by the 

NRC to ensure safe siting and operation of nuclear power plants with respect to 

meteorology. Site-specific data on meteorology and regional climatology is 

used to ensure that continuing staff awareness of both regional and local meteorological trends is maintained to ensure that the design basis meteorology conditions remain sufficient to ensure safe plant operation in accordance with 

the current licensing basis (CLB). 

2.2.2 Safety Issues and Regulatory Requirements 

Nuclear power plants are designed, operated, and maintained such that offsite 

exposure to accidental gaseous releases and their resultant dose to receptors 

at the plant exclusion area boundary, low population zone distance, and the 

population center and offsite exposures from routine normal operational releases at nearby receptors including residences, dairies, farms, schools, etc., 

comply with the requirements of 10 CFR Part 100 and 10 CFR Part 50, Appendix I, 

respectively. 

2.2.3 Evolution of Current Licensing Basis 

"Meteorology and Atomic Energy-1968, 11 AEC 1968, generally, served as a compen- dium of meteorology measurement, dispersion modeling, and dose determination 

during the early 1970s. This document was later used as the foundation for 

issuance of safety guides, regulatory guides, and the Standard Review Plan sections in the meteorology area of review. 

Regarding the regional climatology, the long-term meteorological conditions 

affecting the plant vicinity, as described in the plant final safety analysis 

report (FSAR) submitted by the licensee, and addressed in the staff safety evaluation report (SER), are not expected to change significantly. The basis 

for the expectation of minimal change lies in the meteorological and 

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climatological records collected from as far back as 1895 and published in documents prepared by the National Climatic Center. The published data 

demonstrate that nearly constant climatological conditions exist in a local 

area with only small anomalies on a monthly or seasonal basis. Studies by the 

National Institute of Standards and Technology (formerly National Bureau of 

Standards) using long-term data show the low probabilities of rapid significant changes in long-term temperature, precipitation, and wind speeds. Thus, design 

basis characteristics might only be exceeded for brief time periods in the near 

term (say, next 50 to 100 years), while greater frequency of duration of 11 abnormal11 conditions might provide evidence of a possible worldwide major climatic change. The changes of global extent and recognition of its happening would 

be expected to provide sufficient time to allow the staff to take appropriate 

measures to deal with the changes such that the current licensing basis continues to remain valid for the licensed term. 

Recently, the staff considered the changes to global climate due to atmospheric 

ozone depletion in response to concerns raised by the Council on Environmental 

Quality (CEQ) pursuant to NEPA. The thrust of the memorandum dealt with highand low-level waste repositories and impact of global warming on rainfall and 

flooding due to changes in climate. The original plant site evaluation was 

based on the premise and expectation that site area climate would generally not 

change significantly during the operational life of the plant. Based on current information and staff review experience, site data concerning regional 

climate provided in FSARs and staff conclusions contained in plant SERs are 

expected to remain valid during any extended term of the operating license. 

With respect to the review area of local meteorology, the local meteorology 

conditions, generally, reflect the expected regional climate influence and 

unique topographic features that may result in micro-scale phenomena that had 

beer. addressed in the FSAR and SER and should remain the same as described in 

those documents during any renewal period. However, in the event of a marked 

climatic change, it is probable that parameters representing 11 normals,U such as 

extreme wind, precipitation, temperature, and structural capacities may require 

reevaluation, since climatic change may result in storm systems with greater 

intensities and frequency than those assumed in the design basis of the plant. 

However, this type of reevaluation is a part of the NRC staff's continuous 

plant safety assessment effort to ensure continued adequacy of the 

meteorology-related current licensing basis for operating plants. The licensee 

routinely publishes specific information related to meteorology in semiannual 

reports of meteorological joint frequency data as required by emergency planning requirements. Plant modifications or improvements in the meteorological 

monitoring system dictated by the semiannual reports have been implemented by 

the staff or licensee to ensure the validity of the current licensing basis. 

The current licensing basis requires each plant to have an onsite meteorological monitoring program. This onsite meteorological measurement program continues to monitor local conditions that would affect the dispersion of radioactive 

and toxic gaseous effluent from or to the plant. 

2.2.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the meteorological and climatological factors related to the sit2-4 
ing of the facility met the staff acceptance criteria and the intent of all 

applicable regulations. The acceptance criteria are constructively established 

to ensure adequate protrection from extremely low probability meteorological phenomena. The design basis for these extremely low probability meteorological 

events is related with eonservation and is not experted to change appreciably 

during the next 50 to 100 year period. However, the staff has and will continue to obtain new information related to this subject area through a variety of 

sources such as updates to the FSAR, research reports, operating plant events 

and routine plant inspections. The staff reviews this information and, in the 

past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In 

addition, the staff will continue to review new information in this subject 

area and if the staff determines that new or different requirements are needed, 

the staff has the capability within the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes 

that the current regulatory process has and will continue to provide reasonable 

assurance that the licensing basis of all currently operating plants are sufficient to assure that operation is not inimical to the public health and safety. 

2.3 Hydrologic Engineering 

2.3.1 Scope 

Nuclear power plants interact continuously with their hydrosphere (e.g., rivers, lakes, coastal environments, ground water, and water control structures). 

Such interactions present potential hazards of flooding as a result of severe 

hydro-meteorological conditions. This section discusses the regulations and 

licensing requirements adopted by the NRC to ensure safe operation of nuclear 

power plants against severe flooding hazard over the licensed plant life. The 

continuous review process used by the NRC staff to assess safety impact resulting from changes in hydrometeorological parameters and new information related 

to flooding hazard is also discussed. Lastly, the rationale for an NRC staff 

conclusion that the current licensing basis for operating plants is adequate to 

protect public health and safety is discussed. 

2.3.2 Safety Issues and Regulatory Requirements 

The Commission's regulations require, in part, that systems, structures, and 

components important to safety be designed to withstand the effects of natural 

phenomena such as floods, tsunami, and seiches without loss of capability to 

perform their safety functions. In addition, the regulations require that 

physical characteristics of the site, including meteorology and hydrology, be 

taken into account in determining the acceptability of a site for a nuclear 

power reactor. More specifically, the regulations require that a detailed 

study be performed and that the design bases for seismically induced floods and 

water waves be based on the results of the required geologic and seismic investigations and that these design bases be taken into account in the design of 

the ruclear power plant. 

In order to demonstrate compliance with the above regulatory requirements, nuclear power plants are designed to prevent or mitigate the loss of capability 

for cold shutdown and maintenance thereof resulting from the most severe flood 

conditions that can reasonably be predicted to occur at a site as a result of 

severe hydrometeorological conditions, seismic activity, or both. 

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2.3.3 Evolution of Current Licensing Basis 

The NRC has recognized the potential hazards resulting from the flooding of a 

commercial nuclear power plant since the mid-1960s. The flooding hazard review 

of plants licensed in the late 1960s used a formal and disciplined evaluation 

process but on a plant-specific basis. 

In order to independently evaluate the potential for flooding at proposed reactor sites, the Atomic Energy Commission contracted the U.S. Army Corps of Engineers and the U.S. Geological Survey to evaluate flooding potential at coastal 

sites and river sites, respectively. In 1970 the AEC staff developed specific guidance for use in determining flood protection requirements for all plant sites. The staff adopted the concept of the "Probable Maximum Flood" from the 

Corps of Engineers and applied this concept to sites along streams and rivers. 

Guidance for determining the Probable Maximum Hurricane Surge, Probable Maximum 

Seiche, and Probable Maximum Tsunami flooding was also developed and applied for plant sites along lakes and oceans. · 

During 1973 through 1975, the NRC, based on the above work, published integrat- ed staff positions and acceptance criteria related to acceptable design of nuclear power plants against flooding hazard. 

As part of the Commission 1 s process of reviewing new information related to 

specific technical issues, the staff learned that a number of probabilistic risk assessments completed between 1981 and 1987 concluded that external flooding could be a key contributor to overall plant risk. As a result, the NRC 

staff determined that it would be useful to evaluate the continued adequacy of 

the existing flood protection regulations by performing an individual plant examination at each reactor site. The staff developed regulatory guidance and 

acceptance criteria for the evaluation of plant-specific vulnerabilities to 

beyond design bases events initiated from a severe flooding event and incorpo- rated this evaluation into the Individual Plant Examination External Events 

(IPEEE) program. The staff intends to evaluate the results of the IPEEE program, not on a plant-specific basis, but as an aggregate to determine appropriate changes to the regulations related to protections from potential external 

flooding. If any changes are identified, modification of the regulations will 

proceed and implementation of any plant modifications would be required on 

plant-specific bases to meet the revised regulations. 

In October 1989, NRC issued Generic Letter 89-22 to inform licensees that the 

staff has adopted for future plants the latest probable maximum precipitation (PMP) criteria published by the National Oceanic and Atmospheric Administration 

(NOAA), National Weather Services (NWS), to establish acceptable design config- urations for safety-related nuclear power plant facilities. In this letter the 

staff also concluded and stated that the existing criteria for determining PMP 

at operating plants still provided reasonable assurance of the protection of 

the health and safety of the public and that no additional backfit action by licensees was necessary. 

2.3.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the design to protect the facility from floods met the staff ac2-6 
ceptance criteria and the intent of all applicable regulations. The staff has 

and will continue to obtain new information related to this subject area 

through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information in this subject area and if the staff determines that new or different 

requirements are needed, the staff has the capability within the existing regu- latory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, 

the Commission concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing basis of all currently operating plants are sufficient to assure that operation is not inimical to the 

public health and safety. 

2.4 Geologic, Seismologic, and Geotechnical Engineering 

2.4.1 Scope 

This section discusses the regulations and licensing requirements adopted by the NRC to ensure safe operation of nuclear plants subject to the influence of 

site-specific geologic, seismologic, and geotechnical hazards over the licensed 

plant life. The ongoing review process used by NRC staff to assess the safety impact resulting from changes in parameters related to the hazards and pertinent new information is also discussed. 

2.4.2 Safety Issues and Regulatory Requirements 

The Commission 1 s regulations require that systems, structures, and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be per- formed and that nuclear power plant systems, structures, and components important to safety be designed to withstand the effects of natural phenomena such 

as earthquakes without loss of capability to perform their safety functions. 

In addition, the regulations require, in part, that suitable redundancy be pro- vided for the cooling water system to ensure that its safety function can be 

accomplished and that measures be established to ensure design control, material control, special processes control, and inspection and test controls. The 

regulations also require that all nuclear power plants be designed so that, if 

the safe shutdown earthquake (SSE) occurs, all safety-related systems, structures, and components remain functional. 

In order to demonstrate compliance with these regulatory requirements, nuclear 

power plants are designed to prevent or mitigate the loss of capability for 

cold shutdown and maintenance thereof resulting from the safe shutdown earthquake, foundation settlement, or instability. 

2.4.3 Evolution of Current Licensing Basis 

When the NRC first began to review nuclear power plant seismic designs in the 

1960s, the reviews were performed using a formal evaluation process but on a 

plant-specific approach. 

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In 1971, the General Design Criteria (GDC) for Nuclear Power Plants were formally adopted as the minimum requirements for the principal design standards. 

These requirements generally adopted the existing staff practice in effect at 

that time. The GDC have been used as guidance in reviewing new plant applica- tions since then. 

Because of the evolutionary nature of licensing requirements and the development of technology over the years, nuclear power plants employ a broad spectrum of design features and requirements depending on when the plant was designed and constructed, who was the manufacturer, and when the plant was licensed for 

operation. 

The NRC staff, as part of its routine review process, continues to assess the 

potential safety impact of any new information related to geologic and seismologic issues. New information can be derived from research or additional data 

observations since the issuance of operating licenses. Whenever the results of 

such information indicated the need for remedial actions, including plant modifications, the staff has acted to ensure the implementation of such actions to 

ensure that the current licensing basis at potentially affected plants remains 

adequate to protect the health and safety of the public. 

Examples of this include the discovery of the capable Hosgri fault near Diablo 

Canyon and the assessment of earthquakes occurring near Maine Yankee. One nuclear power plant (Humboldt Bay) and one non-power reactor (General Electric 

Test Reactor) were shut down, and remain permanently shut down, as a direct 

result of geologic concerns. Another example is the 11Charleston Earthquake Issue 11 which was raised as a result of a U.S. Geologic Survey letter in 1982. 

This letter highlighted the possibility that large damaging earthquakes have 

some likelihood of occurring at locations not formally considered in past licen~ing decisions. The staff initiated the Seismic Hazard Characterization 

Project, which provided probabilistic seismic hazard estimates for nuclear power plant sites east of the Rocky Mountains. A stmilar project was carried out 

by the Electric Power Research Institute (EPRI) for the electric utility industry. The staff 1 s purpose in evaluating the probabilistic studies has been to 

identify plants in the central and eastern United States where past licensing decisions have resulted in the potential for plant-specific vulnerabilities to 

beyond design basis events with respect to seismic hazard. The staff 1 s plan for documenting and reconfirming the degree of protection from seismic safety issues is part of the staff resolution of Generic Issue A-46, 11 Seismic Qualification of Equipment in Operating Nuclear Power Plants. 11 

The purpose of the A-46 is to reverify and document the seismic adequacy of 

mechanical and electrical equipment qualification to ensure the survival and 

functionality of equipment required to safely bring the reactor and plant to a 

safe shutdown condition. Consistent with the guidance for developing an unresolved safety issue, the staff performed an analysis to determine if the identified seismic concern might have a significant impact on the protection of the 

public health and safety and, therefore, that immediate remedial action would 

be warranted. The staff 1 s conclusions and their technical bases have been pub- lished in NUREG-1211, 11 Regulatory Analysis for Resolution of Unresolved Safety Issue A-46, Seismic Qualification of Equipment in Operating Plants, 11 February 1987. In this document, the staff concluded that equipment installed in nuclear power plants is inherently rugged and not susceptible to seismic damage. However, the staff also concluded that sufficient justification of a safety 

2-8 
benefit could be made because, although the equipment was inherently rugged and 

not susceptible to seismic damage, failures resulting from seismic loads were 

possible if the equipment was not adequately supported or anchored. As a 

result, the staff issued Generic Letter 87-02 to all operating reactors, which 

required, as a backfit under 50.109, that licensees reverify the seismic quali- fication and anchorage of installed equipment to provide additional assurance 

of the continued protection of public health and safety. 

In the staff review of areas related to plant foundation stability/settlement, 

water control structural safety and heat sink integrity, etc., the staff has 

generally upgraded established geotechnical engineering criteria and methodologies, which have been widely used to ensure full compliance with the necessary regulatory requirements. 

As part of the staff's routine review effort, any new geotechnical engineering data or analysis techniques, which are judged by the staff as pertinent for 

inclusion in the current licensing criteria, have been incorporated following established NRC procedures. Where appropriate, the new analysis techniques 

were applied to assess design adequacy and any plant-specific modifications 

were implemented on plant-specific bases in order to ensure continued validity of the plant-specific licensing basis. Examples of such staff actions are 

(1) resolution of Waterford basemat cracking and structural integrity issue, 

(2) resolution of North Anna buried piping settlement and support integrity issue, and (3) resolution of San Onofre Unit 1 settlement of foundation and 

buried equipment issue. 

As part of the staff's effort to assess the continued adequacy of the existing regulations, the staff has implemented the Individual Plant Examination External Events program. As part of this program, licensees will be requested to 

look for potential plant-specific vulnerabilities to beyond design basis accidents initiated from postulated seismic events and to report their findings to 

the Commission. The staff intends to use the results from all the plants in 

aggregate to determine if deficiencies exist in present regulations governing seismic hazards. If such deficiencies are identified, the staff intends to 

modify the regulations as necessary and would require plant-specific modifications as necessary to establish compliance with the new regulations. 

2.4.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the seismic design of the facility met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will 

continue to obtain new information related to this subject area through a variety of sources such as updates to the FSAR, research reports, operating plant 

events and routine plant inspections. The staff reviews this information and, 

in the past, has required licensees to take actions to upgrade the plant to 

provide continuing assurance of adequate protection of the public health and 

safety. For example, a program currently in process is requiring licensees to 

confirm and document the seismic qualification and anchorage of equipment and, 

where necessary, to make plant specific modifications to ensure the continued 

adequate protection of public health and safety. As part of the continuing assessment of the adequacy of the existing regulations, the Commission has re2-9 
quested licensees to evaluate their plants for potential vulnerabilities for 

beyond design basis accidents resulting from geologic, seismologic, and 

geotechnical hazard considerations. If the results of this evaluation indicate 

that the existing regulations need modification, then the staff will proceed to 

revise the regulations and require plants to meet the revised regulations. In 

addition, the staff will continue to review new information in this subject 

area and if the staff determines that new or different requirements are needed, 

the staff has the capability within the existing regulatory process to require 

additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes 

that the current regulatory process has and will continue to provide reasonable 

assurance that the licensing basis of all currently operating plants are sufficient to assure that operation is not inimical to the public health and safety. 

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3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 

3.1 Scope 

This section addresses the principal criteria required for the materials, design, fabrication, testing, inspection, and certification of all structures, 

components, equipment, and systems that are important to safety. Important to 

safety is defined in the introduction to 10 CFR Part 50, Appendix A, as those 

systems, structures, and components that provide reasonable assurance that the 

facility can be operated without undue risk to the health and safety of the 

public. The topics covered by this section are: the NRC General Design Criteria; classification of systems, structures, and components; wind and tornado 

loadings and water level (flood) design; missile protection; protection against dynamic effects associated with postulated rupture of piping; seismic design; design of Seismic Category I structures, mechanical systems, and components; seismic and dynamic qualification of Seismic Category I mechanical and electrical equipment; and environmental design of mechanical and electrical equipment. 

3.2 Conformance With NRC General Design Criteria 

3.2.1 Safety Issues and Regulatory Requirements 

10 CFR Part 50, Appendix A, contains General Design Criteria (GDC) that establish minimum broad requirements for the principal criteria mentioned in 

Subsection 3.1 above. 

3.2.2 Evolution of Current Licensing Basis 

In the late 1960s and early 1970s, the U.S. Atomic Energy Commission 1 s (now 

Nuclear Regulatory Commission) scope of review of proposed power reactor designs was evolving and somewhat less defined than it is today. The requirements for acceptability evolved as new facilities were reviewed. In 1967, the 

Commission published for comment and interim use proposed General Design Criteria (GDC) for Nuclear Power Plants that established minimum requirements for 

the principal design standards. The GDC were formally adopted in 1971 and have 

been used as guidance in reviewing new plant applications since that time. 

Safety guides issued in 1970 became part of the Regulatory Guide Series in 

1972. These guides describe methods acceptable to the staff for implementing specific portions of the regulations, including certain GDC, and formalize 

staff techniques for performing a facility review. In 1972, the Commission 

distributed for information and comment a proposed 11 Standard Format and Content 

of Safety Analysis Reports for Nuclear Power Plants, 11 now Regulatory Guide 1.70. It provided a standard format for these reports and identified the 

principal information needed by the staff for its review. The Standard Review 

Plan (SRP), NUREG-75/-087, was published in December 1975 and updated in July 1981 (NUREG-0800) to provide further guidance for improving the quality and 

uniformity of staff reviews. This guidance consisted of acceptance criteria 

and review procedures necessary to provide the staff with the basis for concluding that applicable GDC have been satisfied. For the most part, the detailed acceptance criteria prescribed in the SRP were not new; rather they were 

3-1 
methods of review that, in many cases, were not previously published in any regulatory document. 

Because of the evolutionary nature of the licensing requirements discussed 

above and the developments in technology over the years, operating nuclear power plants embody a broad spectrum of design features and requirements depending 

on when the plant was constructed, who was the manufacturer, and when the plant 

was licensed for operation. The amount of documentation that defines these 

safety-design characteristics also has changed with the age of the plant. Although the earlier safety evaluations of operating facilities did not address 

many of the topics discussed in current safety evaluations, all operating facilities have been reviewed more recently against a substantial number of major safety issues that have evolved since the operating license was issued. Conclusions of overall adequacy with respect to these major issues (e.g., emergency core cooling system, fuel design and pressure vessel design) are a matter of 

record. On the other hand, a number of other issues (e.g., seismic considerations, tornado and turbine missiles, flood protection, pipe break effects inside containment, and pipe whip) were not originally reviewed against today's 

acceptance criteria for many operating plants. 

The Systematic Evaluation Program (SEP) was initiated by the staff in 1977 to 

review the designs of older operating nuclear power plants in order to enhance 

the documentation of their safety. The review provided (1) an assessment of 

the significance of differences between current technical positions on safety issues and those that existed when a particular plant was licensed, (2) a basis 

for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The results of the 

staff's SEP reviews are documented in a series of Integrated Plant Safety Assessment Reports. 

3.2.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the plant design met the staff acceptance criteria, and the intent 

of all applicable regulations. The staff has and will continue to obtain new 

information related to this subject area through a variety of sources such as 

updates to the FSAR, research reports, operating plant events and routine plant 

inspections. The staff reviews this information and, in the past, has required 

licensees to take actions to upgrade the plant to provide continuing assurance 

of adequate protection of the public health and safety. In addition, the staff 

will continue to review new information in this subject area and if the staff 

determines that new or different requirements are needed, the staff has the 

capability within the existing regulatory process to require additional analyses 

or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current 

regulatory process has and will continue to provide reasonable assurance that 

the licensing basis of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

3-2 
3.3 Classification of Structures, Components, and Systems 

3.3.l Safety Issues and Regulatory Requirements 

The Commission regulations require that systems, structures, and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be per- formed. Regulatory Guide 1.26, 11Quality Group Classifications and Standards 

for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants,U contains staff guidance that may be used to determine how a 

plant-specific design satisfies the regulations. 

In addition, the regulations require that systems, structures, and components important to safety be designed to withstand the effects of earthquakes without 

loss of capability to perform their safety functions. Regulatory Guide 1.29, 11 Seismic Design Classification/ contains staff guidance that may be used to 

determine how a plant-specific design satisfies the applicable regulations by identifying all systems, structures, and components that should be classified 

as Seismic Category I. 

3.3.2 Evolution of Current Licensing Basis 

During the 1960s, when General Design Criteria and the ASME Section III Code 

for Nuclear Power Plant·components were evolving, the staff reviews of quality 

group and seismic classification were performed on a plant-specific basis. 

Decisions on the classifications for systems, structures, and components were 

made based on staff positions at that time relative to the importance to safety of each item. These positions were first documented in Safety Guides 26 and 29 

in 1970 and 1971 and later in Regulatory Guides 1.26 and 1.29 in 1972. 

Subsequent to an exchange of correspondence between the General Electric Company and the staff during 1973 and 1974, the staff developed its position on 

classifications of the main steam and feedwater lines for BWR/6 plants. This 

position allowed BWR/6 applicants the option of installing a third shutoff 

valve and a seismic restraint downstream of the outside isolation valve in 

these lines. If the applicant chose this option, then the main steam and 

feedwater lines could be classified as Quality Group D (Non-Nuclear Safety) and 

non-Seismic Category I downstream of the seismic restraint. This position was 

implemented by the staff during its reviews of all BWR/6 plants between 1975 

and 1981. 

3.3.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the the quality group and seismic classification met the staff 

acceptance criteria and the intent of all applicable regulations. The staff 

has and will continue to obtain new information related to this subject area 

through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information in this subject area and if the staff determines that new or different 

3-3 
requirements are needed, the staff has the capability within the existing regu- latory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, 

the Commission concludes that the current regulatory process has and will 

continue to provide reasonable assurance that the licensing basis of all currently operating plants are sufficient to assure that operation is not inimical 

to the public health and safety. 

3.4 Wind, Tornado, and Flood Protection 

3.4.1 Safety Issues and Regulatory Requirements 

Commission regulations require, in part, that systems, structures, and components important to safety shall be designed to withstand the effects of natural 

phenomena such as tornadoes, hurricanes, and floods. 

3.4.2 Evolution of Current Licensing Basis 

The Commission has recognized the potential hazards resulting from wind, tornadoes, and floods imposed on a nuclear power plant since the early 1960s. These 

early reviews were performed on a plant-specific basis and the staff's guide- lines for acceptability evolved as new facilities were reviewed. In 1975, the 

staff published sections of the Standard Review Plan addressing these subject 

areas. These guidance documents formally published regulatory positions that 

had been in practice at that time but had not been previously published in any 

type of regulatory document. Minor revisions were made to these SRPs in 1981; 

however, guidelines therein have remained virtually unchanged since 1975. 

3.4.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the designs for protection against tornadoes, and floods met the 

staff acceptance criteria and the intent of all applicable regulations. The 

staff has and will continue to obtain new information related to this subject 

area through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews 

this information and, in the past, has required licensees to take actions to 

upgrade the plant to provide continuing assurance of adequate protection of the 

public health and safety. In addition, the staff will continue to review new 

information in this subject area and if the staff determines that new or different requirements are needed, the staff has the capability within the existing regulatory process to require additional analyses or plant modifications, 

as necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has and 

will continue to provide reasonable assurance that the licensing basis of all 

currently operating plants are sufficient to assure that operation is not inimical to the public health and safety. 

3-4 
3.5 Missile Protection 

3.5.1 Safety Issues and Regulatory Requirements 

Although large steam turbines and their auxiliaries are not safety-related systems as defined by NRC regulations, failures that occur in these turbines can 

produce large, high-energy missiles. If such missiles were to strike and to 

damage plant safety-related systems, structures, and components, they could 

render them unavailable to perform their safety function. Consequently, the 

regulations require, in part, that systems, structures, and components important to safety be appropriately protected against the effects of missiles that 

might result from such failures. 

3.5.2 Evolution of Current Licensing Basis 

The standard review plans for turbine missile protection were first published in NUREG-75/087 in December 1975 and updated in NUREG-0800 in July 1981. Regu- latory Guide 1.115 was revised in July 1977. The staff's guidance in these 

documents indicates that the hazard rate for the loss of essential safety systems from a single turbine missile event must be less than 10- 7 per year. Plants constructed prior to the publication of the staff guidance and evaluated 

as part of the Systematic Evaluation Program (SEP) were reviewed to ensure that 

an adequate level of protection from turbine missiles existed at these plants. Turbine missiles are identified as SEP Topic III-4.B. 

In the initial reviews of this topic, the value for the probability of turbine 

failure resulting in ejection of turbine fragments through the turbine casing (P1 ) was assumed be a constant of 10-4 per year for all turbines. Licensees or 

applicants evaluated the strike probability (P2 ) and the damage probability (P3 ) to ensure that the product of P1 , P2 , and P3 is less than 10- 7 per year. These early reviews indicated that large uncertainties exist in P2 and P3 because of the difficulty in modeling missiles, barriers, obstacles, and trajec- tories and in determining critical impact energies. In an Electric Power 

Research Institute sponsored seminar on 11 Turbine Missile Effects in Nuclear 

Power Plants 11 in October 1982, the staff indicated that, because of the uncertainties in P2 and P3 , they would emphasize the missile generation probability (P1 ) in future turbine missile reviews. P2 and P3 probabilities are to be 

order-of-magnitude estimates that are dependent on the orientation of the turbine to essential safety systems. The revised method also ensures that the 

hazard rate would be less than 10- 7 per year. In this method the staff evaluates the procedures and methods used by turbine manufacturers to calculate the 

total missile generation probability and the associated turbine maintenance and 

inspection procedures. 

