ML19344F218

Proposed Licensing Requirements for Pending Applications for CPs & Mfg Licenses, Draft Rept for Comment

ML19344F218
Person / Time
Issue date:	08/31/1980
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0718, NUREG-0718-DRFT, NUREG-0718-DRFT-FC, NUREG-718, NUREG-718-DRFT, NUREG-718-DRFT-FC, NUDOCS 8009120628
Download: ML19344F218 (86)

Text

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NUREG-0718 l

Proposed Licensing Requirements for Pending Applications i

for Construction Permits and Manufacturing License Draft Report for Comment l

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$0teP7 ug t Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Wcshington, D.C. 20666 p ~%

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ABSTRACT The THI-2 Action Plan, NUREG-0660, does not specifically address requirements for construction permit and manufacturing license applications.

There are currently pending six construction permit applications for eleven plants and one manufacturing license application for eight floating nuclear plants.

Staff review of these applications has been suspended since the TMI-2 cccident pending the formulation of a policy to appropriately reflect the lessons learned from the accident.

This report summarizes an NRC staff assessment of the TMI-2 Action Plan and the manner and extent to which it should be applied to these seven applications prior to the issuance of any licenses.

TABLE OF CONTENTS Section Title M

Abstract iii I

Introduction 1

4 II Discussion 4

III Assessment of TMI-2 Action Plan For Pending CP and ML Applications 6

IV Summary and Conclusions 9

Appendix A ACRS Memorandum, Chairman Plesset to Chairman Ahearne Dated May 6, 1980 A-1 Appendix 8 Requirement Category Assignments for Pending Construction Permit and Manufacturing License Applications B-1 Appendix C Action Plan Items Applicable to Pending Construction Permit and Manufacturing License Applications C-1 l

Appendix D Information Requirements for TMI-2 Action Plan Items in Categories 3, 4, and 5 D-1 l

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a I.

INTL!0 DUCTION After the accident at Three Mile Island, Unit 2, on March 28, 1979, the Commission directed its technical review resources to assuring the safety of operating power reactors rather than to the issuance of new licenses.

Further-more, the Commission decided that power reactor licensing should not continue until the assessment of that accident had been substantially completed and comprehensive improvements in both the operation and regulation of nuclear power plants had been set in motion.

Following the accident at Three Mile Island, Unit 2, the President established a Commission to make reccimendations regarding changes necessary to improve nuclear safety.

In May 1979, the Nuclear Regulatory Commission established a Lessons Learned Task Force to determine what actions were required for new operating licenses and chartered a Special Inquiry Group to examine all facets of the accident and its causes.

These groups have published their reports.

The Lessons Learned Task Force led to NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" and NUREG-0585, "TMI-2 Lessons Learned Task Force Final Report." Following release of the report of the Presidential Commission, the Commission provided a preliminary set of l

responses to the recommendations in that report.

This response provided broad l

policy directions for development of an NRC Action Plan, work on which was begun in November 1979.

During the development of the Action Plan, the i

Special Inquiry Group Report was received, which had the benefit of review by panels of outside consultants representing a cross section of technical and public views.

This report provided additional recommendations.

NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident," was developed to provide a comprehensive and integrated plan for the actions

. Judged appropriate by the Nuclear Regulatory Commission to correct or improve the regulation and operation of nuclear facilities based on the experience from the accident at Three Mile Island, Unit 2, and the official studies and investigations of-the accident.

In developing the Action Plan, the various

V recommendations and possible actions of all the principal investigations were assessed and either rejected, adopted or modified.

Actions to improve the safety of nuclear power plants now operating were judged to be necessary imediately after the accident and could not be delayed until the Action Plan was developed, although they were subsequently included in the Action Plan.

Such actions came from the Bulletins and Orders issued imediately after the accident, the first report of the Lessons Learned Task Force issued in July 1979, the recomendations of the Emergency Preparedness Task Force, and the NRC staff and Commission. Before these imediate actions were applied to operating plants, they were approved by the Commission.

Many of the required imediate,ctions have already been taken by licensees and most are scheduled te be complete by the end of 1980.

On February 7,1980, based on its review of initial drafts of the Action Plan, the Commission approved a listing of near-ters operating license (NTOL) requirements, as being necessary but not necessarily sufficient TMI-related requirements, for granting new operating licenses.

Since then, the fuel load requirements on the NTOL list have been used by the Commission in granting operating licenses, with limited authorizations for fuel loading and low power testing, for three plants.