3.5.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the protection from turbine missiles met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will 

continue to obtain new information related to this subject area through a variety of sources such as updates to the FSAR, research reports, operating plant 

events and routine plant inspections. The staff reviews this information and, 

in the past, has required licensees to take actions to upgrade the plant to 

provide continuing assurance of adequate protection of the public health and 

3-5 
safety. In addition, the staff will continue to review new information in this 

subject area and if the staff determines that new or different requirements are 

needed, the staff has the capability within the existing regulatory process to 

require additional analyses or plant modifications, as necessary, to ensure the 

continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has and will continue to provide reasona~le assurance that the licensing basis of all currently operating plants 

are sufficient to assure that operation is not inimical to the public health 

and safety. 

3.6 Protection Against Dynamic Effects Associated With Postulated Rupture of 

Piping 

3.6.1 Safety Issues and Regulatory Requirements 

Commission regulations require that systems, structures, and components important to safety be appropriately protected against the dynamic effects that may result from equipment failures, including the effects of pipe whipping and discharging fluids. 

3.6.2 Evolution of Current Licensing Basis 

In 1972, the staff documented the deterministic criteria that the staff had 

been using for several years as guidelines for selecting the locations and orientations of postulated pipe breaks inside containment and for identifying the 

measures that should be taken to protect safety-related systems and equipment from the dynamic effects of such breaks. Prior to use of these deterministic 

criteria, the staff used non-deterministic guidelines on a plant-specific basis. The staff criteria were subsequently revised and issued in May 1973 as 

Regulatory Guide 1.46, 11 Protection Against Pipe Whip Inside Containment. 11 

Prior to 1972, the staff did not require postulation of pipe breaks outside 

containment. However, as a result of the continuing review of plant safety during that time period, the staff determined that such breaks should be postu- lated and the effects of these breaks should be evaluated by all licensees of 

operating plants and all applicants for Construction Permits or Operating Licenses. Therefore, generic letters were sent to all licensees and applicants from late 1972 through mid-1973. These letters provided deterministic criteria 

to be used for postulating pipe breaks outside containment and guidelines for 

evaluating the dynamic effects of these breaks. The letters requested that all 

recipients submit a report to the staff that summarized each plant-specific analysis of this issue. All operating reactor licensees and license applicants submitted the requested analyses in separate correspondence or updated the 

safety analysis report for the proposed plant to include the analysis. The 

staff reviewed all of these submitted analyses and prepared safety evaluations 

for all plants. 

In November 1975, the staff published Standard Review Plan sections that 

slightly revised the two generic letters discussed above. As a part of its 

plant-specific reviews between 1975 and 1981, the staff used the guidelines in 

Regulatory Guide 1.46 for postulated pipe breaks inside containment and SRPs 

3.6.1 and 3.6.2 for outside containment. In July 1981, SRPs 3.6.1 and 3.6.2 

were revised to be applicable to both outside and inside containment. On 

June 19, 1987, Generic Letter 87-11 was issued to provide revised guidelines for locations of postulated pipe ruptures. 

3-6 
Another example of the continuing review of new technical issues is the poten- tial problem of asymmetric loading on reactor vessel supports following a pos- tulated reactor coolant pipe rupture. In 1975, the staff was informed that 

asymmetric loading on the reactor vessel supports resulting from a postulated 

reactor coolant pipe rupture at the vessel nozzle had not been eonsidered in 

the original design of PWR plants. Following a brief review of this problem, the staff determined that a reevaluation of the reactor coolant system of all 

PWR plants was necessary to determine its capability to withstand these new 

loads. Unresolved Safety Issue (USI) A-2 was originated to address this prob- lem. All licensees of PWR operating plants were requested to submit 

plant-specific analyses. In response to this request, several licensees formed 

an owners group and, in lieu of an analysis, submitted a report that incorpo- rated advanced fracture mechanics techniques to demonstrate that a full diameter break could not occur in their primary loop piping. In Generic 

Letter 84-04, dated February 1, 1984, the staff agreed that such a break was 

unlikely to occur, provided it could be demonstrated by deterministic fracture 

mechanics analyses that postulated through-wall flaws in plant-specific piping would be detected by the plant 1 s leakage monitoring systems long before the 

flaws could grow to unstable sizes. The concept underlying such analyses is 

referred to as "leak-before-brea~• (LBB). Subsequent evaluations of this issue 

by the staff led to the so-ca 11 ed 11 broad scope rule, 11 which revised GDC-4 in 

1987 to permit the use of LBB-type analyses in both PWRs and BWRs. 

3.6.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the evaluations of the effects of postulated pipe breaks both 

inside and outside containment met the staff acceptance criteria and the intent 

of all applicable regulations. The staff has and will continue to obtain new 

information related to this subject area through a variety of sources such as 

updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance 

of adequate protection of the public health and safety. In addition, the staff 

will continue to review new information in this subject area and if the staff 

determines that new or different requirements are needed, the staff has the 

capability within the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and 

safety of the public. In conclusion, the Commission concludes that the current 

regulatory process has and will continue to provide reasonable assurance that 

the licensing basis of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

3.7 Seismic Design 

3.7.1 Safety Issues and Regulatory Requirements 

The Commission regulations require, in part, that systems, structures, and components important to safety be designed to withstand the effects of earthquakes without loss of capability to perform their safety functions and provides, in 

part, criteria required to determine the suitability of the plant design bases 

that were established by consideration of the seismic characteristics of the 

proposed plant site. 

3-7 
3.7.2 Evolution of Current Licensing Basis 

The Commission has recognized the potential safety-related consequences of the 

occurrence of a significant seismic event at a nuclear power plant site since 

the staff first began reviewing applications for licenses in the late 1950s. 

These early reviews were performed on a plant-specific basis, and the staff's 

guidelines for acceptability evolved as new facilities were reviewed. 10 CFR 

Part 100, Appendix A, was later issued to establish the seismic design basis 

for systems, structures, and components. 

In 1973, Regulatory Guides 1.60 and 1.61 were issued to provide staff positions relative to seismic input levels to be used for plant designs to ensure adequate consideration of historical data, site characteristics, and material behavior. In 1975, SRPs 3.7.1, 3.7.2, and 3.7.3 were issued to provide detailed 

guidelines for analytical modeling techniques for seismic analyses. These 

guidelines were used by the staff to determine that a plant-specific design satisfies applicable portions of the Commission's regulations. These SRPs were 

revised in 1981 to reflect changes in these guidelines since 1975. For example, one of the significant changes was related to the staff position on 

soil-structure interaction which was based on 1975 state-of-the-art. This po- sition was widely debated among industry, ACRS, and the staff. These debates 

provided the basis for significant industry research effort on this issue during the 1970s. The staffs' evaluation of this research provided the basis for 

a change in the staff position which incorporated the research recommendations 

and led to more realistic criteria. 

In the course of evaluating several plant-specific piping designs in the late 

1970s, the staff became aware of significant discrepancies between the original piping seismic analysis computer code and a staff-approved benchmark code. 

This problem led to a March 13, 1979 Order to Show Cause from the Commission, 

which resulted in the shutdown of five plants whose piping designs had involved 

the use of the suspect computer codes. The differences between the computer codes were attributed to the use of an inappropriate method of combining certain seismic-induced loads in the original codes. In April 1979, IE Bulletin 

79-07 was issued to request all licensees and applicants to review their piping analyses and determine if any of their computer codes contained the unaccept- able method of combining loads. In addition, they were requested to verify that all piping computer programs were checked against either staff-approved benchmark problems or other acceptable piping computer programs. All licensees 

were to submit reports to the Commission describing the results of their review. The staff reviewed the submittals from all licensees and applicants and 

the issue was resolved on a plant-specific basis by arriving at one of the following conclusions: 

1. The licensee or applicant used acceptable methods of combining loads in 

their piping analyses. 

2. If the original analyses used the unacceptable method of combining loads, 

all applicable piping was reanalyzed using acceptable methodology. The 

results of these new analyses showed that all piping stresses were within 

the allowable stresses of applicable ASME Section III or ANSI 831.1 Codes. 

3-8 
3.7.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the design for protection against earthquakes met the staff acceptance criteria and the intent of all applicable regulations. The staff has 

and will continue to obtain new information related to this subject area 

through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information in this subject area and if the staff determines that new or different 

requirements are needed, the staff has the capability within the existing regu- latory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, 

the Commission concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing basis of all currently operating plants are sufficient to assure that operation is not inimical to the 

public health and safety. 

3.8 Design of Seismic Category I Structures 

3.8.1 Safety Issues and Regulatory Requirements 

The Commission regulations contain various requirements for the design and construction of concrete and steel containments as well as requirements for all 

other Seismic Category I structures both inside and outside containment. 

3.8.2 Evolution of Current Licensing Basis 

When the NRC staff first began to review design of Seismic Category I structures in the 1960s, its scope and depth of review were not well defined and the 

staff acceptance was generally based on an ad hoc and plant-specific approach 

that provided adequate protection of the gener'ar-public. 

Staff review of design adequacy of containment and other Category I structures 

has been generally upgraded to use established structural design criteria and 

methodologies. The primary codes used in the early 1970s to review the design adequacy were the Building Code Requirements for Reinforced Concrete (ACI-318) 

and AISC, 11Specification for Design, Fabrication, and Erection of Structural 

Steel for Buildings,t' American Institute of Steel Construction, for concrete 

and steel structures, respectively. 

In 1973, the ASME Boiler and Pressure Vessel Code, Section III, Divisions 1 and 

2, became the standards for the design of steel and concrete containments. In 

1975, the staff published the Standard Review Plans (NUREG-0800), which adopted the above-listed codes and standards with appropriate inclusion of new load 

combinations and analysis methods. The Standard Review Plans were revised in 

1981 and Sections 3.8.l through 3.8.5 of the plans form the bulk of the current 

licensing criteria for containments and Category I structures. 

3-9 
The staff, as part of its routine review process, continued to assess the po- tential safety impact of any new information related to design of structures 

and, as appropriate, caused plant modifications to be implemented for affected 

plants to ensure continued conformance with the current licensing basis (e.g., modifications of torus supports and header piping supports for all Mark I 

plants). 

3.8.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the design of seismic catetory I structures met the staff acceptance criteria and the intent of all applicable regulations. The staff has and 

will continue to obtain new information related to this subject area through a 

variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this information 

and, in the past, has required licensees to take actions to upgrade the plant 

to provide continuing assurance of adequate protection of the public health and 

safety. In addition, the staff will continue to review new information in this 

subject area and if the staff determines that new or different requirements are 

needed, the staff has the capability within the existing regulatory process to 

require additional analyses or plant modifications, as necessary, to ensure the 

continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing basis of all currently operating plants 

are sufficient to assure that operation is not inimical to the public health 

and safety. 

3.9 Mechanical Systems and Components 

3.9.1 Safety Issues and Regulatory Requirements 

The Commission regulations contain requirements to ensure that all of the different types of mechanical systems, components, and equipment will maintain 

their structural and functional integrity for the life of the plant. 

3.9.2 Evolution of Current Licensing Basis 

During the early 1960s when 10 CFR 50 General Design Criteria and the ASME 

Section III Code for Nuclear Power Plant Components were evolving, the staff 

reviews of design criteria for mechanical systems and components were performed 

on a plant-specific basis. ASME Section VIII, "Pressure Vessels," and American 

National Standards Institute (ANSI) B31.1, "Power Piping," were the two main 

design standards that were accepted by the staff to ensure the structural integrity of safety-related mechanical systems and components. In 1963, ASME 

Section III, "Nuclear Vessels;' was published and accepted by the staff as a 

replacement for ASME Section VIII. In 1969, ANSI B31.7, "Nuclear Power Pip-

; ng, 11 was pub 1 i shed and accepted by the staff as a replacement for ANSI B31. 1. 

In 1971, ASME Section III was expanded to include rules for vessels, pumps, valves, and piping. ANSI B31.7 was included in ASME Section 111-1971. In that 

same year, 10 CFR 50.55a was added to the regulations to provide a requirement for applicants to use ASME Section III for the design of reactor coolant pressure boundary components. Subsequent editions of ASME Section Ill through the 

present 1989 edition have been required by 10 CFR 50.55a for the designs of 

mechanical systems and components. 

3-10 
In 1975, Standard Review Plans 3.9.1 through 3.9.5 were issued to document 

guidelines that the staff had previously been using in its plant reviews of 

mechanical systems and components to demonstrate compliance with the applicable Commission regulations. SRP 3.9.6 was also issued in 1975 to provide guide- lines for the staff to use in evaluating inservice testing (IST) programs for 

safety-related pumps and valves in non-operating plants. At that time, there 

was no requirement for IST applicable to operating plants. Therefore, in 

February 1976, 10 CFR 50.55a(g) was revised to include specific IST requirements in accordance with ASME Section XI for all licensees of operating plants and all applicants for a license to operate. Included in this revision was a 

requirement for all licensees to submit a new IST program to the staff every 10 

years. These new programs are updated to reflect the latest ASME Section XI 

requirements and staff positions. Each licensee 1 s program is reviewed and ap- proved by the staff. 

Prior to 1979, light water reactors experienced a number of occurrences of improper performance of safety and relief valves installed in the reactor coolant 

system. As a result in 1980, the staff issued NUREG-0737, 11Clarification of 

TMI Action Plan Requirements, 11 Item II.D.1, which required all BWR and PWR 

licensees and applicants to qualify reactor coolant system safety and relief 

valves, block valves, and associated piping and supports under expected operat- ing conditions for design basis transients and accidents. In response to this 

requirement, the Electric Power Research Institute (EPRI) conducted a series of 

tests for licensees and applicants in 1981 and 1982 to demonstrate operability of the components under the required loading conditions. All licensees and 

applicants have submitted information to demonstrate applicability of the EPRI 

test results to their plant-specific equipment. The staff has reviewed all but 

a few plant-specific responses on this issue. The staff's reviews assure that 

all applicable valves, piping, and supports in each plant are enveloped by the 

EPRI test program. These reviews are scheduled for completion in 1990. 

3.9.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the design of mechanical systems and components important to 

safety met the staff acceptance criteria and the intent of all applicable regu- lations. The staff has and will continue to obtain new information related to 

this subject area through a variety of sources such as updates to the FSAR, 

research reports, operating plant events and routine plant inspections. The 

staff reviews this information and, in the past, has required licensees to take 

actions to upgrade the plant to provide continuing assurance of adequate pro- tection of the public health and safety. In addition, the staff will continue 

to review new information in this subject area and if the staff determines that 

new or different requirements are needed, the staff has the capability within 

the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has 

and will continue to provide reasonable assurance that the licensing basis of 

all currently operating plants are sufficient to assure that operation is not 

inimical to the public health and safety. 

3-11 
3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment 

3.10.1 Safety Issues and Regulatory Requirements 

The Commission regulations contain requirements to ensure that mechanical and 

electrical equipment important to safety remain operable under the full range of normal and accident loadings, including seismic. 

3.10.2 Evolution of Current Licensing Basis 

The evolution of the GDC, safety guides, regulatory guides, and SRPs is discussed in Subsection 3.2.2 above. Commission guidance was originally issued in 

March 1976. These guidelines documented staff positions on seismic and dynamic qualification of equipment that had been implemented by the staff since the 

early 1970s. The analysis and test criteria used by the staff to review this 

issue evolved rapidly between 1971 and 1980. Consequently, for plants that 

were reviewed by the staff prior to the early 1970s, the margins of safety pro- vided in equipment to resist seismically induced loads are uncertain. This 

concern led the staff to originate Unresolved Safety Issue (USI) A-46, "Seismic 

Qualification of Equipment in Operating Nuclear Power Plants. 11 

NUREG-1211, "Regulatory Analysis for Resolution of USI A-46, 11 February 1987, 

identified some operating plants to be reviewed under USI A-46 as those plants whose equipment had not been qualified by using IEEE Standard 344-1975 or later 

revision. On February 19, 1987, the staff issued Generic Letter 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operat- ing Reactors, USI A-46. 11 This letter provided the staff 1 s requirements for 

implementing the resolution of USI A-46. A utility owners group has been 

formed and is negotiating with the staff on the implementation schedule in accordance with the NRC policy on integrated schedule for plant modifications, 

Generic Letter 83-20, May 9, 1983. In addition, the staff will conduct detailed audits of a limited number of plants to verify that the licensee has 

implemented its program in accordance with GL 87-02. Enforcement or other reg- ulatory actions could result from these audits. When a licensee completes its 

followup actions, it will be required to submit a letter stating that all plant modifications or followup actions related to USI A-46 have been completed. 

3.10.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the seismic and dynamic qualification of mechanical and electrical equipment met the staff acceptance criteria and the intent of all applica- ble regulations. The staff has and will continue to obtain new information 

related to this subject area through a variety of sources such as updates to 

the FSAR, research reports, operating plant events and routine plant inspec- tions. The staff reviews this information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance 

of adequate protection of the public health and safety. In addition, the staff 

will continue to review new information in this subject area and if the staff 

determines that new or different requirements are needed, the staff has the 

capability within the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and 

3-12 
safety of the public. In conclusion, the Commission concludes that the current 

regulatory process has and will continue to provide reasonable assurance that 

the licensing basis of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

3.11 Environmental Design of Mechanical and Electrical Equipment 

3.11.1 Safety Issues and Regulatory Requirements 

The Commission regulations require that systems, structures, and components important to safety be designed to accommodate the effects of and be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. 

They are to be appropriately protected against the effects of discharging fluids. Environmental qualification is one means of satisfying the above requirement for essential components. Specific requirements for environmental 

qualification of electrical equipment important to safety are contained in 

10 CFR 50.49, which requires each licensee to establish a program for qualifi- cation of essential electrical equipment subject to harsh environmental conditions, and maintain qualification of this equipment for the lifetime of the 

plant. 

For purposes of the discussion on the environmental design basis, it should be 

noted that licensees' current environmental qualification programs will include 

equipment subject to periodic replacement and equipment that has been qualified for the currently licensed plant lifetime. A review of the program covering equipment periodically replaced is not necessary as this process will continue 

during the renewed license life. 

3.11.2 Evolution of Current Licensing Basis 

In November 1977, the Union of Concerned Scientists petitioned the Commission 

to upgrade the environmental qualification of equipment in operating facilities 

to current standards. This petition led to the Commission Memorandum and Order 

of May 23, 1980 (CLI-80-21) which provided guidance and directives to resolve 

this matter in an expeditious manner. Part of this activity included development of a new rule for environmental qualification of electrical equipment. This action culminated in issuance of 10 CFR 50.49 dated January 21, 1983. All 

licensees have implemented programs consistent with 10 CFR 50.49 and supple- mental staff guidelines to ensure the safety function of electrical equipment subjected to harsh environments (radiation, temperature, pressure, and moisture) following postulated design basis accidents. This has resulted in assurance of safe plant shutdown following loss of coolant and steam line break 

accidents. 

3.11.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the environmental qualification program for electrical equipment important to safety met the staff acceptance criteria and the intent of all 

applicable regulations. The staff has and will continue to obtain new information related to this subject area through a variety of sources such as updates 

to the FSAR, research reports, operating plant events and routine plant inspec3-13 
tions. The staff reviews this information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance 

of adequate protection of the public health and safety. In addition, the staff 

will continue to review new information in this subject area and if the staff 

determines that new or different requirements are needed, the staff has the 

capability within the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and 

safety of the public. In conclusion, the Commission concludes that the current 

regulatory process has and will continue to provide reasonable assurance that 

the licensing basis of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

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4 REACTOR 

4.1 Scope 

This chapter addresses the evaluation and supporting information reviewed by the staff to establish the capability of the reactor to perform its safety functions throughout its design lifetime under all normal operational modes,, 

including transient and steady state, and accident conditions. The evaluation 

covers the areas of fuel system design, nuclear design, thermal and hydraulic design, reactor materials, and functional design of reactivity control systems. 

4.2 Fuel System Design 

4.2.1 Safety Issues and Regulatory Requirements 

Commission regulations require that the reactor core and associated coolant, 

control, and protection systems be designed with appropriate margin to ensure 

that specified acceptable fuel design limits (SAFDLs) are not exceeded during 

any condition of normal operation, including the effects of anticipated opera- tional occurrences, and that the reactor core and associated coolant systems be 

designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity. These regulations also address the requirements of 

maintaining the capability to cool the core under postulated accident conditions. Methods of adequately predicting fuel rod failures during postulated accidents are adopted so that radioactivity release estimates are not underestimated and thereby ensure that the plant in question would continue to satisfy the related requirements of 10 CFR Part 100. Also, the acceptable fuel performance limits under a postulated loss of coolant accident are specified in 

10 CFR 50.46 and Appendix K to 10 CFR Part 50. 

4.2.2 Evolution of Current Lice·nsing Basis 

The evolution of the general design criteria (GDC), safety guides, regulatory guides, and standard review plans (SRPs) is discussed in Section 3.2.2 above. 

4.2.3 Conclusions 

The staff review at the time of the initial licensing and before each refueling of a facility determined that the information provided by the applicant was 

sufficient for the staff to conclude that the fuel system design met the staff 

acceptance criteria and the intent of all applicable regulations. The staff 

has and will continue to obtain new information related to this subject area 

through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information in this subject area and if the staff determines that new or different 

requirements are needed, the staff has the capability within the existing 

4-1 
regulatory process to require additional analyses or plant modifications, as 

necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has and will 

continue to provide reasonable assurance that the licensing basis of all currently operating plants are sufficient to assure that operation is not inimical 

to the public health and safety. 

4.3 Nuclear Design 

4.3.1 Safety Issues and Regulatory Requirements 

Commission regulations require that acceptable fuel design limits be specified that will not be exceeded during normal operation, including the effects of 

anticipated operational occurrences, and require that, in the power operating 

range, the prompt inherent nuclear feedback characteristics tend to compensate for a rapid increase in reactivity. The regulations also require that power oscillations that could result in conditions exceeding SAFDL are not possible 

or can be reliably and readily detected and suppressed; and require instrumentation and controls to monitor variables and systems that can affect the fission process over anticipated ranges for normal operation and accident 

conditions, and to maintain the variables and systems within prescribed operat- ing ranges. Further, the regulations require automatic initiation of the reactivity control systems to ensure that SAFDLs are not exceeded; that reliable 

reactivity control systems be provided under normal or accident operating conditions; and that the effects of postulated reactivity accidents neither result 

in damage to the reactor coolant pressure boundary greater than limited local 

yielding, nor cause sufficient damage to impair significantly the capability to 

cool the core. 

4.3.2 Evolution of Current Licensing Basis 

The evolution of the GDC, safety guides, regulatory guides, and SRPs is discussed in Section 3.2.2 above. 

4.3.3 Conclusions 

The staff review at tbe time of the initial licensing and before each refueling of a facility determined that the information provided by the applicant was 

sufficient for the staff to conclude that the nuclear design met the staff acceptance criteria and the intent of all applicable regulations. The staff has 

and will continue to obtain new information related to this subject area 

through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information in this subject area and if the staff determines that new or different 

requirements are needed, the staff has the capability within the existing regu- latory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. ln conclusion, 

the Commission concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing basis of all currently 

4-2 
operating plants are sufficient to assure that operation is not inimical to the 

public health and safety. 

4.4 Thermal and Hydraulic Design 

4.4.1 Safety Issues and Regulatory Requirements 

Commission regulations require that the reactor core and associated coolant, 

control, and protection systems be designed with appropriate margin to ensure 

that specified acceptable fuel design limits (SAFDLs) are not exceeded during 

any condition of normal operation, including the effects of anticipated opera- tional occurrences. 

4.4.2 Evolution of Current Licensing Basis 

The evolution of the GDC, safety guides, regulatory guides, and SRPs is discussed in Section 3.2.2 above. 

4.4.3 Conclusions 

The staff review at the time of the initial licensing and before each refueling of a facility determined that the information provided by the applicant was 

sufficient for the staff to conclude that the thermal and hydraulic design met 

the staff acceptance criteria and the intent of all applicable regulations. The 

staff has and will continue to obtain new information related to this subject 

area through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information 

in this subject area and if the staff determines that new or different requirements 

are needed, the staff has the capability within the existing regulatory process 

to require additional analyses or plant modifications, as necessary, to ensure 

the continued health and safety of the public. In conclusion, the Commission 

concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing basis of all currently operating plants 

are sufficient to assure that operation is not inimical to the public health 

and safety. 

4.5 Reactor Materials 

4.5.1 Safety Issues and Regulatory Requirements 

Commission regulations, in part, require that the reactor coolant pressure boundary have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture under operating, maintenance, testing, and postulated accident conditions. These regulations also require that 

structures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. In addition, the regulations require that one of the reactivity control systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to ensure that fuel design limits are not 

4-3 
exceeded under conditions of normal operation, including anticipated opera- tional occurrences, and that components that are part of the reactor coolant 

pressure boundary be designed to permit periodic inspection and testing of 

critical areas to assess their structural and leaktight integrity. 

To satisfy these regulations, the staff recommends that control rod drive 

structural materials, reactor internals, and core support materials be designed, fabricated, erected, and tested using the published regulatory guidance and inspected to the guidelines of Section XI, Code Class 1, of the American 

Society of Mechanical Engineers Boiler and Pressure Vessel Code (hereafter ASME 

Code). The staff reviews proposed alternatives to the recommendations in the 

current criteria to ensure that they provide an acceptable level of quality and 

safety. 

When inservice inspection requirements of Section XI of the ASME Code are determined to be impractical, the NRC, in accordance with 10 CFR 50.55a(g)(6), 

may grant relief and may impose alternative requirements that are determined to 

be authorized by law, will not endanger life or property or the common defense 

and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were 

imposed on the facility. 

The staff has implemented requirements in addition to Section III of the ASME 

Code because many of the components are constructed of austenitic stainless 

steel material that is susceptible to intergranular stress corrosion cracking in the water environment of the boiling-water reactor (BWR). 

4~5.2 Evolution of Current Licensing Basis 

Sections III and XI of the ASME Code have changed and will continue to change 

as technology and operating experience change. The staff actively participates in the process that revises the ASME Code and reviews these changes to determine whether they should be incorporated into plants• licensing bases. 