On May 15, 1980, after review of the last version of the Action Plan, the Comission approved a list of " Requirements For New Operating Licenses," now contained in NUREG-0694, which the staff recommended for imposition on current operating license applicants.

That list was recast from the previous NTOL list and sets forth the TMI-related requirements and actions for new operating licenses. The Comission also approved the staff's recommendation that the remaining items from the TMI reviews should be implemented or considered over time to further enhance safety.

In approving the schedules for developing and implementing changes in requirements, thc Comission's primary considerations were the safety signif-icance of the issues and the imediacy of the need for corrective actions.

As discussed above, many actions were taken to improve safety immediately or soon after the accident.

These actions were generally considered to be interim 2

l improvements.

In scheduling the remaining improvements, the availability of both NRC and. industry resources was considered, as well as the safety signif-icance of the actions.

Thus, the Action Plan approved by the Commission presents a sequence of actions that will result in a gradually increasing improvement in safety as individual actions are completed and the initial immediate actions are replaced or supplemented by longer term improvements.

Based upon its review and consideration of the issues arising as a result of the Three Mile Island accident, the Commission, in its June 16, 1980 Statement of Policy, concluded that the above-mentioned list of TMI-related requirements for new operating licenses found in NUREG-0694 is necessary and sufficient for responding to the TMI-2 accident.

The Commission has decided that current operating license app'ications should be measured against the regulations, as augmented by these requirements.

The Commission also stated that, in general, the remaining items of the Action Plan should be addressed through the normal process for development and adoption of new requirements rather than through immediate imposition on pending applications.

The THI-2 Action Plan, NUREG-0660 does not specifically address requirements for const'.'uction permits (CP) or manufacturing license (ML) applications.

There are currently pending six CP applications for eleven plants and one ML application for eight floating nuclear plants.

The NRC staff review of these applications has been suspended since the TMI-2 accident pending the formula-tion of a licensing policy to appropriately reflect the lessons learned from the accident.

Therefore the NRC staff initiated a program to propose for Commission approval a course of action that would lead to the establishment or

- TMI-2 related requirements for these applications.

The sections that follow describe the NRC staff review and the proposal for addressing TMI-related requirements on these applications.

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Y II.

DISCUSSION On March 17, 1980 the Office of Nuclear Reactor Regulation formed a Task Group to propose requirements based on the TMI-2 Action Plan for the six pending applications for eleven construction permits.

This effort was later expanded to include the one pending application for a manufacturing license for eight floating nuclear plants.

The applicants for the six pending CP applications formed a group to interact with the staff in the development of the requirements.

A number of meetings were held with the applicants to discuss the NRC staff's program for establishing the requirements.

One meeting was held with an Advisory Committee on Reactor Safeguards (ACRS) Subcommittee and one was held with the Full Committee to discuss the NRC staff's program.

An ACRS letter dated May 6, 1980 from Chairman Plesset to Chairman Ahearne is included as Appendix A.

This letter indicates that the ACRS expects further involveaent in the development of the requirements.

In establishing the program we considered three options:

1.

Resume licensing using the pre-TMI CP requirements augmented by the applicable requirements identified in the Commission's June 16, 1980 Statement of Policy regarding operating licenses.

In effect, this treats the pending CP and ML applications as if they wert the last of tne present generat. ion of nuclear power plants.

2.

Take no further action on the pending CP and ML applications until the rulemaking actions described in the Action Plan have been completed.

This would, in effect, treat the pending applications as the first of a new generation of nuclear power plants.

3.

Resume licensing using the pre-TMI CP and ML requirements augmented by the applicable requirements identi~fied in the Commission's June 16, 1980 Statement of Policy regarding operating licenses and require certain 4

9 additional measures or commitments in related areas, e.g., those that will be the subject of rulemaking.

Option 1 would minimize the review and construction impact, thereby minimizing delays in reaching regulatory decisions on these applications.

The principal disadvantage of Opticn 1 is that it fails to take advantage of the fact that, since construction has not started, it would be relatively easy to provide design flexibility to implement potential significant safety improvements.

Option 2 would maximize the safety improvements but would result'in extensive delays and possible cancellations. We believe that the costs of such delays are not justified provided that design flexibility can be demonstrated.

Option 3 is believed to be a suitable compromise between the extremes of Options 1 and 2.

This option will ensure that approved action items in the Action Plan are applied to the pending CP and ML applications and will provide for early consideration of added safety measures that can be incorporated into the design without the need for inordinately costly backfit.

By establishing a clear statement of policy with respect to the issues to be determined by rulemaking, a degree of stability is introduced into the review process thereby allowing prospective applicants to make better-informed decisions.