Section XI of the ASME Code contains updating provisions. It requires licensees to revise their inservice inspection program every 10 years. The 

revised programs incorporate all the changes required by ASME Section XI of the 

licensee's program and, in accordance with 10 CFR 50.55a(g)(6), the staff may 

grant relief or impose alternative requirements. 

Changes in technologies (i.e., radiation embrittlement, ultrasonic examination, 

etc.), which are not addressed by the ASME Code, are described in generic letters and regulatory guides. These letters and guides are prepared and pub- lished by the staff and become incorporated into the licensing basis of any nuclear power plant. Many of these guides recommend changes to the 

plant-specific licensing basis when environmental conditions change. The staff 

reviews these licensing changes and approves them on a plant-specific basis 

using generic acceptance criteria. 

4.5.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the control rod drive structural materials and reactor internals 

4-4 
met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will continue to obtain new information related to this subject area through a variety of sources such as updates to the FSAR, research 

reports, operating plant events and routine plant inspections. The staff reviews this information and, in the past, has required licensees to take actions 

to upgrade the plant to provide continuing assurance of adequate protection of 

the public health and safety. In addition, the staff will continue to review 

new information in this subject area and if the staff determines that new or 

different requirements are needed, the staff has the capability within the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has 

and will continue to provide reasonable assurance that the licensing basis of 

all currently operating plants are sufficient to assure that operation is not 

inimical to the public health and safety. 

4.6 Functional Design of Reactivity Control Systems 

4.6.1 Safety Issues and Regulatory Requirements 

Commission regulations require that the protection system be designed to fail 

into a safe state or into a state demonstrated to be acceptable on some other 

defined basis if such conditions as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments 

(e.g., extreme heat or cold, fire, pressure, steam, water, radiation) are expe- rienced. These regulations also require that the protection system be designed 

to ensure that SAFDLs are not exceeded for any single malfunction of reactivity control systems, such as accidental withdrawal of control rods, and require that two independent reactivity control systems of different design principles be provided. 

One of the systems uses control rods, preferably including a positive means for 

inserting the rods, and shall be capable of reliably controlling reactivity changes to ensure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for such malfunctions as stuck rods, SAFDLs are not exceeded. The second reactivity control 

system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to ensure that SAFDLs are not exceeded. One of the systems shall be capable of 

holding the reactor core subcritical under cold conditions. 

Further, these regulations require that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes 

to ensure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained; that the reactivity control systems be designed to consider the effects of postulated reactivity accidents and maintain core coolability; and that the reactivity control 

systems be designed to ensure an extremely high probability of accomplishing their safety function in the event of anticipated operational occurrences. 

4.6.2 Evolution of Current Licensing Basis 

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The evolution of the GDC, safety guides, regulatory guides, and SRPs is discussed in Section 3.2.2 above. 

4.6.3 Conclusions 

The staff review at the time of the initial licensing and before refueling of a 

facility determined that the information provided by the applicant was sufficient for the staff to conclude that the functionsl design of reactivity control systems met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will continue to obtain new information related 

to this subject area through a variety of sources such as updates to the FSAR, 

research reports, operating plant events and routine plant inspections. The 

staff reviews this information and, in the past, has required licensees to take 

actions to upgrade the plant to pro~ide continuing assurance of adequate pro- tection of the public health and safety. In addition, the staff will continue 

to review new information in this subject area and if the staff determines that 

new or different requirements are needed, the staff has the capability within 

the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has 

and will continue to provide reasonable assurance that the licensing basis of 

all currently operating plants are sufficient to assure that operation is not 

inimical to the public health and safety. 

4-6 
5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 

5.1 Integrity of Reactor Coolant Pressure Boundary and Reactor Vessels 

5.1.1 Scope 

This section addresses an evaluation of the reactor coolant system and systems connected to it. Special consideration is given to the reactor coolant system and pressure-containing parts out to and including isolation valving which is 

the reactor coolant pressure boundary (RCPB), as defined in 10 CFR 50.2(v). The 

evaluation covers the areas of integrity of RCPB, reactor vessels, and component and subsystem design. 

5.1.2 Safety Issues and Regulatory Requirements 

Commission regulations require that the RCPB have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture under 

operating maintenance, testing, and postulated accident conditions and require that the reactor coolant system and associated auxiliary, control, and protec- tion systems be designed with sufficient margin to ensure that the design conditions of the RCPB are not exceeded during any conditions of normal operation, including anticipated operational occurrences. These regulations also require that components that are part of the RCPB be designed, fabricated, erected, and 

tested to the highest quality standards practical and also that they be designed 

to permit periodic inspection and testing of critical areas to assess their 

structural and leaktight integrity. 

To satisfy these requirements, RCPB components must be designed, fabricated, 

erected, and tested to 10 CFR 50.55a(c), inservice inspection must be performed according to 10 CFR 50.55a(g), and RCPB components must meet the fracture toughness and material surveillance requirements of 10 CFR 50.60. Additional fracture toughness requirements for protection against pressurized thermal shock 

events are contained in 10 CFR 50.61. Also protection against overpressure is 

provided per the requirements of American Society of Mechanical Engineers Boiler 

and Pressure Vessel Code (ASME Code), Section III, Article NB-7000. 

10 CFR 50.55a(c) requires RCPB components to meet the edition and addenda of 

Section III of the ASME Code that was required by Commission regulations at the 

time the regulations were issued. 10 CFR 50.55a(g) requires RCPB components to 

meet Section XI of the ASME Code. Proposed alternatives to these ASME Code 

requirements are permitted in 10 CFR 50.55(a)(3), provided the licensee demonstrates that (1) the proposed alternative would provide an acceptable level of 

quality and safety or (2) compliance with the specified requirements would 

result in hardship or unusual difficulties without a compensating increase in 

the level of quality and safety. When inservice inspection requirements of Section XI of the ASME Code are determined to be impractical, the NRC, in accordance with 10 CFR 50.55a(g)(6), may grant relief and may impose alternative 

requirements that are determined to be authorized by law, will not endanger life 

or property or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could 

result if the requirements were imposed on the facility. 

5-1 
10 CFR 50.GO(a) requires that RCPB components meet the fracture toughness requirements in Appendix G to 10 CFR Part 50, and requires the reactor vessel 

material surveillance program to meet the requirements in Appendix H to 10 CFR 

Part 50. These appendices impose additional requirements on the reactor vessel 

because the reactor vessel is subject to neutron irradiation embrittlement. 

10 CFR 50.GO(b) permits licensees to meet alternative requirements to those 

specified in Appendices G and H when an exemption is granted by the Commission 

under 10 CFR 50.12. A low-temperature overpressure protection (LTOP) system is 

provided to ensure that the pressure-temperature limits per the Appendix G 

requirements are not exceeded. 

5.1.3 Evolution of Current Licensing Basis 

The RCPB is composed of piping, pumps, valves, and vessels. Before 1970, p1p1ng in the RCPB was constructed and fabricated to the American National Standards 

Institute (ANSI) Code B31.1, 11 Power Piping, 11 and pumps and valves were constructed and fabricated to manufacturer specifications. In 1970, the ASME Code 

was revised to include requirements for RCPB piping, pumps, and valves in Section III. Vessels within the RCPB are constructed and fabricated to ASME Code 

requirements. Earlier plants were constructed to Sections I and VIII, and later 

plants were constructed to Section III requirements. 

The fabrication requirements for RCPB piping, pumps, valves, and vessels are 

specified in the plant 1 s final safety analysis report (FSAR). The staff reviewed 

the plant 1 s FSAR to determine that the alternative requirements to Section III 

of the ASME Code provided an acceptable level of quality and safety. 

Sections III and XI of the ASME Code have changed and will keep changing during the licensed lifetime of nuclear power plants including the renewal period. The 

staff reviews these changes to determine whether they should be incorporated into the individual plant licensing bases. 

The most significant change in Section III of the ASME Code, which affects RCPB 

integrity, was a change in fracture toughness requirements initiated in the 

Summer 1972 Addenda to the 1971 Edition of the ASME Code. This change required additional material testing and required pressure/temperature (P/T) limits during heatup, cooldown, and hydrotest of the reactor vessel. All licensees are 

required to heat up, cool down, and hydrotest the reactor vessel in accordance 

with plant-specific P/T limits that are based on linear elastic fracture mechanics technology. However, the Commission decided that plants built before 1972 

and plants that had ordered their reactor coolant pressure boundary material 

before 1972 did not have to perform the additional testing required by the 

Summer 1972 Addenda. Except for reactor vessel materials, the Commission concluded that the earlier test requirements were adequate to ensure RCPB integrity. Commission acceptance of these requirements is documented in Appendix G to 

10 CFR Part 50. For reactor vessel materials, the staff issued Branch Technical 

Position (MTEB) 5-2, 11 Fracture Toughness Requirement. 11 This branch technical 

position described a method of updating the earlier test data to Summer 1972 

Addenda requirements. The test data must be updated in order to calculate P/T 

limits. Licensees utilized this method or developed their own method to update their reactor vessel material test data. Methods developed by licensees were 

reviewed and approved by the staff. 

5-2 
10 CFR 50.55a(g) requires that RCPB components meet Section XI of the ASME Code. 

Section XI of the ASME Code contains updating provisions. It requires licensees 

to revise their inservice inspection programs every 10 years. The revised programs incorporate all the changes required by Section XI of the ASME Code, 

except for those that are impractical. The staff reviews the lic~nsee 1 s program and, in accordance with 10 CFR 50.55a(g)(6), the staff may grant relief or 

impose alternative requirements. 

Changes in technologies (i.e., radiation embrittlement, ultrasonic examination, 

etc.), which are not addressed by the ASME Code, are described in generic letters 

and regulatory guides. These letters and guides are prepared by the staff and 

may be incorporated into the licensing basis of the nuclear power plant. Many of these guides recommend changes to the plant licensing basis when environmental 

conditions change. These licensing changes are reviewed and approved by the 

staff. 

In addition to these requirements. the staff may recommend or require additional 

programs when it determines that the operating environment for the component is 

particularly severe. These programs are imposed through issuance of technical 

specifications or are recommended through issuance of branch technical positions, regulatory guides, standard review plans, or generic letters. Examples of components that operate in a particularly severe environment and for which the staff 

has either recommended or imposed additional requirements are boiling-water 

reactor (BWR) coolant pressure boundary piping, pressurized-water reactor (PWR) 

steam generator tubing, and all light-water reactor (LWR) vessels. Generic Letter 88-01 specified additional recommendations for BWR coolant pressure boundary piping because the piping was subject to intergranular stress corrosion cracking (IGSCC). Generic Letter 88-11 recommended a revised method of calculating neutron irradiation embrittlement of LWR reactor vessels because analysis of surveillance data in Appendix H to 10 CFR Part 50 indicated that the previous method did not adequately address the issue. The staff imposes augmented inspection program requirements on PWR steam generator tubing by issuing technical specifications. 

The staff reviewed the capability to remove contaminants from the reactor 

coolant system. The reactor water cleanup systems in direct-cycle BWR plants, in conjunction with the primary water monitoring system, must have the capability 

to remove contaminants introduced by main condenser leakage. Since the SEP 

effort, the staff has encouraged the preparation and implementation of Electric 

Power Research Institute (EPRI) and BWR owners group normal and hydrogen water 

chemistry guidelines. These guidelines, implemented by all BWR owners, specifi- cally address action recommendations in the event of condenser in-leakage transient implementation of the water chemistry guidelines through the use of NRC 

Inspection Procedure 79501. On the basis of such regulatory activities, the 

staff has concluded that the SEP lesson-learned issue has been acceptably resolved and that no additional regulatory action is necessary to address this 

issue. 

A reliable and sensitive leakage detection system is important to monitor the 

reactor coolant pressure boundary leakage to the containment and interconnecting 

system and it provides operators with an adequate margin of time to initiate 

actions to identify, isolate, and repair the source of a leak. Revisions to 

procedures or technical specifications were made for some plants which reflected 

system limitations and to enhance system reliability. Generic Letters 84-04 and 

5-3 
88-01 addressed improved technical specifications for leakage detection on PWRs 

and BWRs, respectively. All plants have technical specifications that address 

allowable rates for leaks in the primary coolant system. Additionally, proce- dures and operator training improvements also address the detection of primary coolant system leakage and necessary actions. 

5.1.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the reactor coolant pressure boundary as defined in 10 CFR 

50.2(n) met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will continue to obtain new information related 

to this subject area through a variety of sources such as updates to the FSAR, 

research reports, operating plant events and routine plant inspections. The 

staff reviews this information and, in the past, has required licensees to take 

actions to upgrade the plant to provide continuing assurance of adequate pro- tection of the public health and safety. In addition, the staff will continue 

to review new information in this subject area and if the staff determines that 

new or different requirements are needed, the staff has the capability within 

the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the 

public. In conclusion, the Commission concludes that the current regulatory 

process has and will continue to provide reasonable assurance that the licensing basis of all currently operating plants are sufficient to assure that 

operation is not inimical to the public health and safety. 

5.2 Component and Subsystem Design 

5.2.1 Scope 

This section addresses the performance requirements and design features to 

ensure overall safety of the various components within the reactor coolant 

system and subsystems closely allied with the reactor coolant system. This 

component and subsystem include reactor coolant pumps, steam generators, reactor 

coolant piping, main steamline flow restrictions, main steamline isolation system, reactor core isolation cooling system, residual heat removal system, reactor 

water cleanup system: main steamline and feedwater piping, pressurizer, pres- surizer relief discharge system, valves, safety and relief valves, component 

supports, and reactor coolant system high point vents. 

5.2.2 Safety Issues and Regulatory Requirements 

Commission regulations require that systems, structures, and components important 

to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed and that 

systems, structures, and components important to safety shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whip- ping, and discharging fluids, that may result from equipment failures and from 

events and conditions outside the plant. In addition, the regulations require (1) that the RCPB shall be designed, fabricated, erected, and tested so as to 

have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture; (2) that the reactor coolant system and associated auxiliary systems shall be designed with sufficient margin to ensure that 

5-4 
the design conditions of the RCPB are not exceeded during any condition of normal operation, including anticipated operational occurrences; (3) that the RCPB 

shall be designed with sufficient margin to ensure that, when stressed under 

operating, maintenance, testing, and postulated accident conditions, the boundary behaves in a nonbrittle manner and the probability of rapidly propagating 

fracture is minimized; (4) that components that are part of the RCPB shall be 

designed to permit periodic inspection and testing of important areas and features to assess their structural and leaktight integrity; and (5) that a system 

to supply reactor coolant makeup for protection against small breaks in the RCPB 

shall be provided. 

The system safety function shall be to transfer fission product decay heat and 

other residual heat from the reactor core at such a rate that specified acceptable fuel design limits (SAFDLs) and the design conditions of the RCPB are not 

exceeded. Suitable redundancy in components and features and suitable 

interconnections, leak detection, and isolation capabilities shall be provided. 

5.2.3 Evolution of Current Licensing Basis 

The licensing basis for various components within the reactor coolant system and 

subsystems has evolved as reactor events and generic studies by the NRC staff 

provide new information that is determined to improve component and subsystem 

performance. The process of evaluating operating experience and assessing plant 

data to determine the need for additional actions is a continuous one. 

From operating experiences, many different forms of steam generator tube degradation have been identified, including stress corrosion cracking, wastage, intergranular attack, denting, erosion-corrosion, fatigue cracking, pitting, fretting, 

support plate degradation, and mechanical damage resulting from impingement of 

foreign objects or loose parts on the internal components of steam generators. 

These degradations have resulted in extensive steam generator inspections, tube 

plugging, repair, or replacement. Also steam generator tube rupture (SGTR) 

events have occurred in a few operating reactors. Steam generator tube integrity was designated an unresolved safety issue (US!) in 1978 and Task Action 

Plans (TAPs) A-3, A-4, and A-5 were established to evaluate the safety significance of degradation in steam generators of various designs. NUREG-0844 was 

published in September 1988 to present the results of the NRC integrated program 

for the resolution of USis A-3, A-4, and A-5 regarding steam generator tube 

integrity. A generic risk assessment is provided and indicates that risk from 

SGTR events is not a significant contributor to total risk at a given site, nor 

to the total risk to which the general public is routinely exposed. This finding is considered to be indicative of the effectiveness of licensee programs and 

regulatory requirements for ensuring steam generator tube integrity in 

accordance with Appendices A and B to 10 CFR Part 50. This report also identifies a number of staff-recommended actions that can further improve the effectiveness of licensee programs in ensuring the integrity of steam generator tubes 

and in mitigating the consequences of an SGTR. As part of the integrated program, the staff issued Generic Letter 85-02 encouraging licensees of PWRs to 

upgrade their programs, as necessary, to meet the intent of the staff-recommended 

actions; however, such actions do not constitute NRC requirements. In addition, 

the staff is pursuing a number of actions and studies involving steam generator 

issues to gain added assurance that risk from SGTR events will continue to be 

small. 

5-5 
Following the accident at Three Mile Island, Unit 2 (TMI-2), the staff found 

that additional means were necessary to vent noncondensible gases from the reactor coolant system which may inhibit core cooling during natural circulation. 

On the basis of knowledge gained from the TMI-2 accident, Item II.8.1, 11 Reactor 

Coolant System Vents, 11 was incorporated into the licensing bases for individual 

plants when all operating nuclear power plants were required to implement reactor 

coolant system high point venting capability in accordance with these guidelines. 

Also from the experience of the TMI-2 accident, the staff found that operational performance of the relief and safety valves under various operating conditions 

is significant to safety. Performance testing of BWR and PWR relief and safety 

valves was incorporated in individual plant licensing bases when all nuclear 

power plants were required to implement testing requirements in accordance with 

the guidelines contained in TMI Action Plan Item II.D.l. 

5.2.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the components within the reactor coolant system and subsystems 

met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will continue to obtain new information related to this subject area through a variety of sources such as updates to the FSAR, research 

reports, operating plant events and routine plant inspections. The staff reviews this information and, in the past, has required licensees to take actions 

to upgrade the plant to provide continuing assurance of adequate protection of 

the public health and safety. New criteria and improvements necessary for 

safety which have resulted from the continuous staff review have included steam 

generator tube intergrity (GL 85-02), reactor coolant system vents (NUREG-0737, 

Item II.8.1), and performance testing of relief and safety valves (NUREG-0737, 

Item II.0.1). In addition, the staff will continue to review new information in 

this subject area and if the staff determines that new or different requirements are needed, the staff has the capability within the existing regulatory 

process to require additional analyses or plant modifications, as necessary, to 

ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has and will continue to 

provide reasonable assurance that the licensing basis of all currently opera- ting plants are sufficient to assure that operation is not inimical to the 

public health and safety. 

5-6 
6 ENGINEERED SAFETY FEATURES 

6.1 Scope 

Engineered safety features (ESFs) are provided to mitigate the consequences of 

postulated accidents in spite of the fact that these accidents are very unlikely. The engineered safety features included in plant designs vary depending on 

the type of plant (PWR or BWR) under evaluation. This section will discuss 

five general categories of features routinely considered under the subject of 

ESFs. These include: metallic and organic materials, containment systems, 

emergency core cooling systems, habitability systems, and fission product removal and control systems. 

6.2 Metallic and Organic Materials 

6.2.1 Safety Issues and Regulatory Requirements 

The Commission regulations require that the containment boundary be designed with sufficient margin to ensure that, under operating, maintenance, testing, and postulated accident conditions, its ferritic materials behave in a 

nonbrittle manner and the probability of a rapidly propagating fracture is minimized. In addition, the regulations require that systems, structures, and 

components important to safety be designed, fabricated, erected, and tested to 

quality standards commensurate with the importance of the safety function per- formed. Specific guidance on satisfying these requirements is contained in 

applicable regulatory guides that refer to the criteria of ASME Section III for 

metallic materials used in ESF system construction. In specific cases, with 

proper justification, the staff evaluated and found acceptable alternatives to 

these criteria that continue to ensure ESF system integrity and performance. 

In addition, 10 CFR 50.55a(g) requires that essential components in ESF systems built to ASME Section III criteria receive regularly scheduled inservice inspection in accordance with the criteria of ASME Section XI. Relief can be 

granted against the criteria of ASME Section XI when the NRC staff determines 

that alternative measures are in place to ensure fracture prevention of the 

pressure boundary. 

6.2.2 Evolution of Current Licensing Basis 

Sections III and XI of the ASME Code have changed and will continue to change during the plant lifetime of nuclear power plants based on operating experience. The staff reviews these changes to determine whether they should be incorporated into the licensing basis of operating plants. 

Section XI of the ASME Code contains updating provisions. It requires licensees to revise their inservice inspection program every 10 years. The revised 

programs incorporate all changes required by Section XI of the licensee 1 s 

program and, in accordance with 10 CFR 50.55a(g)(6), the staff may grant relief 

or impose alternative requirements. 

6-1 
The staff reviewed the use of organic materials inside containment. The 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 containment during accidents when they may be exposed to high 

temperatures, steam environments, high radiation fields, and containment spray 

systems. Since the completion of the SEP effort, the staff has implemented 

some actions related to this topic: Regulatory Guide 1.54, which ensorsed 

industry standards ANSI NlOl.4 (1972), 11Qual ity Assurance for Protective Coatings Applied to Nuclear Facilities/ and ANSI NlOl.2 (1927), 11 Quality Assurance 

Program Requirements for Nuclear Power Plants. 11 In addition the NRC issued 

NUREG-0897, 11 Containment Emergency Sump Performance, 11 and NUREG/CR-2791, 11Methodology for Evaluation of Insulation Debris Effects, 11 to provide a basis for 

making an evaluation of sump performance and insulation debris effects. 

Based upon the above and industry actions, the staff concludes that adequate guidance and information has been issued and that licensee actions are adequate 

to address this issue. 

6.2.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that ESF system components met the staff acceptance criteria and the 

intent of all applicable regulations. The staff has and will continue to obtain 

new information related to this subject area through a variety of sources such 

as updates to the FSAR, research reports, operating plant events and routine 

plant inspections. The staff reviews this information and, in the past, has 

required licensees to take actions to upgrade the plant to provide continuing 

assurance of adequate protection of the public health and safety. In addition, 

the staff will continue to review new information in this subject area and if 

the staff determines that new or different requirements are needed, the staff 

has the capability within the existing regulatory process to require additional 

analyses or plant modifications, as necessary, to ensure the continued health 

and safety of the publ'ic. In conclusion, the Commission concludes that the 

current regulatory process has and will continue to provide reasonable assurance 

that the licensing bases of all currently operating plants are sufficient to 

assure that operation is not inimical to the public health and safety. 

6.3 Containment Systems 

6.3.1 Safety Issues and Regulatory Requirements 

The Commission regulations require that nuclear power plants be provided with 

an essentially leaktight containment as a barrier against uncontrolled release 

of radioactivity to the environment following accidents. More specifically, the regulations require that containment heat removal systems be designed, inspected, and tested in a manner intended to ensure their safety function and 

that containment atmosphere cleanup systems be designed, inspected, and tested 

6-2 
in a manner intended to ensure their safety function. In addition, the regula- tions require that the containment be designed to (1) withstand post-accident 

temperature and pressure conditions without exceeding the design leak rate, 

(2) prevent fracture, (3) permit periodic integrated leakage testing, (4) permit periodic inspection and pressure testing of resilient seals, and (5) provide appropriate isolation valves. 

6.3.2 Evolution of Current Licensing Basis 

In order to demonstrate that containment designs are capable of withstanding post-accident temperature and pressure conditions without releasing excessive 

radioactivity, licensees and the staff have used mathematical models to establish and confirm acceptable containment performance. These models and the input assumptions are conservative and have demonstrated that containments are 

designed with substantial margin. As new information and research on containment design and post-accident energy release are obtained, such information is 

applied to the analytical methods as appropriate to ensure that adequate margins against excessive leakage are maintained. 

For example, in the early 1970s, General Electric identified concerns regarding post-accident pool dynamic loads on BWR pressure suppression containments. The 

staff and BWR licensees performed significant reanalyses of containment performance based on this newly identified load phenomenon. The result of this effort was the formation of programs for modifications to the Mark I, II, and III 

BWR containment designs in order to reestablish the original containment design margins. 

In the early 1980s, Westinghouse informed the staff that steam line break analyses may not have properly considered superheated steam blowdown conditions 

into the containment which could occur as the steam generator drys out. This 

information led to revised steam line break analyses by licensees which incorporated the new blowdown input. The new analyses confirmed that containment 

performance remains acceptable and appropriate margins are maintained. 

Following the accident at Three Mile Island Unit 2 in March 1979, the NRC staff 

noted several concerns with regard to containment performance during the event 

that warranted improvement. One area of major focus concerned the capability 

to control combustible gas following accidents. Initially, the NRC staff required licensees to provide dedicated hydrogen penetrations to ensure the ability to employ hydrogen recombiners to reduce post-accident hydrogen concentration in the containment. This improvement was implemented as Item 11.E.4.l of 

the TMI Action Plan Clarification, NUREG-0737. However, the NRC staff also 

recognized that further research into combustible gas concerns was necessary. This led to substantial modifications to 10 CFR 50.44 in 1981 and 1985 wherein 

more stringent combustible gas control measures were specified for pressure suppression containment plants. Implementation of these requirements has 

improved combustible gas control capability. 

The TMI-2 accident also pointed out the need for improvements in containment 

isolation dependability. New criteria in this regard were implemented as part of TMI-2 Action Plan, Item 11.E.4.2, which required all licensees to evaluate 

their post-accident containment isolation capability against current criteria 

and make the necessary changes to improve its dependability. 

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Over the past few years, the NRC staff has undertaken research into severe accident effects on containments for all types of operating plants. To date, 

this program has pointed out weaknesses in the capability of BWR plants with 

Mark I pressure suppression containment designs to ensure adequate containment 

integrity under severe accident conditions. This has resulted in issuance of 

Generic Letter 89-16, which indicated the staff's intention to pursue plant- specific backfit procedures for a wetwell vent on all Mark I plants if the 

licensee does not voluntarily install the vent. Implementation of this improvement is currently proceeding. 

Based on the continuous review of Appendix J leak rate test results, the NRC 

staff has periodically updated this rule to incorporate improved containment 

leak rate testing guidelines. One recent change was to permit use of the mass 

point method when conducting a Type A integrated leak rate test. Other revisions to Appendix J are currently pending and will provide further improvement in leak rate testing. 

In addition, inservice inspection requirements for the containment structures 

and components are identified in Appendix J, 10 CFR Part 50. Section V.A in 

Appendix J requires a general inservice inspection of the accessible interior 

and exterior surfaces of the containment structures and components prior to any Type A test to uncover any evidence of structural deterioration that may affect 

either the containment structural integrity or leaktightness. 