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7 III.

ASSESSMENT OF TMI-2 ACTION PLAN FOR PENDING CP AND ML APPLICATIONS In order to assess the extent to which the THI-2 Action Plan should be imple-mented on the seven pending CP and ML applications, the staff developed five requirement categories.

Each of the THI-2 Action Plan requirements were carefully evaluated and then assigned to one of these five categories. A discussion of each of the requirement categories follows.

Category 1 Information of a type not applicable to the pending CP and ML applications for any of the following reasons:

a.

It can only be addressed in operating license applications or by licensees; b.

It is not directed to CP or ML applicants; c.

It does not apply to plants of the type now in review; d.

It has been (or will be) superseded by a more restrictive requirement in the Action Plan; e.

It has already been completed.

Category 2 Information of the type customarily left for the operating license stage.

The applicant should indicate its recognition of the development of operating license or final design requirements and should provide a commitment to imple-ment such requirements in connection with its application for approval of the final design.

Category 3 Studies (and other research and development activities to provide design development information) of the type customarily left for review at the final 6

q stage.

However, to satisfy 50.35(a)(3) the staff believes that items in this category should be completed as early as is practicable so that the results can be most effectively taken into account in developing final design details.

The applicant should provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and a program to assure that the results of such studies are factored into the final design.

Category 4 Information sufficient to demonstrate that any additional design, development and implementation necessary to satisfy the requirement (or to satisfy the goals of the task whose requirements are to be developed in the future) will be satisfactorily completed by the operating license stage.

This is the type of information customarily supplied at the construction permit stage to satisfy 50.35(a)(2), or is similar to the information customarily supplied at the pre-liminary design stage with respect to generic issues to satisfy ALAB-444.

Category 5 Information of the type customarily reviewed at the preliminary design stage to include:

Items for which the required information should be sufficient to demon-a.

strate that the requirement has been satisfied by the application.

This is the kind of information and degree of detail customarily provided at the preliminary design stage with respect to site and major systems and structures to satisfy 50.34(a)(1).

This will also be applicable to items relating to technical qualifications of the applicant and its management for design and construction.

b.

Items for which the required information should be sufficient to assure that the requirement will be met at the final design stage.

This is the kind of information and degree of detail customarily provided at the pre-liminary design stage with respect to the preliminary design of the facility to satisfy 50.34(a)(3)(4), etc.

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It should be noted that in assigning indiv'idual TMI-2 Action Plan requirements to one of tt.e aforementioned categories, the NRC staff did not limit itself to a narrow heterpretation of the Action Plan requirements.

Rather, we took into account the fact that much work could be done by the applicants to address the l

specific concerns of the individual Action Plans.

In such cases we defined the specific concerns that should be addressed by the CP and ML applicants and the level of information to be supplied in order for the staff to verify that the requirement has been (or will be) satisfied.

Tables 1, C.1, C.2, and C.3 from NUREG-0660 list each of the TMI-2 Action Plan requirements.

Appendix B of this report is a reprint of these tables with the NRC staff's category assignments for the pending CP and ML applications.

Appendix C provides a cross-cut of the Action Plan requirements by category assignment for the pending CP and ML applications. Appent'x D provides a description of the specific information to be provided by CP a:id ML applicants for each of the Action Plan requirements assigned to Category 3, 4, and 5.

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IV.

SUMMARY

AND CONCLUSIONS Based upon its extensive review and consideration of the issues arising as a resu!t of the Three Mile Island accident, the Commission recently approved the TMI-2 Action Plan, NUREG-0660.

The Comission noted that the Action Plan presents a sequence of actions that will result in a gradually increasing improvement in safety as individual actions are completed and the initial immediate actions that were taken soon after the accident are replaced or supplemented by longer term improvements.

By Policy Statement dated June 16, 1980, the Commission identified (in NUREG-0694) the set of TMI-2-related requirements for new operating licenses that are necessary and sufficient for responding to the TMI-2 accident.

The Commission further decided that current operating license applications should be measured against the regulations, as augmented by these requirements.

The NRC staff has developed a proposal with respect to the set of necessary and sufficient TMI-related requirements that should be applied in the review of applications for construction permits and manufacturing licenses for nuclear power plants as described in this report.

Upon receipt of public comments and further review by the ACRS, we would plan to finalize our position and present l

appropriate recommendations for Comission approval to resume review of the construction permit and manufacturing license applications.

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ut.iito s1 AT Es APPENDIX A NUCLEAR REGUL ATORY CO.'.*. MISSION j.; k..