As part of the SEP effort the staff reviewed the isolations capability of lines 

that penetrate containment. Isolation provisions for lines that penetrate the 

primary containment maintain an essential leaklight barrier against the uncontrolled release of primary system coolant as result of postulated pipe breaks 

outside containment. The isolation function must be accomplished without endangering the performance of post-accident safety systems. Since the effort, 

the staff has implemented a number of regulatory initiatives that address the 

topic. TMI Action Plant item II.E.4.2 addressed actions relating to containment 

isolation system dependability, such as isolation of non-essential systems, diverse signals, and the reset of the containment isolation system (CIS) actuation. The TMI Action Plan also addressed procedural improvements. Type C testing required by Appendix J and by plant-specific TS periodically verify accept- able leak rates from containment. The NRC inspection program examines the valve 

arrangements and administrative controls on manual valves. Although not required 

to address this SEP issue, the IPE program will also include a review to identify specific vulnerabilities. 

6.3.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the containment design met the staff acceptance criteria and the 

intent of all applicable regulations. The staff has and will continue to obtain 

new information related to this subject area through a variety of sources such 

as 4pdates to the FSAR, research reports, operating plant events and routine 

plant inspections. The staff reviews this information and, in the past, has 

required licensees to take actions to upgrade the plant to provide continuing 

assurance of adequate protection of the public health and safety. In addition, 

the staff will continue to review new information in this subject area and if 

6-4 
the staff determines that new or different requirements are needed, the staff 

has the capability within the existing regulatory process to require additional 

analyses or plant modifications, as necessary, to ensure the continued health 

and safety of the public. In conclusion, the Commission concludes that the 

current regulatory process has and will continue to provide reasonable assurance 

that the licensing bases of all currently operating plants are sufficient to 

assure that operation is not inimical to the public health and safety. 

6.4 Emergency Core Cooling Systems (ECCS) 

6.4.1 Safety Issues and Regulatory Requirements 

The Commission regulations require that nuclear power plants contain abundant 

emergency core cooling capability and specifies the specific safety functions 

for these systems. 10 CFR Part 50.46 and Appendix K to 10 CFR Part 50 establish the criteria and evaluation methods to be used by licensees and vendors to 

evaluate ECCS designs. The ECCS cooling performance must be evaluated using an 

acceptable model and must be evaluated for a number of postulated loss-of-coolant 

accidents of different sizes, locations, and other properties to ensure that the 

range of postulated loss-of-coolant accidents are considered. 

6.4.2 Evolution of Current Licensing Basis 

In June 1971, prior to establishing Part 50.46 or Appendix K, the Commission 

published interim acceptance criteria for ECCS designs by Westinghouse and General Electric reactor plants, and concluded that these criteria provide areasonable assurance that ECCS will be effective in the unlikely event of a 

loss-of-coolant accident. However, research was under way at the time, and 

increased knowledge of heat transfer, fluid flow, and engineering disciplines important to ECCS analysis was anticipated. 

Based on this research, modifications were made to the ECCS analysis guidelines. 

In December 1971, the NRC amended the interim criteria to add evaluation models 

for reactor designs by Babcock and Wilcox and Combustion Engineering. In 

January 1972, the AEC undertook an extensive rulemaking hearing. As a result 

of this proceeding, the Commission established a new Part 50.46 and Appendix K 

in January 1974, setting forth the acceptance criteria and the ECCS evaluation 

models in a final rulemaking. These regulations, which were enacted only after 

extensive rulemaking hearings, established the general approach that remains in 

use today. Between 1974 and 1976, extensive efforts were made to apply the 

requirements and criteria of Part 50.46 and Appendix K to all light water reactors then in operation. All plants subsequently licensed have been found to 

meet Part 50.46 and Appendix K. 

In 1987, the Commission proposed modifications to the regulations because research, performed since the current rule was written, has shown that calculations performed using current methods and in accordance with the current 

requirements result in estimates of cooling system performance that are signif- icantly more conservative than estimates based on the improved knowledge gained 

from this research. 

The final rule incorporating new evaluation models was published in September 1988, but did not force facilities that had used previous models to perform new 

6-5 
analyses. The Commission concluded at that time that existing Appendix K evaluation models should be permitted indefinitely. The Commission also believes 

that the decision to permit continued use of such models can and should be made 

at this time because it believes that both methods provide adequate protection of the public health and safety. 

During the process of licensing, each applicant must submit in the FSAR sufficient information to describe the design bases for each ECCS subsystem, including its functional requirements, reliability requirements, protection from 

physical damage, and environmental conditions. Significant design parameters such as design flow rates, system temperatures, etc., along with piping and 

instrumentation diagrams, are routinely included. 

Prior to granting an operating license, the staff reviews the described ECCS 

design against established acceptance criteria and concludes, in general, that 

the plant-specific design of the ECCS meets all necessary requirements and is 

acceptable. 

The review process does not stop. The performance requirements of the ECCS are 

routinely evaluated during each plant refueling to ensure that operation during the subsequent cycle will be within the safety envelope of the plant design. 

In many instances, technical specification changes or license conditions are 

implemented to govern operation during the period of operation. In the extreme 

case, plant modification may be required to provide continued assurance of pub- lic health and safety. 

Plant operating events also generate new information that may require operating plants to reanalyze the performance of the ECCS and, as necessary, make plant modifications. One such example was the lessons learned from the accident at 

TMI-2. Following this event, all operating reactors were required to reanalyze their plant-specific response to a range of small-break LOCAs. In some cases, 

these reanalyses resulted in plant modifications or changes in operating proce- dures being made. Another result of the TMI event was the requirement to install reactor head vents and to have operating procedures that describe how to 

use these vents in the event of certain postulated accidents. The net result 

was an overall improvement in the level of safety provided by the ECCS at operating nuclear power plants. 

6.4.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the ECCS designs met the staff acceptance criteria and the intent 

of all applicable regulations. The staff has and will continue to obtain new 

information related to this subject area through a variety of sources such as 

updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance 

of adequate protection of the public health and safety. In addition, the staff 

will continue to review new information in this subject area and if the staff 

determines that new or different requirements are needed, the staff has the 

capability within the existing regulatory process to require additional analyses 

or plant modifications, as necessary, to ensure the continued health and safety 

6-6 
of the public. In conclusion, the Commission concludes that the current regulatory process has and will continue to provide reasonable assurance that the 

licensing bases of all currently operating plants are sufficient to assure that 

operation is not inimical to the public health and safety. 

6.5 Habitability Systems 

6.5.1 Safety Issues and Regulatory Requirements 

The Commission regulations require that control rooms at nuclear power plants 

be provided with adequate radiation protection to permit access and occupancy under accident conditions such that personnel do not receive radiation exposures in excess of 5-rem whole body or its equivalent to any part of the body for the duration of the accident. Additional guidelines are contained in regu- latory guides for assuring operator protection against both radioactivity and 

toxic gas (e.g., chlorine) releases following postulated accidents. 

6.5.2 Evolution of Current Licensing Basis 

To satisfy the Commission requirements, all licensees have performed dose analyses using mathematical models to ensure that post-accident radiation levels 

within the control room are within the required limits. Guidelines for conducting these analyses have remained essentially unchanged since the mid-1970s. 

The assumptions used are considered to be conservative in order to account for 

uncertainties in the actual radioactivity release mechanism following an 

accident. 

The accident at Three Mile Island Unit 2 in March 1979 pointed out potential vulnerabilities in the capability of control room habitability systems, i.e., 

the control room ventilation system to ensure adequate radiation protection for 

the operators. Therefore, Item III.D.3.4 of the TMI Action Plan Clarification, 

NUREG-0737, was implemented at all operating plants. This item required licensees to evaluate their control room habitability systems against the criteria of Standard Review Plan Section 6.4 and perform the necessary analyses of 

toxic gas and radiation exposure to the operators in order to demonstrate compliance with these criteria. All plants provided responses to this issue and 

made improvements in the control room ventilation systems, as appropriate. 

The staff also recognized the need to conduct a longer term review of criteria 

for ensuring control room operator protection and began a study in this regard 

under Generic Issue 83 in the mid-1980s. This effort began with a survey of 12 

nuclear power plants to determine what improvements had been made as part of 

the NUREG-0737, Item III.D.3.4, implementation. Based on the results of the 

survey, the staff determined that further guidance to improve control room habitability systems was necessary. This guidance is currently under development and is intended to be issued in a generic letter to all licensees soon. 

6.5.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the control room habitability system met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will 

continue to obtain new information related to this subject area through a variety 

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of sources such as updates to the FSAR, research reports, operating plant events 

and routine plant inspections. The staff reviews this information and, in the 

past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In 

addition, the staff will continue to review new information in this subject area 

and if the staff determines that new or different requirements are needed, the 

staff has the capability within the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued 

health and safety of the public. In conclusion, the Commission concludes that 

the current regulatory process has and will continue to provide reasonable 

assurance that the licensing bases of all currently operating plants are sufficient to assure that operation is not inimical to the public health and safety. 

6.6 Fission Product Removal and Control Systems 

6.6.1 Safety Issues and Regulatory Requirements 

The Commission regulations require that containment atmosphere cleanup systems be designed, inspected, and tested in a manner to ensure their safety function 

following postulated accidents. 

6.6.2 Evolution of Current Licensing Basis 

Staff guidance in the fission product removal and control area has changed little over the years. Most nuclear power plants are equipped with ventilation 

systems containing charcoal and high efficiency particulate air filters for 

fission product removal and prevention of unacceptable radiological releases 

during normal operation and post-accident conditions. Plants with filters have 

technical specifications that require surveillance and testing of those filters 

to ensure their continued satisfactory performance. PWR plants are also 

equipped with containment spray systems that provide both a post-accident heat 

removal and fission product control safety function in containment. As a means 

of controlling pH in the spray water, these plants have utilized a sodium 

hydroxide solution as a spray additive. Over the years, however, the staff 

recognized, through research of the post-accident source term, that a lower 

spray water pH (no lower than 7) was acceptable to ensure iodine retention and 

long-term corrosion control in ECCS systems. As a result, some PWR licensees 

have removed the sodium hydroxide addition system and replaced it with much 

simpler trisodium phosphate baskets placed directly in the containment sump in 

order to achieve necessary pH control. 

In BWRs, blowdown of the reactor through the suppression pool results in some 

fission product removal following an accident. However, the staff had not pre- viously credited this pathway in dose analyses. Staff review of recent analyses by General Electric resulted in a recognition of the suppression pool as a 

means of fission product control and led to a revision of the Standard Review 

Plan to credit an appropriate decontamination factor. Future BWRs will utilize 

this additional credit in post-accident dose analyses as may currently operating pJ ants when proposing changes. 

6.6.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

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conclude that the fission product removal and control system met the staff 

acceptance criteria and the intent of all applicable regulations. The staff 

has and will continue to obtain new information related to this subject area 

through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff ~evi~ws this. 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information in this subject area and if the staff determines that new or different 

requirements are needed, the staff has the capability within the existing reg- ulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, 

the Commission concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing bases of all currently operating plants are sufficient to assure that operation is not inimical to the 

public health and safety. 

6-9 

7 INSTRUMENTATION AND CONTROL SYSTEMS 

7.1 Scope 

The licensing bases and regulatory requirements for instrumentation and control 

(I&C) systems are discussed in the following sections. 

The systems to be discussed in this section include the reactor trip system, engineered safety features actuation system, safe shutdown systems, and safety- related display systems. Remote shutdown systems are included in safe shutdown 

systems and post-accident monitoring and safety parameter display systems are 

included in safety-related display systems. 

7.2 Development of Regulatory Requirements 

Initially the regulatory requirements came from the need to develop highly reliable instrumentation and control systems to monitor and control the operation of nuclear reactors and other critical systems. In response to this need, concepts and methods such as the single failure criterion, failure mode and effects 

analysis, reliability, failure rates, sneak circuit analysis, redundancy, and 

diversity were developed and applied. In August 1968, these concepts and methods 

were originally collected into proposed IEEE Standard 279, "Criteria for Protection Systems for Nuclear Power Generating Stations," which was incorporated into 

10 CFR 50.55a(h) in 1970. In addition, these concepts and methods were made 

part of the Commission regulations governing the design, fabrication, construction, installation, testing, and operation of these highly reliable instrumentation and control systems for nuclear reactors. 

In 1974 and 1975, the staff went further in providing guidance by drafting and 

issuing criteria by which they would review the safety analysis reports (SARs) 

and other information submitted by licensees and applicants. The totality of 

these requirements has become the regulatory requirements that the licensees 

must address for their plant. This body of requirements is frequently revised 

and upgraded to take ·into account technological advances and lessons learned 

from operating experience. 

The licensee, however, is authorized through 10 CFR 50.59 to make changes to 

the plant and its procedures and to conduct tests or experiments not described 

in the SAR without prior NRC approval unless the proposed change, test, or 

experiment involves changes to the technical specifications or introduces an 

unreviewed safety question. This body of requirements as it exists at the time 

application is made for an operating license and as reviewed and approved by the 

staff becomes the specific regulatory requirements for that plant. 

7.3 Reactor Trip System 

7.3.1 Safety Issues and Regulatory Requirements 

The primary safety function of the reactor trip system (RTS) instrumentation is 

to monitor selected reactor and plant parameters related to nuclear power 

7-1 
generation and transfer of the heat from that generation to the power conversion 

devices. When these parameters approach and exceed values deemed unsafe by analysis, the system shall initiate reactor shutdowns that shall promptly make 

the reactor core subcritical, i.e., stop the generation of nuclear power, by rapidly inserting control rods into the core or by other means of rapidly inserting enough negative reactivity into the core to make it subcritical and 

to keep the core subcritical. 

The RTS instrumentation must be highly reliable, minimize false shutdowns, 

possess high availability, be automatically initiated, provide for manual 

initiation, and be designed so that the operators can easily and quickly determine the state of the plant. Applicable design requirements ensure that 

trip parameter monitoring channels and trip logic and actuation trains are 

redundant and independent; that all channels and trains meet the single failure 

criterion; that monitored parameters are sufficiently diverse; and that 

measuring instrumentation possesses adequate range, sensitivity, and accuracy and has adequate capability for test and calibration. These design requirements ensure that parts and components are specified that meet plant-specific seismic and environmental requirements in accordance with IEEE Standard 344 and 

10 CFR 50.49. Fabrication and installation requirements ensure that the system 

or subsystem is built of Class lE parts and components, that it is fabricated 

and installed to meet plant-specific seismic and environmental requirements, and that quality control and quality assurance programs and procedures are used 

that meet the requirements of 10 CFR Part 50, Appendix B, and applicable IEEE 

and ANSI standards. Testing and operational requirements set forth in 10 CFR 

50.36, 10 CFR Part 50 Appendices A and B, IEEE standards, and various regulatory guidance ensure that the system is adequately tested prior to and during operation and that the system is operated within the limits specified in the 

plant technical specifications. 

7.3.2 Evolution of Current Licensing Basis 

As plants became operational during the 1970s, operating experience indicated 

the need for improvements and changes in branch technical positions, technical 

specifications, regulatory guides, and IEEE standards. Changes at the nuclear 

power plants were also recommended in generic letters and bulletins. However, 

several major events occurred that caused major changes to be made to the 

licensing basis. 

The Brown's Ferry fire in 1975 taught lessons about separating and protecting safety-related instrumentation, control, and power cabling. It also emphasized the importance of providing remote initiation capabilities for safety-related equipment that could be made independent of cabling and equipment in the cable 

spreading and main control rooms. Revision of the IEEE standard and the regulatory guide on separation and independence as well as revisions to other IEEE 

standards relating to testing, qualification, and installation of safety-related equipment resulted from staff experience with this event. All plants licensed 

subsequently were reviewed by the staff to confirm that the protection system design precludes the use of components that are common to redundant channels, 

such as: actuation, reset, mode and test switches, common power supplies, or 

any other features that could compromise the independence of redundant channels. 

IEEE Std. 279 Sec. 4.6; IEEE Std. 384; Regulatory Guide 1.75; GDC-22; and SRPs 

7.2 and 7.3 were used as acceptance criteria for these reviews. 

7-2 
The TMI-2 event in 1979 mandated many changes, which included significant revisions to operating procedures, incorporation of human factors concepts into 

the design and arrangement of instrumentation and controls on main control 

boards, monitoring of reactor vessel water level for BWRs and for PWRs, and new 

instrumentation to indicate reactor coolant sub-cooling margin for PWRs. The 

changes were implemented through generic letters and confirmatory orders. 

The Salem ATWS (Anticipated Transients Without Scram) events in 1983 involved 

the only failure of a U. S. reactor to shut down on demand. The rapid intervention of the operators limited the consequences but the implications regarding shutdown reliability were significant and brought changes in operating procedures, reevaluation of on-line testing capability of the RTS, modifications to RTS 

breakers for B&W and Westinghouse plants, and changes to associated maintenance 

procedures. These improvements were requested by Generic Letter 83-28. Each 

licensee response was reviewed and approved by the staff. In 1984, the Commission issued 10 CFR 50.62, which added diverse and independent reactor trip systems to further improve reactor shutdown reliability and reduce the risk from 

potential occurrences of ATWS events. The NRC is presently reviewing and 

inspecting each plant to ensure that the systems have been installed properly. 

7.4 Engineered Safety Features Actuation Systems 

7.4.1 Scope 

In this section are the actuating systems for typical ESF systems such as 

containment and reactor vessel isolation, emergency core cooling, containment 

heat removal, auxiliary feedwater, diesel generators, and standby gas treatment. 

7.4.2 Safety Issues and Regulatory Requirements 

The primary safety function of the engineered safety features actuation system (ESFAS) is to sense the need for, select, and initiate systems that take action 

to terminate or control and contain the effects and consequences of design basis 

accidents and operational occurrences. 

As for the RTS instrumentation and logic, the ESFAS instrumentation, logic, and 

actuation equipment should als.o be highly reliable, minimize spurious actuations, 

possess high availability, be automatically initiated, provide for manual 

initiation of protective action from the control room, and be so designed that 

the operators can readily determine the status of the ESF systems and their 

actuating systems. 

7.4.3 Evolution of Current Licensing Basis 

The ESFAS and the RTS are very similar systems, the difference being principally in the systems controlled or actuated and the mission of those systems. The 

licensing basis for the two systems evolved in very much the same way. The same 

IEEE standards and the same regulatory guides apply to both systems. Some staff 

re~uirements apply to the ESFAS that do not apply to the RTS and vice versa; 

however, the basis for applying the requirements to the ESFAS and RTS subsystems is the same. Similarly, the modifications to the licensing bases for the ESFAS 

caused by the Brown's Ferry fire, TMI, Salem ATWS, the ATWS rule, and the feedback of operating experience are much the same for the ESFAS as they were for 

the RTS and are therefore not presented again. 

7-3 
7.5 Safety-Related Display Instrumentation 

7.5.1 Scope 

This section includes the post-accident monitoring instrumentation (PAM) and the 

safety parameter display system instrumentation with the normal safety-related display instrumentation. 

7.5.2 Safety Issues and Regulatory Requirements 

The primary safety function of the safety-related display instrumentation (SRO!) 

is to assist in meeting the Commission regulations by providing the capability 

to display the instantaneous values of the monitored plant operating parameters that provide the operators the information they need to form and update their 

assessment of the plant's operating status. · 

The primary safety function of the PAM is to provide the capability to monitor 

appropriate plant parameters during and after plant accidents and transients to 

assist the control room operators in preventing and mitigating the consequences of those events. The primary safety function of the safety parameter display 

system (SPDS) is to provide a concise display of critical plant variables to the 

control room operators to aid in rapidly and reliably determining the safety 

status of the plant. 

As with the RTS, PAM instrumentation must be highly reliable, possess high availability,-and be so designed that the operators can readily determine the status 

of the key variables. Applicable design requirements ensure that instrumentation 

channels are redundant and independent; that all channels meet the single failure 

criterion; that monitored parameters are sufficiently diverse; and that the measuring and indicating instrumentation possesses adequate range and sensitivity and has the capability for test and calibration. These design requirements also 

ensure that parts and components are specified that meet plant-specific seismic 

and environmental requirements in accordance with IEEE Standard 344 and 10 CFR 

50.49. Fabrication and installation requirements ensure that the instrumentation 

is built of Class lE parts and components, that it is fabricated and installed 

to meet plant-specific seismic and environmental requirements, and that quality control and quality assurance programs and procedures are used that meet the 

requirements of 10 CFR Part 50 Appendix Band applicable IEEE and ANSI standards. 

Testing and operational requirements set forth in 10 CFR 50.36, 10 CFR Part 50 

Appendices A and B, IEEE standards, and various regulatory guides ensure that 

the instrumentation is adequately tested and that the instrumentation is operated within the limits specified in the plant technical specifications. 

7.5.3 Evolution of Current Licensing Basis 

The TMI-2 event mandated many changes in safety analysis philosophy, operating procedures, incorporation of human factors concepts into the design and arrangement of instrumentation and controls on the main control boards, monitoring of 

reactor vessel water level for BWRs and for PWRs, and reactor coolant sub-cooling margin for PWRs. Following the TMI-2 event, the NRC staff developed a compre- hensive and integrated plan to improve safety at power reactors. As part of 

this plan, the Commission required the installation of improved post-accident monitoring instrumentation and SPDS. These improvements were intended to pro- vide the operator with a broader range of information for accidents, including those beyond the design basis. 

7-4 
The SPDS and Regulatory Guide 1.97, Revision 2, were items identified in the 

TMI Action Plan (NUREG-0737). Additional clarification for implementation of 

these items was addressed in NUREG-0737, Supplement No. 1 via Generic Letter 

82-33. The SPDS and the instrumentation in Regulatory Guide 1.97 were required for all operating plants, applicants for operating licenses, and holders of 

construction permits. The staff has reviewed almost all the submittals on 

conformance to Regulatory Guide 1.97 and SPDS. Generic Letter 89-06 was issued 

to all licensees for the purpose of certifying that the SPDS fully meets or 

will be modified to meet the requirements of NUREG-0737, Supplement 1. 

7.6 Safe Shutdown and All Other Systems Required for Safety 

7.6.1 Scope 

This section includes systems and interlocks required for safe shutdown and safe 

operation of the reactor which were not included as part of either the reactor 

trip system, engineered safety features actuation system, or the safety-related display instrumentation. Examples include, for PWRs: residual heat removal, 

auxiliary feedwater, boration, interlocks, and radiation monitoring systems and 

remote shutdown facilities; for BWRS: reactor core isolation cooling, residual 

heat removal (shutdown cooling mode), standby liquid control, neutron monitoring (including rod block monitor), recirculation pump trip, interlocks, and radiation monitoring systems, low level set instrumentation, and remote shutdown 

facilities. 

7.6.2 Safety Issues and Regulatory Requirements 

The safety functions of the systems in this section vary with the system or 

equipment but, for the majority of these systems, it is preventive for the 

interlocks, boration, and SLCS systems and protective for the shutdown cooling 

systems and radiation monitoring systems. 

In general, the instrumentation and logic systems in this section should meet 

the same safety criteria and regulatory requirements discussed in Section 7.4.1 

for the ESFAS instrumentation and logic systems; however, some systems and por- tions of systems, particularly radiation monitoring systems and portions of 

interlock systems, may not be required to meet all the requirements for Class lE 

systems. In addition; the Commission regulations require the provision for 

remote shutdown facilities that are located outside of the main control room 

and that meet the regulatory requirements. 

7.6.3 Evolution of Current Licensing Basis 

The licensing basis for these systems and equipment contains the same basic 

requirements as the RTS and ESFAS relating to system and equipment reliability; availability; redundancy; independence; ability to meet the single failure criterion; provision of adequate range, sensitivity, and accuracy in sensing and 

monitoring equipment; and provision of capability for test and calibration of 

the systems and equipment to which these requirements apply. 

Modifications to the licensing basis for the safe shutdown and all other systems required for safety that were found necessary by experience gained from the 

Brown's Ferry fire, TMI, Salem ATWS, the ATWS rule, and the feedback of oper- ating experience that updates it are the same for the requirements applicable 

7-5 
to these systems and equipment as they are for the RTS and ESFAS systems. These 

were previously discussed in Sections 7.2.2 and 7.3.2 and will not be discussed 

further. 

7.7 Control Systems 

7.7.1 Scope 

This section includes those control systems used for normal operation that are 

not relied upon to perform safety functions following anticipated operational 

occurrences or accidents but that control plant processes having a significant impact on plant safety. Examples include the reactivity control systems; the 

reactor coolant pressure, temperature, flow, and inventory controls; the secondary system pressure and flow controls; and the environmental control systems for safety-related instruments and instrument sensing lines. 

7.7.2 Safety Issues and Regulatory Requirements 

The licensing of earlier plants in the control system area usually encompassed the review of the interaction of the control systems with the safety systems that have been discussed in the previous sections. This review was performed 

to ensure that no interactions existed that would prevent or inhibit the safety 

system from performing its intended safety function. 

7.7.3 Evolution of Current Licensing Basis 

As the licensing process evolved, the review of the control system mentioned 

above became somewhat more detailed. In addition, the environmental control 

systems were added to the list of significant control systems and a regulatory guide that detailed the review bases for this system was published. During the 

later licensing years, plant-specific studies were performed to determine the 

effects of high energy line breaks on control systems, the effects of the loss 

of power to control systems used to shut the plant down in a normal manner, and 

the results of multiple control system failures on the existing safety analysis. Accordingly, requirements and criteria for the review of these control systems have been included in the SRP, and they have been made a part of the licensing 

bases. 

A generic study was undertaken (USI A-47) that led to the conclusion that 

some modifications to plants should be made and that the failure of some control 

systems would have an impact on the safety analysis and, therefore, surveillance 

of these systems should be included in the technical specifications along with 

the safety systems mentioned in the previous sections. For example, Generic 

Letter 89-19 requested all reactor licensees to install, if not already present, overfill protection instrumentation. All responses to the generic letters as 

well as all changes will be reviewed by the staff. 

7.8 General Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that instrumentation and control systems, including the reactor trip, 

7-6 
engineered safety features actuation, safe shutdown and safety-related display 

systems, met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will continue to obtain new information related 

to this subject area through a variety of sources such as updates to the FSAR, 

research reports, operating plant events and routine plant inspections. The 

staff reviews this information and, in the past, has required licensees to take 

actions to upgrade the plant to provide continuing assurance of adequate protec- tion of the public health and safety. In addition, the staff will continue to 

review new information in this subject area and if the staff determines that new 

or different requirements are needed, the staff has the capability within the 

existing regulatory process to require additional analyses or plant modifications, 

as necessary, to ensure the continued health and safety of the public. In 

conclusion, the Commission concludes that the current regulatory process has 

and will continue to provide reasonable assurance that the licensing bases of 

all currently operating plants are sufficient to assure that operation is not 

inimical to the public health and safety. 