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ADVISORY COf.f.* lite E Ot; RE ACTOR SAFEGUARDS e, ',>.;3Nf f

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p May 6, 1980 Honorable John F. Ahearne Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

NEAR-TERM CONSTRUCTION PERMIT APPLICATIONS

Dear Dr. Ahearne:

Ouring its 241st meeting, May 1-3, 1980, the ACRS reviewed the status of applications for near-term construction permits (NTCPs).

In its review the Committee had the benefit.of discussions with the NRC Staff and with repre-sentatives of the applicants for the NTCPs.

A subcomittee meeting on this subject was held on April 9, 1980.

The six NTCP applicants and the reactor types involved are as follows:

Black Fox Station, Units 1 and 2, Public Service Company of Oklahoma, General Electric BWR/6, Mark III pressure suppres-sion containment Skagit Nuclear Power Project, Units 1 and 2, Puget Sound Pcwer

& Light Company, General Electric BWR/6, Mark III pressure suppression containment Pilgrim Station, Unit 2, Boston Edison Company, Combustion Engineering custom NSSS, large dry containment Perkins Nuclear Station, Units 1, 2 and 3, Duke Power Company, Combustion Engineering CESSAR System 80 NSSS, large dry con-tainment Allens Creek Nuclear Generating Station, Houston Lighting &

Power Company, General Electric BWR/6, Mark III pressure sup-pression containment Pebble Springs Nuclear Plant, Units 1 and 2, Portland General Electric Company, Babcock and Wilcox custom NSSS, large dry containment The NRC Staff has approached this matter primarily by examining the Action Plan and judging the applicability and scheduling of each item to an NTCP.

This procedure has resulted in placing mmy important items in a category wherein the NRC has yet to develop critena applicable to construction per-mit applicants. - Action Plan item II. A on siting introduces questions whose resolution must be achieved prior to issuance of a construction permit.

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'!anarable John F. Ahearne May 6, 1980 item II.B on degraded or melted cores bears directly on containment design, as well as other safety features.

Item II.C on reliability engineering and risk assessment could bear significantly on the design requirements for cany important plant systems.

There are many other, items in the Action Plan and in the ACRS report of April 17, 1980 which also might impact directly on im-portant design aspects of these plants.

Mr. Harold Denton advised the Comittee that he envisaged permitting con-struction to proceed if there are no obvious site-r? lated of the Report of the Siting Policy Task Force (NUREG-0625) questions in terms and if the contain-ment design pressure were such as to withstand hydrogen combustion, on the assumption that other design aspects could be changed later if so requiret.

The utility representatives advised the ACRS that, in their opinion, there was a need for the resolution of several policy questions which relate to how and whether construction permit applications will be processed in the near term.

The utilities identifled the following six policy issues as being in most urgent need of resolution:

1.

Siting 2.

Emergency planning 3.

Degraded core conditions 4.

Control room design 5.

!!anagement for design and construction 6.

Reliability and risk assessment The utility representatives recommended that a concerted effort be under-taken to develop an acceptable interim approach to resolution by the Commis-sion of such issues in the next few months.

The ACRS supports this recom-mendation and urges that appropriate Staff resources be made available for this purpose.

An ACRS Subcommittee plans to work actively with the Staff on the topic with the anticipation that the full Committee would review the NTCP matter within a few months.

Sincerely, Milton S. Plesset Chairman

References:

1.

Memorandum from D. F. Ross, NRC, to R. F. Fraley, ACRS,

Subject:

Trans-mittal of NTCP Requirements List, dated April 22, 1980.

2.

Memorandum from William F. Kane, NRC, to Addressees,

Subject:

Request for Review of Proposed TMI-2-Related Requirements for NTCP Applicants, dated April 4, 1980.

3.

U. S. Nuclear Regulatory Commission, "NRC Action Plans Developed as a Result of the TMI-2 Accident," USNRC Report NUREG-0660 Draft 3, dated March 5, 1980.

4.

U. S. Nuclear Regulatory Commission, " Report of the Siting Policy Task Force " USNRC Report NUREG-0625, dated August,1979.

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APPENDIX B REQUIREMENT CATEGORY ASSIGNMENTS FCf! PENCING CONSTRUCTION PERMIT AND MANUFACTURING LICENSE APPLICATIONS TABLE 1 - PRIORITIES AND STATUS OF ITEMS IN TMI-2 ACTION PLAN REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML I.

Operational Safety I.A Operating Personnel I.A.1 Operating Personnel and Staffing

1. Shift Technical Advisor 2/lc Refer to Appendix D

2. Shift Supervisor Admin.

2/lc Refer to Appendix D Duties

3. Shift Manning 2/lc Refer to Appendix D

4. Long-term Upgrading id/lc Refer to Action Plan Item I.B.l.1 I.A.2 Training and Qualifications of Operating Personnel

1. Inmediate Upgrading of id/lc Refer to Action Plan Item I.B.l.1 Operating and Senior Opera-tor Training and Qualifi-cations

2. Training and Qualifications of Id/lc Refe,' to Action Plan Item I.B.l.1 Operations Personnel

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TABLE 1 (Continued)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML

3. Administration of Training ~

lb/lc Programs for Licensed Operators

4. NRR Participation in IE lb/lc Inspector Training

5. Plant Drills 2/lc

6. Long-Term Upgrading of Training id/l :

Refer to Action Plan Item I.B.1.1 and Qualifications

7. Accreditation of Training lb/lc Institutions b

I.A.3 Licensing and Requalification of Operating Personnel 1.. Revise Scope and Criteria for 2/lc Refer to Action Plan Item I A.3.2 Licensing Exams

2. Operator Licensing Program Changes Ib/lc

3. Requirements for Operator lb/lc Fitness j

4. Licensing of Additional lb/lc Operations Personnel

5. Establish Statement of lb/lc Understanding with INP0 and DOE

TABLE 1 (Continued)

REQUIREENT CATEGORY ACTION ITEM ASSIGNMENT COMENTS CP/ML I.A.4 Simulator Use and Development

1. Initial Simulator Improve-1d /lc Refer to Action Plan Item I.A.4.2 ment

2. Long-Term Training Simulator 4/lc Refer to Appendix D Upgrade

3. Feasibility Study of Procure-lb /lb ment of NRC Training Simulator

4. Feasibility Study of NRC lb /lb Engineering Computer cn la I.B Support Personnel I.B.1 Management for Operations

1. Organization and Management 55 Refer to Appendix D Long-Term Improvements

2. Evaluation of Organization 1d /lc Refer to Action Plan Item I.B.1.1 4

and Management Improvements of NTOL Applicants

3. Loss of Safety function 1b /lc I.B.2 Inspection of Operating Reactors

1. Revise IE Inspection Program Ib /lb

2. Resident Inspector at Operating lb /lb Reactors l

TABLE 1 -(Continued)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML

3. Regiona'l Evaluations lb/lb

4. Overview of Licensee lb/lb Performance I.C Operating Procedures

1. Short-Term Accident Anal-4/4 Refer to Appendix D ysis and Procedures Revision

2. Shift and Relief Turnover 2/lc Procedures 1.

3. Shift Supervisor Responsi-2 /1c bilities

4. Control Room Access 2 /lc

5. Procedures for Feedback of 4 /4 Refer to Appendix 0 Operating Expertence

6. Procedures for Verification of 2 /lc Refer to Appendix D Correct Performance of Operating Activities

7. NSSS Vendor Review of Pro-2 /lc cedures

8. Pilot' Monitoring of Selected lb /lc Emergency Procedures for NT0L Applicants

9. Long-Term Program Plan for 4/lc Refer to Appendix D Upgrading of Procedures

TABLE 1 (Continu d)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML I.D Control Room Design

1. Control Room Design Reviews 4/4 Refer to Appendix D

2. Plant Safety Parameter 4/4 Refer to Appendix D Display Console

3. Safety System Status Monitor-4/4 Refer 'to Appendix D ing

4. Control Room Design Standard 4/4 Refer to lippendix D

5. Improved Control Poom lb/lb w

'n Instrumentation Research

6. Technology Transfer Conference 1b/lb I.E.

Analysis and Dissemination of Operating Experience

1. Office for Analysis and lb/lb Evaluation of Operational Data

2. Program Office Operational Data lb/lb Activities

3. Operational Safety Data lb/lb Analysis

4. Coordination of Licensee, Industry, 4/4 Refer to Appendix 0 and Regulatory Programs

i TABLE 1.(Continued)

REQUIREMFNT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML

5. Nuclear. Plant Reltability lb/lb Data System

6. Reporting Requirements lb/lb

7. Foreign Sources

' lb/l b

8. Human Error Rate Analysis lb/lb I.F Quality Assurance.

1. Expand QA List 4/4 Refer to 4ppendix D

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2. Develop More Detailed QA Criteria 4/4 Refer to appendix D I.G Preoperational and Low-Power Testing