7-7 

8 ELECTRIC POWER 

8.1 Scope 

Electric power systems power safety-related equipment that is necessary for 

mitigating the consequences of design-basis accidents and for bringing the plant 

to a safe condition and maintaining it in that condition. Electric power systems comprise an offsite power system and an onsite power system. These two 

systems will be discussed jointly in this chapter since their licensing basis 

is often contained in common regulatory requirements. 

8.2 Safety Issues and Regulatory Requirements 

Commission regulations establish the basic criteria to which the offsite power 

system and the onsite power system must be designed. These regulations require that each system (offsite and onsite) have sufficient capacity and capability by itself to support vital functions necessary to respond to operational occurrences and mitigate the consequences of design-basis accidents. In addition, 

the onsite power system must be able to withstand a single failure and the offsite power system must have two power circuits designed and located so as to 

minimize to the extent practical their simultaneous failure. 

The regulations also require that the electric power systems be designed to 

permit appropriate periodic testing and inspection and that all operating plants have the capability to withstand and recover from a station blackout (loss of 

all ac power). These regulations also apply to portions of the electric power 

systems insofar as they provide general requirements for safety systems or 

provide requirements for systems that interface with the electric power systems. 

8.3 Evolution of Current Licensing Basis 

Commission regulations published in February 1971 constitute the primary licensing basis for the electric power systems. More recently, these regulations have been supplemented with the requirements in 10 CFR 50.63, published in June 

1988_, that require all plants be able to withstand and recover from a station 

blackout (loss of all ac power). The station blackout rule illustrates how the 

regulatory process functions to modify the licensing basis in the electric power 

systems area when a need is identified. 

As operating experience was accumulated from license event reports (LERs), diesel 

generator failure reports, and feedback from the regions, a concern arose that 

the offsite and onsite emergency ac power systems might be less reliable than 

originally anticipated, even designs that met the requirements of the Commission 

regulations. Some operating plants had experienced a total loss of offsite 

power, and operating experience with onsite emergency power systems included 

many instances of diesel generators failing to start. In a few cases, there 

was even a complete loss of both the offsite and the onsite ac power systems, although ac power was restored in a short time without any serious consequences. In 1975, the results of the Reactor Safety Study (WASH-1400) showed that station 

8-1 
blackout could be an important contributor to the total risk from nuclear power plant accidents. Although this total risk was found to be small, the relative 

importance of the station blackout accident was established. Subsequently, the 

Commission designated the issue of station blackout as an unresolved safety issue (USI), and initiated studies to determine if additional safety requirements 

were needed. 

As a result of the station blackout studies, a proposed rule was published for 

comment in the Federal Register in March 1986. The final rule was published in June 1988. Concurrent with the development of this regulatory guidance, the 

Nuclear Management and Resources Council (NUMARC) also developed detailed guide- lines and procedures for assessing station blackout capabilities at light-water 

reactors (NUMARC 87-00). The staff reviewed and approved NUMARC 87-00 and found 

the guidance therein acceptable for implementing the station blackout rule. The 

purpose of the effort in developing the NUMARC 87-00 guidelines and the staff's 

cooperation with this effort was to iron out differences and misunderstandings in advance and to establish acceptable approaches to various station blackout 

issues for all utilities in responding to the rule. 

In April 1988, the staff received all licensee responses to the station blackout 

rule. The staff is presently reviewing these submittals and will issue safety evaluation reports (SERs) for each plant when the review is completed. It is 

expected that all licensees will have implemented all required modifications and 

procedure changes within 3 years. Although final plans for inspection have not 

been completed, it is likely that an audit inspection will be performed at some 

plants to monitor licensee implementation efforts. 

In addition to rule changes, the staff has employed other less rigorous methods 

to make improvements in the electric power systems area when a need is identified. These include the use of generic letters, bulletins, revision to regulatory guides, creation of new regulatory guides, modifications to the standard 

review plan, and, more recently, cooperation with the nuclear power industry in the development of industry-sponsored guidance documents. 

Generic Safety Issues B-23 and B-48 on degraded grid voltages and station 

electric distribution system voltages are examples of the use of generic letters 

to implement improvements to electric power systems. Events at Millstone and 

Arkansas Nuclear One power plants raised a concern that the offsite power systems required by Commission regulations may not satisfy the capability requirements of the criteria because they may not always provide adequate voltages to 

operate safety-related loads. This could cause loss of or damage to redundant 

safety systems during an event. As a result, the staff issued generic letters 

to all power reactor licensees in June 1977 and August 1979 requesting that they analyze their electric distribution systems for adequate voltages and describe 

to the staff the modifications to upgrade the protection of electrical relaying that separates the offsite power system from the safety loads when voltage levels 

are insufficient to operate these loads. These guidelines were later incorpo- rated into a new branch technical position (PSB-1) in the standard review plan in order to ensure they are consistently applied to new plant license applica- tions. As of today, all operating plants have submitted and received approval for plant modifications that implemented a second level of voltage protection for their safety-related electrical buses. 

8-2 
NRC Bulletin 88-10 on nonconforming molded-case circuit breakers is an example of how the staff has used a bulletin to require that licensees take some action 

to verify that their electrical system is in conformance with Commission requirements. Here again it was a question of whether the plant electric power systems 

were continuing to meet the capability requirements specified in the applicable Commission regulations. The staff found that at some plants circuit breakers 

supplied by a particular vendor were refurbished, not new as indicated by the 

supplier, and that several breakers did not meet required performance specifi- cations. If these breakers were used in safety-related circuits, the reliable 

functioning of the circuit could not be assured. As a result, the staff issued 

Bulletin 88-10 in November 1988 requesting that all licensees verify traceability of certain circuit breakers used in safety systems and test those breakers where 

traceability to the original manufacturer could not be shown and report the 

results to the Commission. The staff is reviewing the licensee responses to 

determine if the licensee has implemented the actions contained in the bulletin. 

If the staff determines that the licensee has implemented the actions contained 

in the bulletin, no further staff action will be performed. If a licensee proposes alternative actions, the staff will handle these proposals on a case-bycase basis. 

Generic Safety Issue (GSI) B-56 on diesel generator reliability improvement 

serves as an example of the use of revised regulatory guides and industry- sponsored guidance documents to implement improvements to electric power safety 

systems. GSI B-56 was initiated as a response to the lower than expected reliability of diesel generators as emergency power sources in the onsite power 

systems. It is related to the station blackout issue in that it is one of the 

primary sources of unreliability in the total loss of ac power event. As a 

result, the station blackout regulatory guidance called for a reliability program at nuclear power plants designed to monitor and maintain the reliability of the diesel generators and improve the reliability if an acceptable level is 

not achieved. Specific guidance to the utilities on how to implement such a 

program is being provided under the GSI B-56 resolution in the form of a revision 

to Regulatory Guide 1.9, which will reference a NUMARC document for the diesel 

generator reliability program recommendations. The NUMARC document (NUMARC 

87-00, Appendix D) was produced by the nuclear power industry with input from 

the NRC staff (as described above) in the discussion on station blackout. The 

resolution of GSI B-56 will be complete when, consistent with the requirements of the station blackout rule, each licensee implements an emergency diesel 

generator reliability program to enhance the reliability of the onsite diesel 

generators. 

Problems in the electric power systems, such as those discussed above, are 

identified by the staff on an ongoing basis through the review of LERs and other 

licensee notification requirements and through the staff's various license 

review and inspection activities. In addition to the vehicles identified above 

for making changes to the licensing basis for electric power systems, the staff 

often issues information notices to notify licensees of problems found in the 

electric systems at some plants. Although the notices do not require licensees 

to take any action, they serve to quickly advise licensees of problems that may exist in their plants, while the staff determines what, if any, additional action 

is warranted. 

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8.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the design of the facility's electrical power systems met the 

staff acceptance criteria and the intent of all applicable regulations. The 

staff has and will continue to obtain new information related to this subject 

area through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. For example, the Commission recently issued 10 CFR 50.63 to 

require electrical system upgrades. In addition, the staff will continue to 

review new information in this subject area and if the staff determines that 

new or different requirements are needed, the staff has the capability within 

the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has 

and will continue to provide reasonable assurance that the licensing bases of 

all currently operating plants are sufficient to assure that operation is not 

inimical to the public health and safety. 

8-4 
9 AUXILIARY SYSTEMS 

9.1 Scope 

Auxiliary systems are those secondary systems provided to support operation and 

function of primary engineered safety features (ESFs); other systems not directly related to safe reactor operation and safe shutdown are also included. Their 

primary function is to remove heat from essential components (e.g., cooling 

water and ventilation systems) or provide motive power (e.g., compressed air) 

to equipment needed for safe reactor operation and postaccident shutdown. Support systems include cooling water systems (e.g., station service water, reactor 

auxiliaries cooling water, and the ultimate heat sink); compressed air systems; heating, ventilation, and air conditioning (HVAC) systems for various plant 

areas; and diesel generator auxiliaries (e.g., fuel oil, cooling water, lubrication, and combustion air systems). Other auxiliary systems not directly related to safe reactor operation and shutdown include new and spent fuel storage and handling systems, process sampling system, equipment and floor drainage 

system, fire protection system, and communication and lighting systems. In 

addition, the chemical and volume control system that provides normal reactor 

coolant system inventory in pressurized-water reactor (PWR) plants and the 

standby liquid control system that provides an emergency backup means of reactivity control in BWR plants are also within the scope of the auxiliary systems. 

9.2 Safety Issues and Regulatory Requirements 

Commission regulations require that cooling water systems supporting primary ESF systems be designed, tested, and inspected in a manner intended to ensure 

their safety function and that ESF systems be compatible with environmental conditions, which includes HVAC systems relied on to provide proper ESF equipment operating conditions. These regulations also require that spent fuel storage and handling systems be designed with features that ensure a safe spent fuel 

storage facility and that nuclear power plants be designed to minimize the prob- ability of fires and have fire protection features to minimize the adverse 

effects of fires. Specific additional fire protection requirements are 

contained in 10 CFR 50.48 and Appendix R to 10 CFR Part 50. 

9.3 Evolution of Current Licensing Basis 

The licensing basis for auxiliary systems has evolved as reactor events and 

generic studies by the NRC staff provide new information that is determined to 

be helpful in improving the performance of auxiliary systems. The process of 

evaluating operating experience and assessing plant data to determine the need 

for additional actions is a continuous one. 

One important source of operating experience information is the reports on events 

and equipment failures prepared by all licensees in accordance with the requirements of 10 CFR 50.72 and 10 CFR 50.73. The NRC Office for Analysis and Evaluation of Operational Data (AEOD) reviews this information and develops recommendations for action. Examples of this with regard to auxiliary systems are 

discussed below. 

9-1 
Another means of identifying the need for further actions is through the 

process of identifying, prioritizing, and evaluating generic issues when poten- tial safety concerns arise that require longer term study. Examples of the 

generic issue process on auxiliary systems are also discussed below. 

In 1980, the NRC staff became aware through reported events of fouling of 

service water systems and the resulting degradation in system performance caused 

by biological organisms. This resulted in the issuance of IE Bulletin 81-03, 

"Flow Blockage of Cooling Water to Safety-Related Components by Corbiocula Sp. (Asiatic Clam) and Mytilus Sp. (Mussel), 11 dated April 10, 1981. Licensees were 

requested to assess the potential for biofouling at their sites and implement appropriate monitoring or corrective actions. Subsequent service water system problems were also identified and generically communicated in IE Information 

Notice (IN) 85-30, 11Microbiologically Induced Corrosion of Containment Service 

Water System" (April 19, 1985); IN 86-11, "Inadequate Service Water Protection 

Against Core Melt Frequency" (February 25, 1986); and IN 86-96, "Heat Exchanger Fouling Can Cause Inadequate Operability of Service Water Systems 11 (November 20, 

1986). 

In response to service water system degradation problems, several generic issues 

(Gis), but primarily GI-51, "Proposed Requirements for Improving Open Cycle Service Water Systems," were initiated to study the need for further recommendations for improving service water system performance. In addition, as part of 

its responsibility to evaluate operational data, AEOD undertook a study of service water system problems. The AEOD findings were eventually published in 

"Operational Experience Feedback Report - Service Water System Failures and 

Degradations," NUREG-1275, Volume 3, dated November 1988. 

The AEOD report and GI-51 resolution led to development and issuance of Generic 

Letter (GL) 89-13, "Service Water System Problems Affecting Safety-Related Equip- ment," which recommended additional performance monitoring and design verifications in order to ensure the safety function of the service water system. All 

plants were required to respond to GL 89-13 by indicating their plans for accomplishing the NRC staff's intent to improve service water system performance. Through the inspection program, the NRC staff is performing audits of the imple- mentation of the actions identified by the licensees in response to the generic letter and will assess the adequacy of the licensee 1 s actions. 

A similar process was followed during the early 1980s to correct reported failures and problems with degradation in the instrument air systems. Parallel 

NRC staff evaluations were conducted under GI-43, "Air System Reliability," and 

in AEOD, which resulted in publication of "Operational Experience Feedback 

Report - Air System Problems," NUREG-1275, Volume 2, dated December 1987. The 

GI-43 resolution and the AEOD report led to development and issuance of GL 88-14, 

"Instrument Air Supply System Problems Affecting Safety-Related Equipment," which 

requested that licensees perform a design-basis verification of their instrument 

air systems and make the necessary improvements to ensure its proper function. 

All licensees were required to respond to GL 88-14 indicating that they had 

accomplished the recommended actions. Within the framework of the inspection 

program, the NRC staff is performing audit inspections of instrument air systems 

to assess the adequacy of the licensees• actions for improving the systems• 

performance. 

9-2 
After the TMI-2 accident, it became apparent from analysis of the event that 

additional means were necessary to ensure prompt and accurate postaccident sampling of the containment environment and reactor coolant conditions in order 

to get the information needed to manage recovery. On the basis of experience gained from the TMI-2 accident, Item II.B.3, "Post-Accident Sampling Capability," 

was incorporated in NUREG-0737. All nuclear power plants were required to 

implement postaccident sampling capabilities in accordance with these guidelines. 

Because of the safety significance of the fire at Browns Ferry Unit 1 in 1975, 

the staff undertook a comprehensive effort to develop more specific criteria to 

improve fire safety. This effort resulted in issuance of various staff positions and guidance in 1976, which included Branch Technical Position (BTP) Auxiliary and Power Conversion System Branch (APCSB) 9.5-1 and Appendix A to BTP APCSB 

9.5-1. The Commission eventually codified fire protection requirements when 

it issued 10 CFR 50.48, "Fire Protection," and Appendix R to 10 CFR Part 50, 

dated November 19, 1980. All licensees whose plants had been licensed to operate before January 1, 1979, were required to compare their plant fire protection features to the new criteria and make the necessary modifications. Plants not 

licensed in 1979 or later were reviewed against similar criteria as part of the 

normal prelicensing review. Completion of these actions has resulted in substantial improvement in fire protection and post-fire safe shutdown capability in all plants. 

With the recognition in the mid-1970s that spent fuel from commercial nuclear 

power plants would not be reprocessed, it became apparent that much greater quantities of spent fuel would be stored in onsite spent fuel pools. This led 

to the development of additional guidance for ensuring safe spent fuel storage when license amendments were requested for expanding capacity in spent fuel 

pools. This guidance was issued to all nuclear power reactor licensees by a 

generic letter, dated April 14, 1978. This guidance continues to serve as a 

basis for ensuring safe onsite storage of spent fuel. Subsequent generic 

concerns regarding such spent fuel storage safety were evaluated by the NRC 

staff under Generic Issue 82, "Beyond Design Basis Accidents in Spent Fuel 

Pools." This effort led to the determination that additional criteria beyond these currently established for ensuring safe spent fuel storage were not necessary. In addition, the requirements of 10 CFR Part 72 must be satisfied if a 

licensee proposes to store spent fuel in an independent storage facility 

separate from the spent fuel pool itself. 

Concerns with regard to the safe handling of heavy loads at nuclear power plants 

were the subject of a generic study under Generic Technical Activity A-36 during the late 1970s. This study resulted in publication of NUREG-0612, "Control of 

Heavy Loads at Nuclear Power Plants, 11 dated July 1980, and issuance of a generic letter, dated December 22, 1980. The generic letter requested that, as stated 

in NUREG-0612, all licensees implement improvements to procedures, training, identification of safe load paths, and crane and lifting device maintenance and 

testing, in order to reduce the probability of a heavy load drop near spent fuel 

or safety-related equipment that could lead to an unacceptable release of radioactivity. The staff reviewed and provided a safety evaluation of each licensee's 

proposed actions to handle heavy loads more safely. Through the inspection program, the NRC staff audited licensee-identified actions to satisfy the concerns 

identified in NUREG-0612. Following these reviews, the staff undertook a pilot 

program to assess the need for implementation of additional NUREG-0612 guide- lines. On the basis of the pilot program, the staff determined that further 

9-3 
actions recommended in NUREG-0612, which included the installation of single- failure-proof cranes and performance of load drop analyses, were not necessary. This conclusion was described in Generic Letter 85-11, dated June 28, 1985, 

wherein it was determined that actions already completed by licensees have 

satisfactorily reduced the probability of unacceptable heavy load drops. 

Since the TMI-2 accident, the staff has begun in a systematic manner to review 

the capability of nuclear power plants to cope with beyond-design-basis (severe) 

accidents. This effort has relied largely on probabilistic risk assessment 

techniques. A major objective of these reviews was to identify potential plant vulnerabilities and take corrective actions accordingly. These reviews show that 

auxiliary and support systems can be dominant contributors to risk, and attention to their continued proper operation is important to plant safety. Future 

licensee activities requested by the staff as part of the Individual Plant 

Examination of internally initiated events (IPE) and the Individual Plant Examination of External Events (IPEEE) will include a focus on auxiliary systems and 

their contribution to plant safety. 

9.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the facility's auxiliary systems met the staff acceptance criteria 

and the intent of all applicable regulations. The staff has and will continue 

to obtain new information related to this subject area through a variety of 

sources such as updates to the FSAR, research reports, operating plant events 

and routine plant inspections. The staff reviews this information and, in the 

past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. Examples of instances where staff review and examination has led to issuance 

of new criteria and implementation of improvements to auxiliary systems include 

service water system upgrades (GL 89-13), instrument air system upgrades (GL 88-14), post accident sampling in public upgrades (NUREG 0737, Item 11.B.3), 

fire protection upgrades (10 CFR 50.48 and Appendix R to 10 CFR Part 50), spent fuel storage upgrades (GL of April 14, 1978), and heavy load capability upgrades (NUREG-0612). In addition, the staff will continue to review new information 

in this subject area and if the staff determines that new or different requirements are needed, the staff has the capability within the existing regulatory 

process to require additional analyses or plant modifications, as necessary, to 

ensure the continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has and will continue to 

provide reasonable assurance that the licensing bases of all currently operating plants are sufficient to assure that operation is not inimical to the public health and safety. 

9-4 
10 STEAM AND POWER CONVERSION SYSTEM 

10.1 Scope 

The steam and power conversion system consists of those balance-of-plant systems 

necessary to provide feedwater to the reactor in boiling-water reactors (BWRs) 

and steam generators in pressurized-water reactors (PWRs) in order to produce the main steam supply to the turbine for generating power as part of the normal 

operating function of the nuclear power plant. With the exception of system piping interfaces to the primary (in BWRs) and secondary pressure boundary (in 

PWRs), these systems have no safety function and are not relied on to ensure a 

safe postaccident shutdown with one exception. The auxiliary feedwater system in PWR plants has an important postaccident and transient decay heat removal 

safety function, as is discussed below. 

10.2 Safety Issues and Regulatory Requirements 

Commission regulations require that the reactor coolant pressure boundary have 

an extremely low probability of abnormal leakage, of rapidly propagating failure, 

and of gross rupture under operating, maintenance, testing, and postulated accident conditions. These requirements pertain to the steam and power conversion system in PWRs because control of the secondary water chemistry and inservice inspection to technical specifications limits are essential to ensuring 

steam generator tube integrity and preventing unacceptable primary coolant leakage into the secondary (steam) system. 

In addition, the regulations require that nuclear power plants have a system to 

remove residual heat following accidents and transients and specify design requirements for the system. In PWRs, the auxiliary feedwater system provides this function for most events, except for postulated large reactor coolant piping 

failures. 

10.3 Evolution of Current Licensing Basis 

The 1979 accident at Three Mile Island, Unit 2 (TMI-2) heightened the NRC staff 1 s 

awareness of the importance of the postaccident decay heat removal safety function provided by the auxiliary feedwater system. Improper isolation of the 

auxiliary feedwater flow path at TMl-2 delayed the initiation of decay heat 

removal through the steam generators. This resulted in implementation of Items 

11.E.1.1 and 11.E.1.2 of NUREG-0737, which required upgrades to the auxiliary feedwater system in all PWR plants to improve its reliability. The specific improvements were identified in NUREG-0611 and NUREG-0635 and included changes in system design, including initiation and flow indication, operating procedures, and technical specifications. The staff reviewed licensee responses to this 

item and wrote a safety evaluation for each plant. 

Despite the improvements obtained by this effort, concern with auxiliary feedwater system reliability remained. This concern grew out of the review of 

auxiliary feedwater system reliability studies and continued failures noted from 

10-1 
operating experience data reviews. The specific concern was that the availability of an auxiliary feedwater system with two pumps was not sufficient to ensure 

the secondary decay heat removal safety function when compared to that of a 

three-pump system. This concern was amplified by the loss of all feedwater 

event at Davis-Besse in June 1985. The staff pursued the issue under Generic 

Issue (GI) 124 where those few plants with just two auxiliary feedwater pumps 

were evaluated to determine the need to make further hardware changes to improve their auxiliary feedwater system reliability. 

GI 124 was ultimately resolved with a requirement that two-pump auxiliary feedwater system plants backfit a third means of removing decay heat through the steam generators. Those few licensees affected by this decision have 

committed to implement this additional improvement for ensuring the auxiliary feedwater system safety function. 

In addition, although the remaining portions of the power and conversion system do not perform a direct function in ensuring postaccident plant safety, events 

involving balance-of-plant systems have made the staff recognize that certain 

improvements were necessary in order to ensure the safety function of interfacing 

systems or reduce the likelihood of unanticipated plant trips. Two such areas 

of improvement include preventing erosion/corrosion and preventing waterhammer. 

In the 1970s, waterhammer events in main feedwater systems at several PWR 

plants, including Indian Point Unit 2, Calvert Cliffs, and others, demonstrated 

the need for hardware improvements in order to reduce the chance of breaching the secondary side of the steam generator. The staff evaluated this issue under 

Unresolved Safety Issue (USI) A-1. As a result of this effort, PWRs have 

installed J-tubes on the feedwater ring header within the steam generator to 

reduce the likelihood of steam void formation in the feedwater line and potential waterhammer from collapse of the steam bubble on auxiliary feedwater system initiation. These actions have been effective at reducing the probability of 

damaging waterhammer. 

The feedwater line break event at Surry Unit 2 in December 1986 pointed out the 

adverse consequences to plant safety from unplanned reactor trips caused by to 

balance-of-plant failures and to personnel from high energy system steam 

releases. As a result, the staff issued Bulletin 87-01, which requested that 

all licensees examine plant piping for wall thinning and take corrective action 

as necessary. A subsequent audit of licensee actions in response to Bulletin 

87-01 indicated that continued programs to monitor for future erosion/corrosion 

were not in place at the plants. Therefore, the staff issued Generic Letter 

89-08, which requested that licensees implement a continuous monitoring program 

to detect unacceptable pipe wall thinning and certify that the program is in 

place. These programs provide the necessary assurance against the type of severe 

wall thinning event that challenges plant safety systems. Through the inspection 

program, the staff audits licensee actions to ensure that adequate implementation 

has been undertaken. 

As a result of steam generator tube degradation and leakage problems at many PWR 

plants in the 1970s, PWR licensees, nuclear steam supply system (NSSS) vendors, 

and the NRC staff initiated studies to improve steam generator tube integrity. One major early outcome of these studies was a recognition that typical secondary 

water chemistry programs that included sodium phosphate were potentially contributing to the tube degradation being experienced. As a result, NSSS vendors 

10-2 
recommended a change to an all volatile treatment (AVT), secondary water chemistry program utilizing ammonia and hydrazine. Licensees have adopted this change, and subsequent operating experience indicates that it has been effective in 

improving steam generator tube integrity. Steam generator tube integrity was 

also designated an unresolved safety issue (LISI) by the NRC staff in 1978 and 

Task Action Plans (TAPs) A-3, A-4, and A-5 were established to evaluate the 

safety significance of degradation in Westinghouse, Combustion Engineering, and 

Babcock and Wilcox steam generators, respectively. These studies were later 

combined into one effort because many of the problems being experienced at these 

plants were similar. The staff prepared a draft LISI report regarding this issue, 

which primarily considered corrosion-related failure mechanisms, including the 11 denting11 mechanism, since those failures were the main concern during the 

period when most of the technical studies were performed. 

In May 1982, subsequent to the Ginna steam generator tube rupture (SGTR) event, 

the staff initiated an integrated program to consider the lessons learned from 

the Ginna SGTR event, and from the three previous domestic SGTR events. The 

staff also considered the recommendations in the draft LISI report. The objec- tive of the integrated program was to resolve USis A-3, A-4, and A-5 and determine the need for further requirements to improve steam generator tube integrity. 

Concurrent with the completion of the staff study under USis A-3, A-4, and A-5, 

in 1985, the NRC staff issued Generic Letter 85-02, which requested that all PWR 

licensees describe their programs, including secondary water chemistry control, 

for ensuring steam generator tube integrity. The NRC staff reviewed these 

programs and accepted them with necessary changes made by licensees. 

The results of the NRC staff integrated program for resolution of USis A-3, A-4, 

and A-5 were ultimately documented in NUREG-0844, 11 NRC Integrated Program for 

the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity, 11 September 1988. In NUREG-0844, the NRC staff concluded 

that sufficient regulatory requirements were in place, in conjunction with 

industry initiatives, to ensure that SGTRs do not contribute significantly to 

nuclear power plant risk, and thus, no further regulatory requirements were 

necessary. 

10.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that those portions of the steam and power conversion system performing essential safety functions met the staff acceptance criteria and the intent of 

all applicable regulations. The staff has and will continue to obtain new information related to this subject area through a variety of sources such as updates 

to the FSAR, research reports, operating plant events and routine plant inspec- tions. The staff reviews this information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of 

adequate protection of the public health and safety. The TMI-2 accident led to 

significant improvements in the availability and reliability of the auxiliary feedwater system, which is the portion of the steam and power conversion system providing a decay heat removal safety function in PWRs. Events at other plants in balance-of-plant systems led to improvements to reduce the likelihood of 

10-3 
damaging waterhammer and unanticipated plant trips due to errosion/corrosion 

of piping. In addition, the staff will continue to review new information in 

this subject area and if the staff determines that new or different requirements 

are needed, the staff has the capability within the existing regulatory process 

to require additional analyses or plant modifications, as necessary, to ensure 

the continued health and safety of the public. In conclusion, the Commission 

concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing bases of all currently operating plants 

are sufficient to assure that operation is not inimical to the public health 

and safety. 

10-4 
11 RADIOACTIVE WASTE MANAGEMENT SYSTEMS 

11.1 Scope 

Radioactive waste management systems are provided to control releases of 

radioactive materials to the environment in liquid and airborne effluents and 

to handle radioactive solid wastes produced during normal reactor operation. Process and effluent radiological monitoring and sampling systems are provided for monitoring effluent discharge paths for radioactivity that may be released 

from normal operations and from postulated accidents. 

Radioactive liquid and solid waste management systems are relatively independent of the type of plant; however, radioactive gaseous waste management systems at 

boiling-water reactors (BWRs) differ significantly from those at pressurized 

water reactors (PWRs). 

11.2 Safety Issues and Regulatory Requirements 

Commission regulations limit and otherwise govern the radioactivity in effluents 

released to unrestricted areas for the purpose of providing protection against the hazards of radiation from normal plant operation. These regulations (1) pro- vide requirements regarding the characteristics of radioactive waste prepared and packaged for transfer to offsite disposal sites, (2) provide design objec- tives for equipment to control releases of radioactive materials in effluents, 

and (3) require technical specifications to keep releases of radioactive 

materials to unrestricted areas during normal operations as low as is reasonably achievable. 

Furthermore, Commission regulations require that (1) the plant design include 

means to suitably control the release of radioactive materials in gaseous and 

liquid effluents and to handle radioactive solid wastes produced during normal 

operations, including anticipated operational occurrences; (2) radioactive waste 

systems be designed to ensure adequate safety under normal and postulated accident conditions; and (3) means be provided for monitoring effluent discharge paths for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents. 

11.3 Evolution of Current Licensing Basis 

Significant changes to the regulations governing radioactive waste management 

systems occurred with the establishment of Appendix I to 10 CFR Part 50 in 1975 

and 10 CFR 20.311, and 10 CFR Part 61 in 1982. Appendix I was issued because 

the NRC recognized that specific numerical criteria were necessary to ensure 

that licensees were maintaining radioactivity levels within normal effluent 

releases to as low as reasonably achievable limits as required by 10 CFR Part 20. 

The technical specifications established under 10 CFR 50.36a ( 11Appendix I 

technical specifications 11 ) for all plants are intended to ensure that radioactive 

waste processing operations are conducted within specific limits. These technical specifications provide limiting conditions for operation and surveillance 

11-1 
requirements regarding (1) operation of the liquid and gaseous radwaste treatment systems, (2) radioactive materials in liquid and gaseous effluents, 

(3) offsite doses due to radioactive materials in liquid and gaseous effluents, 

(4) total offsite doses, (5) radiological environmental monitoring, (6) content 

of liquid and gaseous waste storage tanks, (7) explosive mixtures in gaseous radwastes management systems, and (8) processing of solid radioactive wastes. 

Subsequent experience and problems with the acceptability of solid waste packages from commercial nuclear power plants intended for burial at licensed offsite 

facilities resulted in issuance of 10 CFR Part 20 requirements to more closely control transfers of radioactive waste intended for disposal at a licensed land 

disposal facility. These problems included excessive amounts of water in solid 

waste packages and overly rapid deterioration of the waste form itself. Under 

10 CFR 20.311, licensees are required to (1) prepare all solid wastes so that 

they can be classified according to 10 CFR 61.55, (2) meet the waste characteristic requirements of 10 CFR 61.56, and (3) conduct a quality control program 

to ensure compliance with 10 CFR 61.55 and 10 CFR 61.56. These requirements are 

intended to ensure that future solid waste packages from all plants will be 

acceptable for burial at storage facilities and will maintain their long-term integrity. 

Additional generic requirements that have evolved governing radioactive waste 

management systems are as follows. NUREG-0737, 11 Clarification of TMI Action 

Plan Requirements," Item 11.F.1, issued in November 1980, provided new generic requirements regarding additional monitoring and sampling, and analysis of post- accident releases of radioactive materials. These requirements were added 

because of weaknesses noted at Three Mile Island, Unit 2 (TMI-2) in this area 

after the TMI-2 accident. 

In the early 1980s, uncertainty arose with regard to future availability of 

low-level waste disposal capacity at the licensed burial sites. This concern 

resulted in the issuance of Generic Letter (GL) 81-38, 11 Storage of Low-Level 

Radioactive Wastes at Power Reactor Sites, 11 dated November 10, 1981, which provided generic guidance to be used by licensees in the design, construction, and 

operation of such onsite storage facilities. Licensees continue to use this 

guidance to ensure proper storage of low-level waste at nuclear power plants. 

11.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the radioactive waste management system met the staff acceptance 

criteria and the intent of all applicable regulations. The staff has and will 

continue to obtain new information related to this subject area through a variety 

of sources such as updates to the FSAR, research reports, operating plant events 

and routine plant inspections. The staff reviews this information and, in the 

past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. 

In addition, the staff will continue to review new information in this subject 

area and if the staff determines that new or different requirements are needed, 

the staff has the capability within the existing regulatory process to require 

additional analyses or plant modifications, as necessary, to ensure the continued 

11-2 
health and safety of the public. In conclusion, the Commission concludes that 

the current regulatory process has and will continue to provide reasonable 

assurance that the licensing bases of all currently operating plants are sufficient to assure that operation is not inimical to the public health and safety. 

11-3 

12 RADIATION PROTECTION 

12.1 Scope 

General standards are provided for the protection of the individual from radiation hazards associated with activities licensed by the NRC. In this chapter, the staff will discuss the different control strategies in place to limit the 

exposure of occupational workers and the general public to ionizing radiation. 

12.2 Safety Issues and Regulatory Requirements 

Commission regulations provide standards for the protection of licensees, their 

employees, and the general public against the radiation hazards arising out of 

the possession or use of special nuclear, source, or byproduct material under 

license issued by the NRC. Certain precautionary procedures and administrative 

controls are provided to ensure that the evaluation of these radiation hazards 

are adequate and that the resulting radiation doses are kept as low as is reasonably achievable (ALARA). Different limits and controls are provided for 

occupationally exposed individuals and members of the general public. 

12.2.1 Occupational Exposures 

10 CFR Part 20 prescribes dose limits that govern the exposure of personnel to 

radiation from sources external to the body. In addition, limits on the 

quantities of radioactive material taken into the body through inhalation or 

absorption are provided to control the doses to individual organs and tissues 

from internal sources. 

10 CFR Part 19.12 prescribes that plant workers be informed of the radiation 

hazards to which they are subjected and be instructed in the purpose and function 

of radiation protection devices in use and controls that they must observe. 

12.2.2 Exposures to the General Public 

10 CFR Part 20 provides controls for radiation exposure to the general public by limiting the radiation levels that can exist in areas not controlled by the 

licensee and concentrations of radioactive material that may be discharged from 

the facility in gaseous and liquid form, and by regulating the transportation and disposal of radioactive wastes. 

In addition to the limits of 10 CFR Part 20 that apply to all NRC licensees, 

design specifications and operating requirements are provided in Appendix I to 

10 CFR Part 50 to ensure that each power reactor licensee operates its facility 

so that the quantities of radioactive materials released to the environment in 

gaseous and liquid effluents are maintained ALARA. 

Also, NRC licensees are subject to regulations promulgated by other agencies. The Environmental Protection Agency (EPA) provides limitations on the dose to 

members of the public from facilities in the uranium fuel cycle (including those 

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licensed by the NRC) in 40 CFR Part 190. The Department of Transportation (DOT) 

regulates the shipment of radioactive materials in Title 49 to the Code of 

Federal Regulations (49 CFR). 

12.3 Evolution of Current Licensing Basis 

The adoption of 10 CFR Part 20 in 1957 established the basic framework currently employed for the protection of licensee personnel and the public from exposure 

to radiation. Extensive changes to the dose limits and the permissible concentrations of radioactive material in air and water, contained in 10 CFR Part 20, 

were adopted in 1960. These dose limits and permissible concentrations were 

based on the latest scientific knowledge of the time on the biological effects 

of radiation exposure. 

An assumption basic to the radiation protection methods used in 10 CFR Part 20 

is that any exposure to ionizing radiation results in a proportional health risk 

and that there should be no radiation exposure without a commensurate benefit. 

From 1970 to 1975, the Commission undertook a series of rule changes to improve the framework in 10 CFR Part 20 for ensuring that reasonable efforts are made 

to keep exposures to radiation, and releases of radioactivity in effluents, ALARA 

and to specify design and operating requirements in Appendix I to 10 CFR Part 50 

to restrict quantities of radioactive materials released in gaseous and liquid effluents from light-water reactors (LWRs). 

The dose criteria specified in Appendix I to 10 CFR Part 50 correspond to 

continuous effluent releases that are a small fraction of the concentration 

limits in 10 CFR Part 20. Licensees were required to implement technical 

specifications to ensure plant operations within the Appendix I requirements. 

In 1981, the Commission amended 10 CFR Part 20 to incorporate the EPA requirements in 40 CFR Part 190, 11 Environmental Radiation Protection Standards for 

Nuclear Power Operations"; 40 CFR Part 190 provides that LWRs be operated so 

that releases of radioactive material and the resulting radiation doses to the 

public are below specified limits. These dose limits are comparable to, and in 

some cases more restrictive than, the dose objectives and operating conditions 

contained in Appendix I to 10 CFR Part 50. 

In 1983, 10 CFR 20.311, was adopted to establish administration procedures and 

recordkeeping requirements to support the licensing requirements for land disposal of radioactive wastes contained in 10 CFR Part 61. The waste manifests, 

specified in 10 CFR 20.311, document that radioactive wastes are properly classified, described, packaged, marked, and labeled, and are in proper condition 

for transportation according to the applicable DOT regulations. 

During the licensing process, pursuant to 10 CFR 50.34, the applicant must 

submit a final safety analysis report (FSAR) that describes the facility, the 

kinds and quantities of radioactive materials expected to be produced, and the 

means for controlling and limiting radioactive effluents and radiation exposures 

within the limits of 10 CFR Part 20. 

Additional design criteria are provided in Appendix A to 10 CFR Part 50 governing the radioactive exposure to plant operators under accident conditions, the exposure to radiation during fuel handling and storage, and the adequate monitoring of radioactive concentrations in plant effluents and radiation levels in plant 

environs during normal operations and postulated accidents. 

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Before granting a license, the staff reviews the FSAR against established 

acceptance criteria and concludes that the facility design, and the radiological controls proposed, are adequate, and that the facility can be operated within 

all applicable limits and radiation exposures will be maintained ALARA. 

The NRC inspection program ensures that each licensee adequately implements the 

radiation protection controls described in the plant FSAR and incorporated in 

the plant technical specifications. The performance of each licensee's radiation protection program is inspected by the NRC resident inspectors on a weekly basis, by region-based specialists routinely, and by teams of specialists whenever deemed necessary. When deficiencies are noted, the licensee is required to 

modify its program or implement additional controls as corrective action. An 

example of this is the finding from the 1980-1981 Health Physics Appraisal Team 

inspections that efforts at maintaining occupational radiation exposure ALARA 

lacked licensee support. Subsequently, each licensee implemented additional 

programs to ensure occupational exposures are maintained ALARA. 

Plant operating events also provide new information that may require changes to 

a plant's licensing bases. Two examples are the serious exposure of plant workers during a fuel transfer operation in 1978 and the radiation protection expe- riences during the 1979 accident at Three Mile Island. In both cases, each LWR 

licensee was required to reanalyze its plant design and make appropriate modifications to ensure adequate protection to workers during spent fuel transfer oper- ations or during postulated accident situations. In addition, the TMI accident 

indicated a need throughout the industry to improve accident assessment and monitoring capabilities related to potential radioactive releases off site during 

an accident. Upgrades in radiation protection during fuel handling operations 

were made at all operating reactor licensees as a required response to Bulletin 

78-08, "Radiation Levels From Fuel Element Transfer Tubes, 11 dated June 12, 1978. 

Improvements in radiation protection programs to protect workers during a post- ulated plant accident were contained in NUREG-0660 and NUREG-0737, and the 

implementation at each plant was subsequently required by Commission order. 

Advances in the scientific and technical knowledge of radiobiology and the risks 

associated with radiation exposure have been made since the current limits in 

10 CFR Part 20 were adopted in 1960. The current recommendations of the International Commission oa Radiological Protection (ICRP) provide a radiation pro- tection framework that relates the risks of nonuniform irradiation to individual 

tissues and organs from internally deposited radionuclides to the risk of uniform 

irradiation of the total body. This method differs from the current standards 

in 10 CFR Part 20, which limit internal and external exposures separately. 

The Commission staff has reviewed the ICRP recommendations and has concluded 

that the framework more firmly establishes health risk as the basis for radiation protection than was evident for the current standards. The Commission is 

currently engaged in a rulemaking proceeding to revise 10 CFR Part 20 to be consistent with these international recommendations and practices, even though the 

standards in the current 10 CFR Part 20, in concert with the ALARA programs implemented at LWRs, result in doses generally far below the limits specified in the proposed revision. Thus, revising the limits in 10 CFR Part 20 will have 

little impact on reactor licensees. For example, limiting the sum of the dose 

from internally deposited radioactivity and the dose from external sources will 

not be a significant impact for LWRs since engineering and administrative controls, already required, generally reduce the intake of radioactive material to 

12-3 
insignificant levels. If approved by the Commission, all NRC licensees will be 

required to change their programs to implement the radiation protection strategy provided by the revised 10 CFR Part 20. As additional scientific findings become 

available, such as those recently published by the National Academy of Sciences 

in its fifth report on the biological effects of ionizing radiation (BEIR V), 

the Commission will consider their significance and make changes to its rules 

and regulations as appropriate. 

12.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the control strategies in place to limit the exposure of occupa- tional workers and the general public to ionizing radiation met the staff acceptance criteria and the intent of all applicable regulations. The staff has and 

will continue to obtain new information related to this subject area through a 

variety of sources such as updates to the FSAR, research reports, operating plant 

events and routine plant inspections. The staff reviews this information and, 

in the past, has required licensees to take actions to upgrade the plant to 

provide continuing assurance of adequate protection of the public health and 

safety. In addition, the staff will continue to review new information in this 

subject area and if the staff determines that new or different requirements are 

needed, the staff has the capability within the existing regulatory process to 

require additional analyses or plant modifications, as necessary, to ensure the 

continued health and safety of the public. In conclusion, the Commission concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing bases of all currently operating plants 

are sufficient to assure that operation is not inimical to the public health 

and safety. 

12-4 
13 CONDUCT OF OPERATIONS 

The following sections discuss a number of subject areas that generally affect 

the conduct of operations around operating facilities. These subject areas 

include discussions of the management, operations, and technical support organ- ization; training programs; emergency planning; licensee's self-assessment 

capabilities; plant procedures; and physical security. 

13.l Management, Operations, and Technical Support Organizations 

13.1.1 Scope 

During the licensing of an operating plant, the staff reviews the licensee's 

management and support organizations. This particular review area is limited 

to ensuring that corporate management is involved with, informed about, and 

dedicated to the safe design, test, and operation of the plant; that there are 

sufficient technical resources available to ensure plant operational safety; and that the structure, functions, and responsibilities of the licensee's 

onsite organization are acceptably defined. 

13.1.2 Safety Issues and Regulatory Requirements 

Commission regulations require that a licensee must be technically qualified to 

operate the plant before a license can be granted and that provisions relating 

to organization and management be included in the administrative controls 

section of the plant technical specifications. The regulations also describe 

licensed operator requirements during operation of a facility. The TMI Action 

Plan (NUREG-0737) also describes specific requirements with respect to the 

responsibility of both the shift supervisor and shift technical advisor. 

13.1.3 Evolution of Current Licensing Basis 

In general the safety evaluation report or its supplements contain descriptions of the management, operations, and technical support organizations for each facility at the time the license was issued. A licensee's management, operations, and technical support organizations continually change during the term of the 

license. Changes to the management, operations, and technical support organiza- tion are monitored throughout the term of the license, and new criteria are 

applied, if applicable. 

10 CFR Part 50.71(e) requires each licensee to periodically update the Final 

Safety Analysis Report (FSAR) for their facility. The FSAR contains information 

with respect to the management, operations, and technical support organizations. If changes are made to provisions relating to management and organization that 

are in the technical specifications, the Commission reviews and approves these 

changes. Thereby, the Commission is continually updated on the licensee's 

management and technical support organizations at each facility. 

The management, operations, and technical support organizations at each facility 

are evaluated continually. The Commission, on a continuing basis, interacts 

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with the licensee through the evaluation of reportable events, license changes, 10 CFR Part 50.59 changes, the Commission resident inspector program, and 

special inspections. These activities provide insight into the capability of 

the technical support organization for the facility. The integration of all 

these interfaces with the licensee provides a continual evaluation of the management, operations, and technical support organization at each facility. 

13.1.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the applicant's technical qualifications to manage and support plant operations met the staff acceptance criteria and the intent of all 

applicable regulations. The staff has and will continue to obtain new information related to this subject area through a variety of sources such as 

updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this information and, in the past, has required Jicensees to take actions to upgrade the plant to provide continuing assurance 

of adequate protection of the public health and safety. Reviews and approval of 

technical specification changes, interactions with the staff, and the inspection 

program provide the Commission with a continuing evaluation of the licensee's 

management, operations, and techncial support organizations. In addition, the 

staff will continue to review new information in this subject area and if the 

staff determines that new or differet requirements are needed, the staff has 

the capability within the existing regulatory process to require additional 

analyses of plant modifications, as necessary, to ensure the continue health 

and safety of the public. In conclusion, the Commission concludes that the 

current regulatory process has and will continue to provide reasonable assurance 

that the licensing bases of all currently operating plants are sufficient to 

assure that operation is not inimical to the public health and safety. 

13.2 Training 

13.2.1 Scope 

This section describes information relating to the operational training and 

licensed operator requalification programs of the plant. The purpose of these 

programs is to provide assurance that the licensee will adequately train a staff 

to safely operate the plant and, thereby, protect the public health and safety. 

13.2.2 Safety Issues and Regulatory Requirements 

Commission regulations require that licensees provide training and instruction 

to individuals who manipulate the controls of a facility or direct any licensed 

activity of a licensed individual and provide information concerning organiza- tional structure, personnel qualifications, and related matters to ensure that 

proper administrative and managerial controls are in place to ensure safe 

operation. 

13.2.3 Evolution of Current Licensing Basis 

Following the accident at Three Mile Island Unit 2 (TMI-2), the NRC emphasized the need to upgrade training and qualifications of nuclear power plant person- nel. In the 11 NRC Action Plan Developed as a Result of the TMI-2 Accident 11 

13-2 
(NUREG-0660, July 1980), the NRC cited its ongoing study of accreditation of 

training as a possible means of upgrading training programs in the industry. 

In the 11Clarification of TMI Action Plan Requirements 11 (NUREG-0737, November 

1980), the NRC cited interim procedures to improve training programs and to 

upgrade the qualifications of personnel prior to accreditation of the facility training programs. Since that time, the Institute of Nuclear Power Operations (INPO), with its associated National Academy for Nuclear Power Operations (Academy), has developed a training accreditation program that the NRC has 

found to be an acceptable means of self-improvement of training. 

On March 20, 1985, the Commission published its policy statement on training and qualification of nuclear power plant personnel allowing the industry a 

minimum of two years of accreditation activity without the introduction of new 

NRC training regulations. In the policy statement, the Commission further 

endorsed the training accreditation program managed by INPO, as it encompasses the elements of effective performance-based training and provides the basis to 

ensure that personnel have qualifications commensurate with the performance requirements of their jobs, and recognized the accreditation of 10 utility training programs. 

On November 18, 1988, the Commission published a revised policy statement that 

reflected the minor modifications made by the Academy to its accreditation program and the NRC staff to the methods by which it monitors the industry training 

programs. Specifically, the amendments of the revised policy statement are: 

(1) recognition of the establishment of an eleventh accredited training program; (2) NRC staff will monitor the industry training programs and training program results by conducting post-accreditation reviews; and (3) NRC will conduct inspections, as deemed necessary, and take appropriate enforcement action in accordance with the Commission's enforcement policy in 10 CFR Part 2, Appendix C, 

when regulatory requirements are not met. However, the Commission's policy has 

been successfully challenged [Public Citizens v. U.S. NRC No. 89-1017 - D.C. 

Circuit, April 17, 1990]. The Commission has this matter under consideration. 

To ensure that the nuclear industry's training program improvements are effective, the NRC monitors the accreditation process and its results by attending and observing Accreditation Board meetings, observing training accreditation 

team visits, conducting operator licensing and requalification exams, and conducting performance-oriented training inspections to assess the level of 

knowledge of plant personnel. 

13.2.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the applicant's training of licensed operators met the staff 

acceptance criteria and the intent of all applicable regulations. The staff 

has and will continue to obtain new information related to this subject area 

through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In particular, 10 CFR 50.34 and 10 CFR Part 55 establish 

the requirements for the development and implementation of training and 

13-3 
requalification programs for facility personnel. The Commission has determined 

that the INPO training accreditation program with its periodic re-evaluation 

requirements is an adequate means of self-improvement training. In addition, 

the staff will continue to review new information in this subject area and if 

the staff determines that new or different requirements are needed, the staff 

has the capability within the existing regulatory process to require additional 

analyses of plant modifications, as necessary, to ensure the continue health 

and safety of the public. In conclusion, the Commission concludes that the 

current regulatory process has and will continue to provide reasonable assurance 

that the licensing bases of all currently operating plants are sufficient to 

assure that operation is not inimical to the public health and safety. 

13.3 Emergency Planning 

13.3.l Scope 

This section discusses the requirement that reactor licensees develop and 

implement emergency plans to ensure the continued protection of the public health and safety in the event of a radiological accident. 

13.3.2 Safety Issues and Regulatory Requirements 

Prior to the issuance of a full power operating license, the emergency planning regulations require a finding that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. The regulations were adopted as an added conservatism to the defensein-depth philosophy. They differ in character from most of the NRC 1 s siting and engineering design requirements which are directed at achieving or 

maintaining a minimum level of public safety protection. 

13.3.3 Evolution of Current Licensing Basis 

13.3.3.1 Onsite Emergency Planning 

In June 1979, NRC began a formal consideration of the role of emergency planning for ensuring the continued protection of the public health and safety in areas 

around nuclear power plant facilities. A final rule, effective November 3, 1980, 

was published in the Federal Register on August 19, 1980 (45 FR 55402). It pro- vides that an initial operating license will not be granted unless NRC can make 

a favorable finding that the integration of onsite and offsite emergency planning provides reasonable assurance that adequate protective measures can and will be 

taken in the event of a radiological emergency. NRC will base its finding on a 

review of Federal Emergency Management Agency (FEMA) findings and determinations 

as to whether State and local emergency plans are adequate and capable of being implemented, and on the NRC assessment as to whether the applicant's onsite emergency plans are adequate and capable of being implemented. In the case of an 

operating reactor, if it is determined that there are such deficiencies that a 

favorable NRC finding is not warranted, and if the deficiencies are not corrected 

within four months of that determination, the Commission will determine expeditiously whether the reactor should shut down or whether some other enforcement 

action is appropriate. In any case, where the Commission believes that the public health, safety, or interest so requires, the plant will be required to shut 

down immediately. Licensees, however, will have an opportunity to demonstrate 

13-4 
to the satisfaction of the Commission, for example, that deficiencies in emergency plans are not significant for the plant in question, that adequate interim 

compensating actions have been or will be taken promptly, or that there are 

compelling reasons to permit plant operation. 

The 1980 rule required that emergency planning considerations be extended to 

emergency planning zones and that these consist of an area of about 10 miles 

in radius for exposure to the radioactive plume that might result from an accident in a nuclear power reactor and an area of about 50 miles in radius for 

food that might become contaminated. Additionally, the final rule sets forth 

16 emergency planning standards that must be met by onsite, State, and local 

emergency plans within the emergency planning zones. 

13.3.3.2 Offsite Emergency Planning 

Section 109 of the NRC FY 1980 Authorization Bill (PL 96-295) required that NRC 

consult with the Director of the Federal Emergency Management Agency (FEMA) on 

the status of State radiological emergency response plans with respect to the 

issuance of an operating license for a reactor facility. 

In response, FEMA issued a rule concerning review and approval of State radiological emergency plans and preparedness (44 CFR Part 350, September 28, 1983). 

This rule established policy and procedures for review and approval by FEMA of 

State emergency plans and preparedness for coping with the offsite effects of 

radiological emergencies that might occur at nuclear power facilities. The 

rule sets out criteria that are used by FEMA in reviewing, assessing, and evaluating the plans and preparedness; it specifies how and where a State may submit plans; and it describes certain of the processes by which FEMA makes findings and determinations as to the adequacy of State plans and the capability of 

State and local governments to implement these plans and preparedness measures. 

Such findings and determinations are to be submitted to the Governors of 

affected States and to NRC for use in its licensing proceedings. 

13.3.3.3 Current Program 

As experience was gained in the implementation of the revised onsite and offsite 

emergency plans by both the licensees and the State and local governments, revisions to the regulations were deemed appropriate. For example, the 1980 

regulations required that the licensees and State and local governments within 

the IO-mile plume exposure pathway emergency planning zone conduct an annual, 

full-participation exercise. After considering the experience gained by all of 

the participants in these annual exercises, the Commission proposed, and then 

adopted in 1984 a change to a biennial full-participation exercise. The revised 

rule continued to require an annual onsite exercise for licensees, required 

State and local governments to participate every 2 years with a provision for 

remedial exercises to ensure that any deficiencies are corrected, and provided the opportunity for State and local government participation in the annual licensee exercise, if desired. The rationale behind the change was that (1) expe- rience in observing and evaluating over 150 exercises had shown a disproportionate amount of Federal, State, and local government resources were being expended 

to conduct and evaluate the annual exercises, (2) State and local governments respond to a variety of actual emergencies on a continuing basis, thus exercising their emergency preparedness capabilities, and (3) the flexibility provided for 

13-5 
in a biennial frequency would provide an incentive for State and local governments to perform in a satisfactory manner in order to avoid conducting remedial 

exercises. 