1. Training Requirements ib/lc

2. Scope of Test Program lb/lc II.

Siting and Design II.A Siting

1. Siting Policy Reformulation lb/lc

2. Site Evaluation of Existing 5/lc Refer to Appendix D Facilities

TABLE 1 (Continu d)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML II.B Consideration of Degraded or Melted Cores in Safety Review

1. Reactor Coolant System Vents 4/4 Refer to Appendix D

2. Plant Shielding to Provide 4/4 Refer to Appendix D Access to Vital Areas and Protect Safety Equipment for Post-accident Operation

3. Post-accident Sampling 4/4 Refer to Appendix D

4. Training for Mitigating Core 2/lc Damage O

5. Research on Phenomena lb /lb Associated with Core De-gradation and Fuel Melting

6. Risk Reduction for Operating ic /lc -

Reactors' at Sites with High Population Densities

7. Analysis of Hydrogen Control id /ld Refer to Action Plan Item II.B.8

8. Rulemaking Proceeding on Degraded 5 /5 Refer to Appendix D Core Accidents II.C Reliability Engineering and Risk Assessment

l. Interim Reliability Evaluation lb/l b Program (IREP)

2. Continuation of IREP lb/lb

TABLE 1 (Continued)

REQUIREMENT CATEGORY COMMENTS ACTION ITEM ASS}GNMENT j

3. Systems Interaction lc /lc

4. Reliability Engineering 3/3 Refer to Appendix 0 II.D Reactor Coolant System Relief and Safety Valves

1. Testing Requirements 4 /4 Refer to Appendix D

2. Research on Relief and Safety 3 /3 Refer to Appendix D Valve Test Requirements

3. Relief and Safety Valve Posi-4/4 Refer to Appendix 0 tion Indication

{

II.E System Design II.E.1 Auxiliary Feedwater System

1. Auxiliary Feedwater System 3/3 Refer to Appendix 0 Evaluation

2. Auxiliary Feedwater System 4/4 Refer to Appendix 0 Automatic Initiation and Flow Indication

3. Update Standard Review Plan lb/lb and Develop Regulatory Guide II.E.2 Emergency Core. Cooling System

1. Reliance on ECCS 2/2

2. Research on Small Break LOCAs lb/lb and Anomalous Transients

3. Uncertainties in Performance 3/3 Refer ~ to Appendix D Predictions

TABLE 1 -(Continued)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML II.E.3 Decay Heat Removal

1. Reliability of Power 4/4 Refer to Appendix D Supplies for Natural Circulation

2. Systems Reliability lb/lb

3. Coordinated Study of Id/ld Refer to Action Plan II.C.4 Shutdown Heat Removal

4. Alternate Concepts Research lb/lb

5. Regulatory Guide lb/lb II.E.4 Containment Design

1. Dedicated Penetrations 5/5 Refer to Appendix D

2. Isolation Dependability 4/4 Refer to Appendix U

3. Integrity Check 2/lc

4. Purging 4/4 Refer to Appendix D II.E.5 Design Sensitivity of B&W Reactors

1. Design Evaluation 4/lc Refer to Appendix D

2. B&W Reactor Transient Response 4/lc Refer to Appendix D Task Force II.E.6 In-situ Testing of Valves

1. Test Adequacy Study ib/lb l

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TABLE 1 (Continutd)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML i

i II.F Instrumentation and Control

1. Additional Accident Monitor-4/4 Refer to Appendix 0 ing Instrumentation

2. Identification of and Recovery from 4/4 Refer to Appendix 0 Conditions Leading to Inadequate Core Cooling

3. Instrumentation for honitor-44 Refer to appendix D

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ing Accident Conditions (Reg. Guide 1.97)

4. Study of Control and Protective 1b/lb co g

Action Design Requirements II.G Electrical Power ^

1. Power Supplies for Pressuri-4/4 Refer to Appendix D zed Relief Valves, Block Valves, and Level Indicators II.H TMI-2 Cleanup ar.d Examination i

1. Maintain Safety of TMI-2 and lb/lb Minimize Environmental Impact-

2. Obtain Technical Data on the 1b/lb Conditions Inside the TMI-2 Containment Structure

3. Evaluate'and Feedback lb/lb Information Obtained from TMI

4. Determine Impact of TMI on 1b /lb Socioeconomic-and Real Property

' Values

TABLE 1 (Continued) l REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML II.J General Implications of TMI for Design and Construction Activities II.J.1 Vendor Inspection Program

1. Establish a Priority System for lb/lb Conducting Vendor Inspections

2. Modify Existing Vendor Inspection Ib/lb I

Program

3. Increase Regulatory Control lb/lb l

7 Over Present Non-licenses C

4. Assign Resident Inspectors to lb/lb r

Vendors and Architect-Engineers II.J.2 Construction Inspection Program

1. Reorient Inspection Program

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2. Increase Emphasis on Independent

-lb/lb Measurement in the Construction Inspection Program

3. Assign Resident Inspectors to all lb/lb Construction Sites I

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TABLE 1 (Continued)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML II.J.3 Hanagement for Design and Construction

1. Organization and Staffing to 5/5 Refer to Apperdix 0 Oversee Design and Construction

2. Issue Regulatory Guide Ib/lb

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II.J.4 Revise Deficiency Reporting Requirements

1. Revise Deficiency Reporting lb/lb a

g Requirements II.K Measures to Mitigate Small-Break LOCAs and Loss of Feedwater Accidents

1. IE Bulletins See Table C.1

2. Commission Orders on B&W plants See Table C.2

3. Final Recommendations of B&O See Table C.3 Task Force l

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TABLE 1. (Continued)

REQUIREMENT CATEGORY

-ACTION ITEM ASSIGNMENT COMMENTS CP/ML III.

Emergency Preparations and l

Radiation Effects III.A NRC and Licensee Preparedness III.A.1 Improve Licensee Emergency Preparedness - Short-term

1. Upgrade Emergency Prepared-5/lc Refer to Appendix D ness

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2. Upgrade Licensee Emergency 4/4 Refer to Appendix D C

Support Facilities

3. Maintain Supplies of Thyroid 2/lc Applies to workers only. Category lb ior oublic.

' Blocking Agent (Potassium Iodide)

III.A.2 Improving Licensee Emergency l

Preparedness - Long-term i

1. Amend 10 CFR 50.and 10 CFR 5/lc

. Refer to Appendix D 50, Appendix E

2. Development of Guidance and 5/5 Refer to Appendix D Criteria III.A.3' Improving NRC Emergency Preparedness

1. NRC Role in Responding to lb/lb

~ Nuclear Emergencies i

2. Improve Operations Centers 1b/lb

3. Communications 4/le' Refer to Appendix D

't TABLE 1 (Continued)

'I REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML

4. Nuclear Data Link lb/ib 5.- Training,. Drills, and Tests 2/lc

6. Interaction of NRC with Other 1b/lb Agencies III.B Emergency Preparedness of State and Local Governments

1. Transfer of Responsibilities 1b/lb T

to FEMA 5

2. Implementation of NRC's and lb/lb FEMA's Responsibilities III.C - Public Information

1. Have Information Available 1b/lb for the News Media and the Public

2. The Office of Pubite Affairs 1b/lb will Develop Agency Policy and Provide Training for Interfacing with the News Media and Other Interested Parties

TABLE 1 (Continu:d) i REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML VI.D Radiation Protection l

III.D.1 Radiation Source Control

1. Primary Coolant Sources 4/4 Refer to Appendix D Outside the Containment Structure

2. Radioactive Gas Management 4/4 Refer to Appendix D

3. Ventilation System and 4/4 Refer to Appendix D Radiciodine Adsorber Criteria

4. Radwast'e System Design lb/lb os A

Features to Aid in Accident Recovery and Decontamination III.D.2 Public Radiation Protection Improvement

1. Radiological Monitoring of lb/l b Effluents

2. Radiciodine, Carbon-14, and lb /lb Tritium Pathway Dose Analysis

3. Liquid Pathway Radiological 2 /2' Control

4. Offsite Dose Measurements 2 /lc

5. Offsite Dose Calculation Manual 2 /lb

6. Independent Radiological ib /lb Measurements

TABLE 1 (Continued)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS' CP/ML III.D.3 Worker Radiation Protection Improvements

1. Radiation Protection Plans 4/4 Refer to Appendix D

2. Health Physics Improvements lb/lb

3. Inplant Radiation Monitoring 4/4 Refer to Appendix D

4. Control Room Habitability 4/4 Refer to Appendix D

5. Radiation Worker Exposure vata la/lc J.

Base o.

IV.