In order to ensure that emergency preparedness around licensed nuclear facilities 

continues to reflect current conditions and circumstances, licensees are permit- ted to make changes to emergency plans without NRC approval if those changes do 

not decrease the effectiveness of these plans, and the plans, as changed, continue to meet the standards of 10 CFR 50.47(b) and the requirements of 10 CFR 

Part 50, Appendix E. Changes made without approval must be reported within 30 

days after the changes are made. Proposed changes that decrease the effectiveness of the approved plans may not be implemented without application to and 

approval by the NRC. This requirement is found in 10 CFR 50.54(q) of the 

regulations. 

As required by 10 CFR 50.47(b)(8), licensees are to provide and maintain adequate 

emergency facilities and equipment. To satisfy this requirement, licensees must 

inspect and perform operability checks of emergency equipment and instruments at 

frequent intervals throughout the year. In addition, the NRC performs an annual 

inspection of the licensee's program and equipment to ensure that essential 

emergency facilities, equipment, instrumentation, and supplies are being maintained in a state of operational readiness. Based on the above discussion, the 

staff does not believe that equipment used in assuring the effectiveness of the 

emergency preparedness program needs to be evaluated as part of the plant assessment of aging degradation required by the new Part 54. 

13.3.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the facility's emergency plans met the staff acceptance criteria 

and the intent of all applicable regulations. The staff has and will continue 

to obtain new information related to this subject area through a variety of 

sources such as updates to the FSAR, research reports, operating plant events 

and routine plant inspections. The staff reviews this information and, in the 

past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information in this subject 

area and if the staff determines that new or differet requirements are needed, 

the staff has the capability within the existing regulatory process to require additional analyses of plant modifications, as necessary, to ensure the continue 

health and safety of the public. In conclusion, the Commission concludes that 

the current regulatory process has and will continue to provide reasonable 

assurance that the licensing bases of all currently operating plants are 

sufficient to assure that operation is not inimical to the public health and 

safety. 

13.4 Review and Audit 

13.4.1 Scope 

This section discusses the licensee's operational review program. The purpose of this program is to implement the licensee's responsibility-related proposed 

13-6 
changes, test evaluations of unplanned events, and provisions for the evaluation of plant operations. 

13.4.2 Safety Issues and Regulatory Requirements 

The Commission regulations require that one of the considerations in granting a 

license is the technical qualifications of the applicant. NUREG-0737, "Clarification of TMI Task Action Plan Requirements," describes the requirements for an 

Independent Safety Engineering Group (ISEG) for post-TM! licensed plants. In 

addition, the regulations require that certain provisions relating to administrative controls be incorporated into the administrative controls section of 

the plant-specific technical specifications. 

13.4.3 Evaluation of Current Licensing Basis 

In general, the safety evaluation report for each facility contains a descrip- tion of the operational review program at the time the facility was licensed. 

A licensee's operational review program is often revised during the term of the 

license. Changes to the program are monitored by the Commission throughout the 

term of the license, and new criteria are applied if applicable. 

10 CFR Part 50.71(e) requires each licensee to periodically update the Final 

Safety Analysis Report (FSAR) for their facility. The FSAR contains a descrip- tion of the facility operational review program. In addition, the Commission 

reviews and approves any changes to the operational review program that are in 

the facility technical specifications. Thereby, the Commission is periodically updated on the current operational review program at each facility. 

The Commission's inspection program provides for periodic evaluation of the 

facility operational review program. Inspection Procedure 40500, "Evaluation of 

Licensee's Self-Assessment Capability, 11 and Inspection Procedure 88005, 11Management Organization and Controls, 11 provide for the periodic inspection and evaluation of the facility operational review program. 

13.4.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the applicants operational review program met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will 

continue to obtain new information related to this subject area through a 

variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to 

upgrade the plant to provide continuing assurance of adequate protection of 

the public health and safety. In addition, the staff will continue to review 

new information in this subject area and if the staff determines that new or 

differet requirements are needed, the staff has the capability within the 

existing regulatory process to require additional analyses of plant modifications, as necessary, to ensure the continue health and safety of the 

public. In conclusion, the Commission concludes that the current regulatory 

13-7 
process has and will continue to provide reasonable assurance that the 

licensing bases of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

13.5 Plant Procedures 

This section discusses two general categories of procedures: administrative 

procedures and operating and maintenance procedures. Administrative procedures include (1) those that provide the administrative controls with respect to pro- cedures, and (2) those that define and provide controls for operational activities of the plant staff. Operating and maintenance procedures are used by the 

operating organization (plant staff) to ensure that routine operating, offnormal, emergency, and maintenance activities are conducted in a safe manner. 

13.5.1 Administrative Procedures 

13.5.1.1 Safety Issues and Regulatory Requirements 

The Commission regulations require that one of the considerations in granting a 

license is the technical qualifications of the applicant to engage in the activities of the license and require that the licensee designate individuals to be 

responsible for directing the activities of licensed operators. The regulations also require provisions relating to administrative controls in the Administrative 

Controls Section of the Technical Specifications. Further, NUREG-0737, 11Clarification of TMI Task Action Plan Requirements, 11 describes certain requirements with respect to administrative procedures requirements. 

3.5.1.2 Evolution of Current Licensing Basis 

The safety evaluation report (SER) and its supplements describe the administrative 

controls program at the time of licensing of the facility. The administrative 

controls relate to, in part, procedures and programs and to designating individuals to be responsible for directing the activities of licensed operators. 

A licensee 1 s procedures and program for the control of procedures may change during the term of the license. Changes to these procedures and programs are 

monitored during the term of the license and new criteria applied if applicable. 

10 CFR Part 50.71(e) requires each licensee to periodically update the Final 

Safety Analysis Report (FSAR) for their facility. The FSAR contains the administrative controls program for procedures. Changes to the procedures control 

program that are included in the technical specifications are reviewed and ap- proved by the Commission. Thereby, the Commission is continually aware of the 

administrative controls program. 

The licensee 1 s administrative controls program at each facility is periodically evaluated through the Commission 1 s inspection program. Xn particular, Inspection Procedure 71707, 11 0perational Safety Verification, 11 and Inspection Procedure 

88005, 11Management Organization and Controls, 11 provide for a continual review of 

the licensee 1 s administrative procedures control program. 

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13.5.1.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the applicant's administrative controls program met the staff 

acceptance criteria and the intent of all applicable regulations. The staff 

has and will continue to obtain new information related to this subject area 

through a variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to 

upgrade the plant to provide continuing assurance of adequate protection of 

the public health and safety. In addition, the staff will continue to review. 

new information in this subject area and if the staff determines that new or 

differet requirements are needed, the staff has the capability within the 

existing regulatory process to require additional analyses of plant modifications, as necessary, to ensure the continue health and safety of the 

public. In conclusion, the Commission concludes that the current regulatory 

process has and will continue to provide reasonable assurance that the 

licensing bases of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

13.5.2 Operating and Maintenance Procedures 

13.5.2.1 Safety Issues and Regulatory Requirements 

The regulations applicable to administrative procedures require the determination that the licensee is technically qualified to engage in licensing activities and that the licensee designate individuals to be responsible for directing the licensed activities of licensed operators. 

Commission regulations also govern operating procedures used by licensed 

operators in the control room and other operating procedures and maintenance 

procedures. Additionally, the TMI Action Plan (NUREG-0660 and NUREG-0737) 

requires licensees to upgrade their Emergency Operating Procedures {EOPs). 

Commission regulations also require that activities affecting quality be 

prescribed by and accomplished in accordance with documented instructions, 

procedures, and drawings. 

13.5.2.2 Evolution of Current Licensing Basis 

Requirements for the commercial nuclear power industry to improve the quality and usability of plant procedures were established as a result of the accident 

at Three Mile Island (TMI). Following TMI, the NRC Office of Nuclear Reactor 

Regulation developed the TMI Action Plan (NUREG-0660 and NUREG-0737) which required licensees of operating reactors to reanalyze transients and accidents 

and to upgrade EOPs (Item I.C.l). NUREG-0660 (Item I.C.9) committed the NRC to 

develop a long-term plan for the overall improvement of nuclear power plant 

proeedures. 

Requirements for EOPs were further defined in Generic Letter 82-33. Generic 

Letter 82-33 transmitted Supplement 1 to NUREG-0737, 11 Requirements for Emergency 

13-9 
Response Capability, 11 and directed each licensee to submit to the NRC a Procedures Generation Package (PGP) from which licensees were to develop function or 

symptom-based EOPs. This document also indicated that the NRC staff would audit 

EOPs on a selective basis. 

Early reviews of EOP programs identified potential concerns with their imple- mentation. In response to these findings, the NRC staff conducted inspections 

to monitor the industry 1 s procedure upgrade programs. Initial inspections revealed a number of problems, and Information Notice 86-64 was issued in August 1986 to alert licensees to these problems. Subsequent inspections revealed similar results and Information Notice 86-64, Supplement 1, was issued on April 20, 

1987, to describe further problems with EOPs and PGPs and to inform the industry that the inspection effort would be intensified. NRC Temporary Instruction 2515/ 

92 was issued in April 1988 and defines the objectives of the EOP inspection. The inspection effort now extends to all operating reactors in the United States 

and has two objectives: (1) to assess the adequacy of the EOPs themselves, and 

(2) to establish that the supporting programs and documents are sufficient to 

ensure the integrity and continued adequacy of the EOPs. 

13.5.2.3 Conclusions. 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the applicant 1 s procedure revision process met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will 

continue to obtain new information related to this subject area through a , variety of sources such as updates to the FSAR, research reports, operating plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to 

upgrade the plant to provide continuing assurance of adequate protection of 

the public health and safety. In addition, the staff will continue to review 

new information in this subject area and if the staff determines that new or 

different requirements are needed, the staff has the capability within the 

existing regulatory process to require additional analyses of plant modifications, as necessary, to ensure the continue health and safety of the 

public. In conclusion, the Commission concludes that the current regulatory 

process has and will continue to provide reasonable assurance that the 

licensing bases of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

13.6 Physical Security 

13.6.1 Scope 

This section discusses the evolution of the basis of the reactor security program. The licensee 1 s security program consists of the following three plans: security, security contingency, and guard training and qualification. These 

three plans provi.de the physical protection envelope that provides the assurances that the operation of these plants does not constitute an unreasonable 

risk to the public health and safety. 

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13.6.2 Safety Issues and Regulatory Requirements 

The Commission regulations require licensees to establish and maintain a physical protection system and security organization that provides high assurance against radiological sabotage. 

13.6.3 Evolution of Current Licensing Basis 

The purpose of nuclear power reactor security requirements is to protect against the design basis threat of radiological sabotage. The design basis threat is 

generally considered to be the worst-case scenario of attack by sever a 1 we 11-

trai ned and dedicated individuals and an individual inside the facility. In 

1977, specific requirements of physical protection of licensed nuclear facilities against radiological sabotage were set forth by the NRC in 10 CFR 73.55. 

In publishing this rule, the Commission stated the following: 11The level of 

protection specified in Part 73.55 is adequate and prudent at this time. The 

kind and degree of threats will continue to be reviewed by the Commission. 

Should such reviews show change that would dictate different levels of protec- tion, the Commission would consider changes to meet the changed conditions 

(42 FR 10836). 11 

The Commission has since made a number of changes to the requirement to maintain or increase the level of assurance. In 1978, the Commission issued requirements for a safeguards contingency plan and guard training and qualifica- tion plans to be prepared and noted in a facility's license conditions. Subsequent changes in Part 73 have required the reporting of physical security 

events, the protection of unclassified safeguards information, and the 11Miscellaneous Amendments. 11 Those amendments include a refined vital area access 

policy, authority to suspend safeguards during safety emergencies, protection of certain safeguards equipment, and upgrades to key and lock controls. Most 

recently the regulations have been revised to require that any individual in 

need of unescorted access at a facility submit to a Federal Bureau of Investigation fingerprint check and chemical testing to determine that they are 

fit-for-duty. 

These changes were made to ensure that the level of protection remains adequate considering all new information and potential threats. In 1989 the Commission 

requested licensees to include in their safeguards contingency plan procedures for short-term actions to protect against attempted radiological sabotage involving a land vehicle bomb if such a threat were to materialize. 

The NRC has conducted Regulatory Effectiveness Reviews (RERs) since 1982 to 

ensure that safeguards required by NRC's regulations, as implemented by licensees, provide the intended level of protection without compromising safety of 

operations. The RER teams use NRC security personnel and members of the U.S. 

Army Special Forces to test plant security systems and personnel. Regional safeguards inspectors continue their routine unannounced and special inspec- tions at all licensed facilities. 

In addition to continued NRC review of industry-wide conditions, the status of 

physical security measures are reviewed at each individual plant in the System- atic Assessment of Licensee Performance (SALP) program. Both headquarters and 

regional safeguards staff provide comments for the 11 Security11 functional area. 

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Licensee-initiated changes to approved security plans (also contingency and 

guard training) may only be made by two methods. Changes made pursuant to 10 

CFR 50.54(p) may be made without prior Commission approval if the changes do 

not decrease the safeguards effectiveness of the plan. The changes must be 

submitted to the Commission within two months and changes are reviewed by the 

staff. The second method for plan changes involves the amendment process as 

specified in 10 CFR 50.90 to include reviews by the staff and Federal Register notices soliciting public comment. These changes may involve measures that are 

not contained in 10 CFR 73.55(b) through (h), but provide the equivalent high 

assurance against radiological sabotage. 

Age-related degradation of safeguards equipment is not a license renewal issue 

because it is an issue that is being currently experienced and managed. A number of the originally licensed sites have reached the life expectancy of certain 

types of security equipment. Because of the general performance objectives and 

requirements of 10 CFR 73.SS(a) and the site-specific commitments contained in 

the individual plant security plans, normal inspection activities will force the 

replacement of degraded equipment or subject the licensee to enforcement action. 

13.6.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the physical security program met the staff acceptance criteria 

and the intent of all applicable regulations. The staff has and will continue 

to obtain new information related to this subject area through a variety of 

sources such as updates to the FSAR, research reports, operating plant events 

and routine plant inspections. The staff reviews this information and, in the 

past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. In addition, the staff will continue to review new information in this subject 

area and if the staff determines that new or different requirements are needed, 

the staff has the capability within the existing regulatory process to require additional analyses of plant modifications, as necessary, to ensure the continue 

health and safety of the public. In conclusion, the Commission concludes that 

the current regulatory process has and will continue to provide reasonable 

assurance that the licensing bases of all currently operating plants are 

sufficient to assure that operation is not inimical to the public health 

and safety. 

13-12 
14 INITIAL TEST PROGRAM 

14.1 Safety Issues and Regulatory Requirements 

Commission regulations require, in part, that an applicant for a license to 

operate a production or utilization facility include the principal design criteria for the proposed facility in the safety analysis report (SAR). These 

regulations state that these principal design criteria are to establish the 

necessary design, fabrication, construction, testing, and performance requirements for systems, structures, and components important to safety, that is, 

systems, structures, and components that provide reasonable assurance that the 

facility can be operated without undue risk to the health and safety of the 

public. 

These regulations also require that a test program be established to ensure 

that systems, structures, and components will perform satisfactorily in service. 

Since all functions designated in the general design criteria are important 

to safety, all systems, structures, and components required to perform these 

functions need to be tested to ensure that they will perform properly. These 

functions, as noted throughout the specific general design criteria, are those 

necessary to ensure that specified design conditions of the facility are not 

exceeded during any condition of normal operation, including anticipated opera- tional occurrences, or as a result of postulated accident conditions. 

14.2 Evolution of Current Licensing Basis 

The NRC safety evaluation report (SER) and supplements describe and attest to· 

the adequacy of the initial test program for each facility at the time the 

license was issued. Upon completion of the test program, the results are documented in a final test report subsequent to issuance of an operating license 

for each facility. The satisfactory completion of the test program provides 

assurance that the systems, structures, and components important to safety will 

perform as designed and that the facility can be operated without undue risk to 

the health and safety of the public. During the term of the initial license, 

Commission oversight, regulatory actions, and the implementation of technical 

specifications provkte assurance that the plant continues to meet the current 

licensing basis. This is sufficient to conclude that the level of safety is 

also adequate for continued operation during any renewal period. 

14.3 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the licensee's test program met the staff acceptance criteria 

and the intent of all applicabli regulations. The staff has and will continue 

to ootain new information related to this subject area through a variety of 

sources such as updates to the FSAR, research reports, operating plant events 

and routine plant inspections. The staff reviews this information and, in the 

past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. 

14-1 
In addition, the staff will continue to review new information in this subject 

area and if the staff determines that new or different requirements are 

needed, the staff has the capability within the existing regulatory process to 

require additional analyses or plant modifications, as necessary, to ensure 

the continued health and safety of the public. In conclusion, the Commission 

concludes that the current regulatory process has and will continue to provide reasonable assurance that the licensing basis of all currently operating plants are sufficient to assure that operation is not inimical to the public health and safety. 

14-2 
15 ACCIDENT ANALYSES 

15.1 Scope 

This chapter addresses the analyses of the response of the plant to postulated accidents and to postulated malfunctions or failures of equipment. Such safety analyses provide a significant contribution to the selection of limiting conditions of operation, limiting safety system settings, and design specifications for components and systems from the standpoint of public health and safety. Also, 

the effects of anticipated accidents and postulated component failures are examined to determine their consequences and to evaluate the capability built into 

the plant to control or accommodate such failures and situations. The situations 

analyzed include anticipated operational occurrences (e.g., a loss of electrical 

load resulting from a line fault), off-design transients that include a small 

amount of fuel failures, and postulated accidents of low probability (e.g., the 

sudden loss of integrity of reactor coolant pressure boundary). The analyses include an assessment of the consequences of an assumed fission-product release. 

15.2 Safety Issues and Regulatory Requirements 

Commission regulations require, in part, that the reactor core and associated 

coolant, control, and protection systems shall be designed with appropriate margin to ensure that specified acceptable fuel design limits (SAFDLs) are not 

exceeded during any condition of normal operation, including the effects of 

anticipated operational occurrences; that the reactor coolant system and 

associated auxiliary, control, and protection systems shall be designed with 

sufficient margin to ensure that the design conditions of the reactor coolant 

pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences; and that redundant and reliable 

reactivity control systems are provided to ensure that under conditions of 

normal operation, including anticipated operational occurrences, SAFDLs are not 

exceeded. 

15.3 Evolution of Current Licensing Basis 

The licensing basis for transient and accident analyses has evolved as reactor 

events provide new information that is determined to provide improvement in the 

methods of evaluation. The process of evaluating operating experience and 

assessing plant data to determine the need for additional actions is a continuing 

one. 

General staff guidance specifies that the transients and accidents analyzed in 

the plant safety analysis report ensure that a sufficiently broad spectrum of 

initiating events has been considered; ensure that initiating events of certain 

types and expected frequencies of occurrence be analyzed so that only the limiting cases in each group are quantitatively evaluated; and permit the consistent 

application of specific acceptance criteria for each postulated initiating event. 

In general, each initiating event is assigned to one of three frequency groups: incidents of moderate frequency, infrequent incidents, or limiting faults. The 

15-1 
quantitative evaluation of each initiating event in each of the three frequency 

groups establishes the limiting conditions of operation for the required safety 

systems and the limiting parameters are routinely placed in the plant technical 

specifications to ensure that the plant is operated within its established 

design envelope. 

The evolution of the current licensing basis regarding the performance of the 

emergency core cooling system following a postulated loss-of-coolant accident 

is discussed in Section 6.4.2 of this report. 

The 1979 accident at Three Mile Island, Unit 2 (TMI-2) raised several issues 

(TMI action items) that affect the management of plant transients and accidents. 

Items II.E.1.1 and II.E.1.2 of NUREG-0737 required upgrades to the auxiliary feedwater system in all pressurized-water-reactor (PWR) plants to improve its 

reliability. The specific improvements include changes in design regarding system initiation and flow indication. Item II.K.3.5 provides guidelines on automatic trip of the reactor coolant pump during a postulated loss-of-coolant accident. Item II.K.3.44 requires an evaluation of anticipated transient with single failure to verify no significant fuel failure. 

Within the framework of the Systematic Evaluation Program (SEP), the staff 

evaluated the need for all plants to adopt the reactor coolant system specific activity limits found with the Standard Technical Specifications (SEP Lessons 

Learned Issue 6.3). The coolant activity levels have a proportionate effect 

on those accidents involving primary coolant release (without core damage) to 

the environment. Implementation of Standard Technical Specifications (STS) 

limits are usually adequate to alleviate the concerns regarding resultant 

offsite doses. 

The staff examined the plant 1 s technical specifications to determine the degree of compliance with the appropriate STS. An evaluation was performed to determine the adequacy of the existing plant technical specification limits in 

restricting offisite dose. The review covered those accidents whose primary dose contribution is from reactor coolant leakage to the atmosphere (e.g., main 

steam line break outside of containment, steam generator tube rupture, and small 

line breaks outside containment). 

With respect to boiling-water reactors (BWRs), the staff reviewed Generic Safety Issue (GSI) 74. On the basis of this review, the staff determined that GSI 74 

had low safety significance and that no further staff review was warranted. 

Additionally, as a result of improvements in nuclear fuel performance, steam 

generator performance for PWRs, and chemistry control for both PWRs and BWRs, 

the potential safety significance of this issue has been further reduced. 

Therefore, the staff concluded that this SEP lessons learned issue have been 

adequately addressed by other regulatory initiatives. The staff no longer believes that this is an issue for any generating plant. 

15.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that transient and accident analyses conducted by the licensee met 

the staff acceptance criteria and the intent of all applicable regulations. The staff has and will continue to obtain new information related to this 

15-2 
subject area through a variety of sources such as updates to the FSAR, 

research reports, operating plant events and routine plant inspections. The 

staff reviews this information and, in the past, has required licensees to 

take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. Before each plant refueling, the 

transient and accident analyses are reviewed by each licensee to verify the 

changes resulting from new core don not result in an unreviewed safety question. If an unreviewed safety question arises, or if any technical 

specifications require modification, staff review and approval is required before plant restart. In addition, the staff will continue to review new 

information in this subject area and if the staff determines that new or 

different requirements are needed, the staff has the capability within the 

existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued health and safety of the 

public. In conclusion, the Commission concludes that the current regulatory 

process has and will continue to provide reasonable assurance that the 

licensing bases of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

15-3 

16 TECHNICAL SPECIFICATIONS 

16.1 Scope 

Each applicant for an operating license is required to submit proposed technical 

specifications and their bases for the facility as a chapter in the final safety analysis report (FSAR). They should be consistent with the content and format 

of the Standard Technical Specifications available from the Commission for the 

appropriate nuclear steam supply system (NSSS) vendor. After review and needed 

modification by the NRC staff, these technical specifications are issued by the 

Commission as Appendix A to the operating license. 

The equipment included in the technical specifications is a broad spectrum of 

structures and electrical and mechanical systems and components taken from the 

safety analyses of the FSAR or updated safety analysis report (USAR). It includes such structures as the reactor vessel and containment, such systems such 

as the emergency core cooling system and reactor protection system, and such 

components as circuit breakers, valves, and pumps in these systems. 

16.2 Safety Issues and Regulatory Requirements 

Commission regulations require that each license issued by the NRC authorizing operation of a utilization facility shall include technical specifications. These regulations also describe the required contents of the technical specifi- cations. Safety limits, settings for automatic protective devices, and limiting conditions for operation are required to be included in these technical 

specifications. Surveillances are also required to ensure that the necessary quality of systems and components is maintained, that important parameters are 

maintained within specified limits, and that the limiting conditions for opera- tion are satisfied. Compensatory actions, which may include shutting down the 

reactor, are required when it is found that these conditions are not met. 

The technical specifications are derived from the analyses in the final (or 

updated) safety analysis report. They ensure that the plant will be operated 

so that the assumptions of these safety analyses remain valid. The assumptions include both initial conditions and availability of equipment. 

16.3 Evolution of Current Licensing Basis 

Technical specifications are required by the Atomic Energy Act of 1954 (Section 

182), which states, in part, that 

The applicant shall state such technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of 

the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that 

the utilization or production of special nuclear material will .... provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued. 

16-1 
In 1968, the Commission issue 10 CFR 50.36 to specify the content of the 

technical specifications. From this time until 1973, each plant's technical 

specifications were unique but similar. In 1973, the concept of Standard Technical Specifications was introduced in an attempt to make technical specifica- tions of different plants more consistent. When a plant is ready to be licensed, 

it uses the applicable Standard Technical Specifications as a starting point. 

Licensees typically request changes to the technical specifications in accordance with existing regulations as the plant is operated throughout its life 

to reflect modifications to the design and different methods of operation. When such changes are requested, the NRC must review and approve the requested changes before they can be implemented. In addition, the NRC also requires changes to the technical specifications as new safety and licensing issues 

arise. For example, changes were required to plant technical specifications as 

a result of the Three Mile Island, Unit 2 accident. The resolution of other 

important issues, such as the potential for overpressurization of reactor 

vessels at low temperatures and isolation of low-pressure systems from highpressure systems, have led to additions to technical specifications to ensure 

that plant operation is in conformance with the resolution of these problems. 

As discussed above, the surveillances required by the technical specifications 

ensure that the plant is operated so that the technical specifications requirements are met. Technical specifications requirements on equipment are primarily 

a check on operability of the equipment. Degradation (as, for example, from 

aging) is in most cases not specifically required to be measured, although the 

American Society of American Engineers Boiler and Pressure Vessel Code (ASME 

Code) (which is incorporated in most technical specifications) requires a limited 

amount of trending of performance for pumps and valves. However, the surveillances are generally done frequently enough so that degradation is not expected 

to occur to the extent that operability is affected between surveillances. 

16.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that the plant's technical specifications met the staff acceptance criteria and the intent of all applicable regulations. The staff has and will 

continue to obtain new information related to this subject area through a 

variety of sources such as updates to the FSAR, research reports, operating 

plant events and routine plant inspections. The staff reviews this 

information and, in the past, has required licensees to take actions to 

upgrade the plant to provide continuing assurance of adequate protection of 

the public health and safety. In addition, the staff will continue to review 

new information in this subject area and if the staff determines that new or 

different requirements are needed, the staff has the capability within the 

existing regulatory process to require additional analyses of plant modifications, as necessary, to ensure the continue health and safety of the 

public. In conclusion, the Commission concludes that the current regulatory 

process has and will continue to provide reasonable assurance that the 

licensing bases of all currently operating plants are sufficient to assure 

that operation is not inimical to the public health and safety. 

16-2 
17 QUALITY ASSURANCE 

17.1 Scope 

The quality assurance (QA) program of licensees applies to systems, structures, 

and components that prevent or mitigate the consequences of postulated accidents 

that could cause undue risk to the health and safety of the public. The QA program of each licensee is reviewed by the NRC to ensure that it meets the requirements of Appendix B to 10 CFR Part 50, and the NRC performs inspections to 

determine whether the program is being implemented effectively. 

17.2 Safety Issues and Regulatory Requirements 

Commission regulations require that licensees establish and maintain a QA 

program for the design, construction, and operation of systems, structures, and 

components that prevent or mitigate the consequences of postulated accidents 

that could cause undue risk to the health and safety of the public. 

17.3 Evolution of Current Licensing Basis 

In publishing a proposed rule in 1970 to add Appendix B to 10 CFR Part 50, the 

Commission stated that its purpose was to establish QA requirements for systems, 

structures, and components to prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. Further, the Commission stated that the requirements of Appendix B would apply 

to all activities during design, construction, and operation of such systems, 

structures, and components and that the criteria of Appendix B would be used 

for guidance in evaluating the adequacy of the QA programs in use by holders of 

both construction permits and operating licenses. In essence, Appendix B establishes the minimum acceptable QA requirements for providing reasonable assurance 

that (1) applicable regulatory requirements and the design basis for systems, 

structures, and components, as specified in the license application, are correctly translated into specifications, drawings, procedures, and instructions 

and (2) subsequent activities, such as construction, operation, testing, refueling, repair, maintenance, modification, and decommissioning, are conducted and verified in accordance with appropriate procedures and instructions. 

In the early 1980s, the Commission identified a concern with plant-specific implementation and modification of NRC-approved QA programs. The Commission 

noted that changes being made in previously approved QA programs could diminish 

their effectiveness and result in unacceptable QA programs at some licensed 

facilities. In publishing a final rule addressing this concern, the Commission 

stated that an NRC-approved QA program becomes a principal inspection and 

enforcement tool in ensuring that a licensee is in compliance with QA requirements for protecting the public health and safety. In addition, the final rule 

[10 CFR 50.54(a) and 10 CFR 50.55(f)] established a procedure requiring review 

and approval by the Commission before implementing any change to a previously approved QA program that would reduce its effectiveness. 

17-1 
The final rule also required that each licensee submit a current description of 

its QA program and thereafter submit any revisions annually for NRC review and 

re-approval. Through these requirements, the Commission established an accept- able baseline for a QA program at each plant against which future changes to the 

program would be judged, and ensured that future changes would be available. 

This rule change created a regulatory process by which the Commission ensures 

that an acceptable QA program will remain in place at a licensed facility throughout the life of the license and that changes to that program would be 

routinely reviewed and evaluated to ensure that the program would continue to 

satisfy the regulatory requirements of Appendix B to 10 CFR Part 50. Toward 

ensuring this end, the NRC routinely inspects the implementation of QA programs. Safety-related activities undertaken by licensees to obtain a renewed license 

are also subject to the requirements of Appendix 8. 

17.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that QA program met the staff acceptance criteria and the intent of 

all applicable regulations. The staff has and will continue to obtain new 

information related to this subject area,through a variety of sources such as 

updates to the FSAR, research reports, operating plant events and routine 

plant inspections. The staff reviews this information and, in the past, has 

required licensees to take actions to upgrade the plant to provide continuing 

assurance of adequate protection of the public health and safety. Under 

10 CFR 50.54(a) and 10 CFR 50.55(e), any subsequent change to relax or reduce 

the previous commitments of the QA program of a licensee must receive NRC 

approval before the licensee can implement the change. In addition, the staff 

will continue to review new information in this subject area and if the staff 

determines that new or different requirements are needed, the staff has the 

capability within the existing regulatory process to require additional 

analyses or plant modifications, as necessary, to ensure the continued health 

and safety of the public. In conclusion, the Commission concludes that the 

current regulatory process has and will continue to provide reasonable 

assurance that the licensing bases of all currently operating plants are 

sufficient to assure that operation is not inimical to the public health and 

safety. 

17-2 
18 HUMAN FACTORS ENGINEERING 

18.l Scope 

This chapter describes the regulatory requirements in the human factors area. 

Human factors engineering has played a significant role in control room design and in the technical areas of safety parameter display. 

18.2 Control Room 

18.2.1 Scope 

Nuclear power plants have a control room from which employees can operate the 

plant safely under normal and accident conditions. Outside the control room, 

equipment has the design capability for prompt hot shutdown of the reactor, 

which includes the necessary instrumentation and controls to maintain the plant in a safe condition during hot shutdown, and with a potential capability for 

cold shutdown. 

18.2.2 Safety Issues and Regulatory Requirements 

The safety issue addressed is to confirm that the design of the plant's control 

room and remote shutdown capability facilitates the plant operator's ability to 

prevent accidents or cope with accidents if they do occur. 

The basis for regulating the design of the plant's control room and remote 

shutdown capability is given in the enclosure to Generic Letter 82-33, "Supplement 1 to NUREG-0737--Requirements for Emergency Response Capability." 

18.2.3 Evolution of Current Licensing Basis 

Requirements for commercial nuclear power plants to review their control room 

design and correct deficiencies were established as a result of the Three Mile 

Island (TMI) accident. In May 1980, NUREG-0660, 11 TMI Action Plan Developed as 

a Result of the TMI-2 Accident,11 was issued. Item I.D.l, "Control Room Design Reviews, 11 stated that 11 NRR wi 11 require that operating reactor licensees and 

applicants for operating licenses perform a detailed control room design review 

to identify and correct design deficiencies." The review was to be performed on 

a schedule consistent with the implementation of other requirements for enhancing operator effectiveness, including necessary retraining. In November 1980, 

the NRC published NUREG-0737, "Clarification of TMI Action Plan Items, 11 which 

identified the requirements associated with detailed control room design reviews 

(DCRDRs). Guidance published as NUREG-0700, "Guidelines for Control Room Design Reviews" (1981), was also issued to the industry. In December 1982, "Supplement 1 to NUREG-0737--Requi rements for Emergency Response Capability, 11 was issued as 

Generic Letter 82-33. This document implemented existing requirements for plants 

to conduct a DCRDR and identify human engineering discrepancies and provided additional clarification. For some, but not all, plants, the NRC issued confirmatory orders, which required plants to submit schedules for completing a program plan and a summary report (including a proposed schedule for implementation) of 

their DCRDRs. 

18-1 
For operating plants, the staff has reviewed the program plans for conducting and implementing the DCRDR. These plants also have submitted a summary report of their completed review, which outlines proposed control room changes and 

implementation schedules. Using established criteria, the staff reviews plant- specific summary reports and determines whether a preimplementation audit is 

necessary. After completing its review, the staff issues a safety evaluation 

report documenting the acceptance of the licensee 1 s proposals. 

Since the issuance of Supplement 1 to NUREG-0737, no new requirements have been 

identified for completing the DCRDR. Through periodic resident and regional inspections, and 10 CFR 50.59 reviews, the staff will ensure that future modifications to plant control rooms and remote shutdown facilities are implemented in a manner consistent with the plant 1 s approved DCRDR process and NRC acceptance criteria. The staff will review and evaluate advances in technology that may affect the design of the plant control room or remote shutdown facility through periodic plant inspections and by sponsoring research in advanced control room 

design. If changes to the current requirements are needed, they could be imple- mented using existing regulatory programs, as necessary, to ensure continued 

public health and safety. 

18.2.4 Conclusions 

The staff review at the time of the initial licensing of a facility determined 

that the information provided by the applicant was sufficient for the staff to 

conclude that Control Room Design met the staff acceptance criteria and the 

intent of all applicable regulations. The staff has and will continue to 

obtain new information related to this subject area through a variety of 

sources such as updates to the FSAR, research reports, operating plant events 

and routine plant inspections. The staff reviews this information and, in the 

past, has required licensees to take actions to upgrade the plant to provide continuing assurance of adequate protection of the public health and safety. 

For example, the requirements for the DCRDRs were established in response to 

the TMI accident and are contained in NUREG-0737, Supplement 1. The staff 

reviews and approves plant-specific DCRDR efforts and documents these approvals in published safety evaluation reports after the review is completed. Resident 

and regional inspections, and 10 CFR 50.59 reviews, will ensure that future 

modifications to plant control rooms and remote shutdown facilities are made 

in accordance with NRC approved DCRDR programs and NRC acceptance criteria. 

In addition, the staff will continue to review new information in this subject 

are and if the staff determines that new or different requirements are needed, 

the staff has the capability within the existing regulatory process to require additional analyses or plant modifications, as necessary, to ensure the continued 

health and safety of the public. In conclusion, the Commission concludes that 

the current regulatory process has and will continue to provide reasonable 

assurance that the licensing bases of all currently operating plants are sufficient 

to assure that operation is not inimical to the public health and safety. 

18-2 
18.3 Safety Parameter Display System 

18.3.1 Scope 

In addition to upgrading the design of their control rooms, licensees are to 

install a safety parameter display system (SPDS) as an aid to operating person- nel in rapidly and reliably determining the safety status of the plant and in 

assessing whether abnormal conditions warrant corrective actions to avoid a 

degraded core. 

18.3.2 Safety Issues and Regulatory Requirements 

The safety issue addressed is to confirm that the design and implementation of 

the plant 1 s SPDS facilitate the user 1 s ability to rapidly and reliably determine 

the safety status of the plant. 

In May 1980, requirements for commercial nuclear power plant licensees to 

install an SPDS were established as a result of the TMI accident. The basis 

for regulating the design and implementation of the plant 1 s SPDS is given in the 

enclosure to Generic Letter 82-33, 11 Supplement 1 to NUREG-0737--Requirements for 

Emergency Response Capability. 11 

18.3.3 Evolution of Current Licensing Basis 

NUREG-0660, Item I.D.2, 11 Plant Safety Parameter Display Console/ stated that 

11 In conjunction with the control room design upgrade described in Item I.D.1, 

NRR will require all licensees and applicants to install a safety parameter display system that will display to operating personnel a minimum set of parameters (safety state vector) which define the safety status of the plant. 11 In November 

1980, the NRC published NUREG-0737, 11Clarification of TMI Action Plan Items,U 

which identified the specific requirements associated with the SPDS. Guidance 

published as NUREG-0695, 11 Functional Criteria for Emergency Response Facilities 11 

(1980), and NUREG-0835, 11 Human Factors Acceptance Criteria for the Safety Parameter Display System, Draft Report for Comment 11 (1981), were also issued to the 

industry. In December 1982, 11 Supplement 1 to NUREG-0737--Requirements for Emergency Response Capability/ was issued as Generic Letter 82-33. This document 

implemented the requirements to install an SPDS and provided additional clarification. For some, but not all, plants, the NRC issued confirmatory orders, 

which required plants to submit schedules for the design, installation, and 

implementation of an SPDS. 

In 1986, the staff issued NUREG/CR-4797, 11 Progress Reviews of Six Safety Parameter Display Systems, 11 and concluded that utilities may be having major difficulties in designing and implementing their SPDSs. The staff subsequently issued NRC Office of Inspection and Enforcement (IE) Information Notice (IN) 

86-10, 11 Safety Parameter Display System Malfunctions 11 (1986), to inform licensees of the results of the survey. After issuing IN 86-10, the staff received 

several requests from licensees for extensions to implementation schedules, 

requests for clarification regarding the definition of an 11 operational SPDS,U 

and questions about SPDS deficiencies and how to resolve them. In response to 

the continuing concerns related to SPDS designs, NRC published NUREG-1342, 11 A 

Status Report Regarding Industry Implementation of Safety Parameter Display Sys- tems,11 which described methods used by some licensees to implement the SPDS in a 

18-3 
manner acceptable to the staff and issued Generic Letter 89-06 in April 1989 

requesting that licensees certify the operational status of their SPDSs to the 

NRC, using guidance contained in the generic letter and NUREG-1342. Ihe staff 

is presently reviewing licensee submittals requested by GL 89-06. 

Since the issuance of NUREG-0737, Supplement 1, and the guidance described above, 

no new requirements have been identified for installing and implementing the 

SPDS. Through the process of periodic resident and regional inspections and 

reviews, the staff will ensure that future modifications to plant SPDSs are 

implemented in a manner consistent with the plant 1 s approved design and NRC 

acceptance criteria. The staff will review and evaluate advances in technology that may affect the design of the SPDS through periodic plant inspections and 

by sponsoring research in advanced SPDS design. If changes to the current 

requirements are needed, they could be implemented by rulemaking on existing requirements or under the backfit rule to ensure continued health and safety to 

the public. 

18.3.4 Conclusions 

The requirements for the SPDS were established in response to the TMI accident 

and are contained in NUREG-0737, Supplement 1. Since NUREG-0737, Supplement 1, 

was issued, there have been no new requirements for SPDS. The staff reviews and 

approves plant SPDS designs, and a safety evaluation report is issued after the 

staff completes its review. Resident and regional inspections and 10 CFR 50.59 

reviews will ensure that future modifications to plant SPDS designs are made in 

accordance with NRC guidance and acceptance criteria. Advances in design technology will be reviewed by the staff and if changes to the existing requirements 

are necessary, they will be implemented within existing regulatory programs, as 

necessary, to ensure the continued public health and safety. In conclusion, the 

Commission concludes that the current regulatory process has and will continue 

to provide reasonable assurance that the licensing bases of all currently operating plants are sufficient to assure that operation is not inimical to the 

public health and safety. 

18-4 
19 RESOLUTION OF SAFETY ISSUES: TECHNICAL AND IMPLEMENTATION STATUS 

19.1 Scope 

The NRC has an integrated program in place for reviewing and analyzing operating experience in order to identify specific events and generic situations in which 

the margin of safety established by design through the licensing process has 

been degraded, or where new information or insights lead to new concerns. The 

program also includes steps to identify and implement corrective actions that 

will restore the intended margin of safety. 

19.2 Safety Issues and Regulatory Requirements 

NRC licensees must report any unexpected occurrence in operation that has actual 

or potential safety significance. Some events must be reported within one hour 

via dedicated direct phone lines, and many must be reported in writing within a 

few weeks. These written reports, required by 10 CFR 50.73, are called licensee 

event reports (LERs) and provide a clear, narrative description of the event and 

the cause of each component or system failure, if the cause is known. The staff 

reviews these LERs to determine the adequacy of short-term corrective actions 

and the need for possible action at other plants, or to identify potential generic problems and significant safety concerns warranting further study. 

19.3 Regulatory Process and Implementation Status 

For many safety-related operational events, NRC resident inspectors perform the 

initial NRC investigations, and the appropriate NRC regional office conducts 

reviews. In addition, the technical aspects of potentially significant opera- tional events are studied by appropriate organizations within the NRC, including the Office for Analysis and Evaluation of Operational Data (AEOD) and the Offices 

of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES). 

The AEOD analyzes and evaluates all operational safety data and provides a 

strong technical capability that is independent of regulatory activities associated with licensing and inspection. Engineering evaluations are performed to 

examine the implications of operating experience and to determine if intensive 

analysis and evaluation as a case study are warranted. If necessary, in-depth 

case studies are performed to determine the level of safety concern, and findings and recommendations are sent to the appropriate NRC office for action. 

The AEOD recommendations and suggestions addressed to NRR are reviewed and 

prioritized according to a judgment of their safety significance. If an item 

appears to have a high degree of safety significance, the need for an information notice, generic letter, bulletin, or other appropriate prompt action is 

determined. If the recommendation does not appear to warrant immediate action, 

it is considered within NRR for appropriate action or a determination whether it 

can be addressed as part of an existing issue (such as a generic issue) or by creation of a new generic issue. If this occurs, the issue is formally transmitted to RES for its consideration and prioritization or for inclusion into an 

existing generic issue. 

19-1 
The AEOD also screens the recommendations and suggestions contained in its 

studies and evaluations for identification as potential generic issues. A gen- eric issue is an issue that is applicable to all, several, or a class of reactors 

or reactor-related facilities. Such issues are identified to RES, which then 

evaluates and prioritizes the issue in accordance with established procedures 

(see below). Generic issues may also be suggested by individuals within the NRC, 

the Advisory Committee on Reactor Safeguards (ACRS), the nuclear power industry, 

or the public. 

The generic issues management program comprises six distinct stages. In 

addition to the identification stage (discussed above), the stages are prioritization, resolution, imposition, implementation, and verification. Each new gen- eric safety issue (GSI) is prioritized by developing a quantitative assessment 

of safety benefits (risk reduction) and impact (cost) for the utility, the NRC, 

and any other entities involved, as described in NUREG-0933, 11 A Prioritization 

of Generic Safety Issues. 11 On the basis of the extent of potential risk reduction to the public and the value/impact ratio developed from this assessment, 

and as further adjusted by qualitative judgments, a priority is assigned to each 

GSI. Following peer review of the initial prioritization, a final priority is 

recommended and assigned. 

Issues that receive a high or medium priority are designated for resolution by 

the staff. An issue given a low or drop priority is, by nature of the rating standard, of so low a public risk reduction potential that resolution of the 

issue is not pursued. All issues are documented in the catalog of generic issues maintained in NUREG-0933. 

The resolution process requires the development of a plan and schedule for the 

work that needs to be done to resolve the issue. The plan also identifies 

needed resources and coordination points. Following completion of the technical studies, a final resolution package is prepared that includes a regulatory analysis describing various potential solutions and justification for any pro- posed requirements based on a consideration of value and impact. The resolution package is considered by the ACRS and by the Committee for Review of 

Generic Requirements (CRGR) if new requirements are proposed. Resolved issues 

are forwarded to NRR for imposition, implementation, and verification. This 

includes issuance of generic correspondence to licensees informing them of the 

issue resolution, establishment of an acceptable schedule for implementation of 

the resolution by the affected licensees, and verification that the required 

improvements have been made in an acceptable manner. 

Value impact analyses were employed as part of the basis of resolving some 

GSis. In the tradeoffs between net safety benefit and net cost, the remaining plant operating term ordinarily enters the calculations. Both the safety value 

and the cost impact can increase over time more than the cost impact, as would 

be the case when costs are largely one-time initial costs but the risk reduction benefit accumulates year after year with continued operation. Consequently, the consideration of extended plant life, and also any increase in population 

around nuclear plant sites, may alter the resolution bases of GSis that have 

been resolved but not backfitted. 

The staff performed a systematic evaluation of all GSis resolved through October 

1990 NUREG/CR 5382 to determine those whose resolution bases could be affected by 

an additional 20 years of plant life. A screening analysis was performed on 249 

19-2 
GSls that were resolved through October 1990, and 139 GSls were identified that 

did not result in backfit requirements. This total includes issues that were 

resolved without backfit and issues that were prioritized low and not considered 

further. Three GSls were identified in which the value-impact estimates played 

a relatively significant role in the resolution and the revised estimates were 

judged to warrant a further reconsideration of decision not to backfit. These 

issues are: 

• GSI-111.A.l.3(2) Maintain Supplies of Thyroid Blocking Agent 

• GSI-82 Beyond-Design-Basis Accident in Fuel Pools 

• GSI-101 BWR Water Level Redundancy 

Due to considerations aside from license renewal, GSI-III.A.1.3(2) has been 

returned to the issue resolution process and is currently under reevaluation. 

The resolution bases of GSis-82 and 101 were also reconsidered and the staff 

determined that the decision not to backfit is still appropriate. 

The screening analysis of issues that were originally prioritized in the low 

category indicated that four issues could be placed in the medium category. 

These issues are: 

• 

• 

• 

• 

GSI-II.D.2 

GSI-III.D.2.1 

GSI-35 

GSI-80 

Research on Relief and Safety Valve Test Requirements 

Radiological Monitoring of Effluents 

Degradation of Internal Appurtenances in LWRs 

Pipe Break Effects on Control Rod Drive Hydraulic Lines 

in the Drywells of BWR Mark I and II Containments 

The prioritization of the issues has been reconsidered and determined that 

GSis II.D.2, III.D.2.1, 35, and 80 should remain in the low category. 

The generic issues management program was initiated in 1981. At that time, 511 

issues were identified to be prioritized; 369 were TMI followup items (NUREGs0660 and -0737) and 142 were identified by previous assessments of generic issues 

(NUREGs-0371 and -0471). These issues included 22 issues (Appendix A) that had 

previously been identified as unresolved safety issues (USis). In the past 10 

years, an additional 264 issues have been identified, for a total of 775; this 

number includes various human factors issues and issues identified by the staff 

assessment of the Chernobyl accident. As of December 1990, 721 issues have been 

resolved, including all the USis. Of the remaining 54 issues, 26 are to be 

prioritized and 28 are in the resolution process. 

The implementation status of USis was recently reviewed by the NRR staff. NRR 1 s 

findings indicate that, in general, most USis have been implemented and that 

unimplemented USis are being addressed on a schedule satisfactory to the staff. 

The implementation status of the remaining generic safety issues is currently 

being assessed by NRR. 

19.4 Conclusions 

The NRC has an effective program in place for reviewing and analyzing operating 

experience and other new information, and for implementing any necessary modifications at operating reactors. The process allows for early notification of 

licensees of potential concerns, if deemed necessary, or for more thorough evaluation through the generic issues management program. Plant modifications are 

19-3 
implemented following an evaluation of various reasonable alternative solutions 

and justification based on an assessment of value and impact. The licensing basis for individual operating plants includes changes resulting from resolution 

of generic issues determined to be applicable. 

19-4 
Number 

A-1 

A-2 

A-3 

A-4 

A-5 

A-6 

A-7 

A-8 

A-9 

A-10 

A-11 

A-12 

APPENDIX A 

UNRESOLVED SAFETY ISSUES FOR WHICH A FINAL 

TECHNICAL RESOLUTION HAS BEEN ACHIEVED 

Title Report Number 

Water Hammer NUREG-0927, Rev. 1 

NUREG-0933 

Asymmetric Slowdown Loads NUREG-0609 

on Reactor Primary Coolant 

Systems 

Westinghouse Steam NUREG-0844 

Generator Tube Integrity 

CE Steam Generator Tube NUREG-0844 

Integrity 

B&W Steam Generator Tube NUREG-0844 

Integrity 

Mark I Short-Term Program NUREG-0408 

Mark I Long-Term Program NUREG-0661 

NUREG-0661 Suppl. 

Mark II Containment Pool NUREG-0808 

Dynamic Loads 

Anticipated Transients NUREG-0460, Vol. 4 

Without Scram 

BWR Feedwater Nozzle NUREG-0619 

Cracking 

Reactor Vessel Material NUREG-0744, Rev. 1 

Toughness 

Fracture Toughness of Steam NUREG-0577, Rev. 1 

Generator and Reactor Coolant Pump Supports 

1 

Date 

March 1984 

November 1980 

September 1988 

September 1988 

September 1988 

December 1977 

July 1980 

August 1981 

September 1980 

November 1980 

October 1982 

September 1982 
UNRESOLVED SAFETY ISSUES FOR WHICH A FINAL 

TECHNICAL RESOLUTION HAS BEEN ACHIEVED 

(continued) 

Number Title Report Number Date 

A-17 Systems Interact ions NUREG-1229 August 1989 

Generic Letter 89-18 

A-24 Qualification of Class IE NUREG-0588, Rev. 1 July 1981 

Safety-Related Equipment 

A-26 Reactor Vessel Pressure NUREG-0224 September 1978 

Transient Protection 

A-31 Residual Heat Removal SRP 5.4.7 1978 

Shutdown Requirements 

A-36 Control of Heavy Loads Near NUREG-0612 July 1980 

Spent Fuel 

A-39 Determination of SRV Pool NUREG-0802 September 1982 

Dynamic Loads and Pressure 

Transients 

A-40 Seismic Design Criteria NUREG-1233 September 1989 

A-42 Pipe Cracks in Boiling NUREG-0313, Rev. 1 July 1980 

Water Reactors 

A-43 Containment Emergency Sump NUREG-0897, Rev. 1 October 1985 

Performance 

A-44 Station Blackout Regulatory Guide 1.155 August 1988 

NUREG-1032 June 1988 

NUREG-1109 June 1988 

A-45 Shutdown Decay Heat Removal NUREG-1289 September 1988 

Requirements NUREG/CR-5230 

A-46 Seismic Qualification of NUREG-1030 February 1987 

Equipment in Operating NUREG-1211 

Plants 

A-47 Safety Implications of NUREG-1217 September 1989 

Control Systems NUREG-1218 

Generic Letter 89-19 

A-48 Hydrogen Control Measures NUREG-1370 September 1989 

and Effects of Hydrogen Burns on Safety Equipment 

A-49 Pressurized Thermal Shock Regulatory Guide 1.154 February 1987 

2 
NRC FORM 335 12-891 

NRCM 1102, 

3201, 3202 

2. TITLE AND SUBTITLE 

U.S. NUCLEAR REGULATORY COMMISSION 

BIBLIOGRAPHIC DATA SHEET 

/See instructions on the reverse) 

foundation for the Adequacy of the Licensing Bases 

A Sl:pplement to the Statement of Considerations for the 

Rule on Nuclear Power Plant License Renewal (10 CFR Part 54) 

Final Rernrt 

5. AUTHOR (S) 

1. REPORT NUMBER 

3. 

(Aulgned by NRC. Add Vol., Supp., Rev .. 

ond Addendum Numbers, If any.) 

NUREG-1412 

DATE REPORT PUBLISHED 

MONTH 1 YEAR 

December 1991 

4. FIN OR GRANT NUMBER 

6. TYPE OF REPORT 

Regulatory 

7. PERIOD COVERED /Inclusive Dates! 

B. PER FORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, provide 

name and mailing address.) 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC ·20555 

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (lfNRC, type "S.measabove";ifconrractor,provideNRCDivision, Office or Region, U.S. Nuclear Regulatory Comminion, 

•nd mailing •ddress. J 

Same as above 

10. SUPPLEMENTARY NOTES 

11, ABSTRACT (200 words or less) 

The objective of this report is to describe the regulatory processes that assures that 

any plant-specific licensing bases will provide reasonable assurance that the operation 

of nuclear power plants will not be inimical to the public health and safety to the end 

of the renewal period. It is on the adequacy of this process that the Commission has 

determined that a formal renewal licensing review against the full range of current 

safety requirements would not add significantly to safety and is not needed to assure 

that continued operation throughout the renewal term is not inimical to the public 

health and safety or common defense and security. This document illustrates in general 

terms how the regulatory process has evolved in major safety issue areas. It also 

provides examples illustrating why it is ,JI1I1ecessary to re-review an operating plant's 

basis, except for age-related degradation unique to license renewal, at the time of 

license renewal. The report is a supplement to the Statement of Considerations for the 

Nuclear Regulatory Commission's rule (10 CFR Part 54) that establishes the criteria 

and standards governing nuclear power plant license renewal. 

12. KEY WORDS/DESCR !PTORS (List words or phrases that will assist researchers in locating the repart,) 

Licensing Bases, License Renewal, Nuclear Power Plant 

NRC FORM 33S 12-891 

13. AVAILABILITY STATEMENT 

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(This Page) 

Unclassified 

/This Report) 

Uncla~1=;i -F; oM 

15. NUMBER OF PAGES 

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