Practices and Procedures IV.A Strengthen Enforcement Process

1. Seek Legislative Authority lb/lb

2. Revise Enforcement Policy lb/lb IV.B Issuance of Instructions and Information to Licensees IV.B.1 Revise Practices for Issuance of lb/lb Instructions and Information to Licensees IV.C Extend Lessons Learned to Licensed Activities Other than Power Reactors IV.C.1 Extend Lessons Learned from TMI to lb/lb other NRC Programs IV.D NRC Staff Training 10/lb IV.D.1 NRC Staff Training

TABLE 1 (Continued)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML IV.E Safety Decision-Making

1. Expand Research on lb/lb Quantification of Safety Decision-Making

2. Plan for Early Resolution of lb/lb

. Safety Issues

3. Plan for Resolving Issues at lb/lb Construction Permit Stage

.L -

4. Resolve Generic Issues by lb/lb Rulemaking

5. Assess Currently Operating lb/lb Reactors IV.F Financial Disincentive to Safety

1. Increased IE Scrutiny of Power lb/lb Ascension Test Program

2. Evaluate the Impact of Financial lb/lb Disincentives to the Safety of Nuclear Power Plants IV.G Improve Safety Rulemaking Procedures

1. Develop a Public Agenda for lb/lb Rulemaking

2. Periodic and Systematic lb/lb Reevaluation of Existing Fules

3. Improve Rulemaking Procedures 1b/lb

TABLE 1 (Continued)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML

4. Study Alternative for lb/lb Improved Rulemaking Process IV.H NRC Participation in the Radi-ation Policy Council V.

NRC Policy, Organization and Management 1.

Develop.NRC Policy Statement lb/lb

-on Safety

2. Study Elimination on Non-safety lb/lb Responsibilities 5

3. Strengthen Role of ACRS lb/lb

4. Study Need for Additional lb/lb Advisory Comittees

5. Improve Public and Intervenor lb/lb Participation in Hearing Process

6. Study Construction-During-lb/lb Adjudication Rules

7. Study Need for TMI-Related lb/lb Legistration

8. Study the Need to Establish lb/lb an Independent Nuclear Safety Board

9. Study t.be Reform of the Licensing lb/lb Process 1

TABLE 1 (Continued)

REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML

10. Study NRC Top Management lb/lb Structure and Process

11. Reexamine Organization and lb/lb Functions of NRC Offices

12. Revise Delegations of lb/lb

. Authority to Staff

13. Clarify and Strengthen the 1b/lb Respective Roles of Chairman, Commission, and EDO m

U

14. Authority to Delegate Emergency lb/lb Response Functions to a Single Commissioner

15. Achieve Single Location -

lb/lb Long-term

16. Achieve Single Location - Interim lb/lb

17. Reexamine Comission Role in lb/lb Adjudication

3 TABLE C.1 0FFICE OF INSPECTION ANO ENFORCEMENT BULLETINS Req /ML CP uirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 1.

Review TMI-2 PNs and detailed chronology 79-05&05A (Item 1)

BWR and PWR l'd/l d Refer to Action Plan of the TMI-2 accident.

79-06&O6A (Item 1)

Items I.A.2.2 and 79-06&O68 (Item 1)

I.A.3.1 79-08 (Item 1) 2.

Review transients similar to THI-2 that 79-05&05A (Item 2)

• B&W ld/lc Refer to Action Plan have occurred at other facilities and Items I.A.2.2 and NRC ovaluation of Davis-Besse transient.

I.A.3.1

[, 3.

Review operating procedures for recog-79-05&05A (Item 3)

PWR ld/lc Refer to Action Plan c:

nizing, preventing, and mitigating void 79-06&O6A (Item 2)

Item I.C.1 formation in transients and accidents.

79-06&O68 (Item 2) 4.

Review operating procedures and training 79-05&05A (Item 4.a) PWR and BWR ld'/l c Refer to Action Plan instructions to ensure that:

79-05B (Item 2)

Items I.C.1, I.C.7,79-06A (Item 7.a)

I.C.8, and I.G.1 a.

Operators to not override ESF 79-06B (Item 6.a) actions unless continued operation 79-08 (Item 5.a) is unsafe; b.

HPI system operation NUREG-0645 (App. G)

W, CE ld/lc Refer to Action Plan HUREG-0565 B&W Item I.C.1 (Rec. 104)69-110 6002-00 ANO-1 (11/1/79 69-110 6003-00 Davis-Besse 1 (11/20/79)69-110 6001-00 Oconee 1, 2 & 3 (11/1/79)

Crystal River 3 Rancho Seco 1

TABLE C.1 (continued)

LP/ML Requirement Source for Category Reqrirement Operating Reactors Applicability Assignment Comments c.

RCP operation NUREG-0623 PWR ld/lc Refer to Action Plan Items I.A.l.3 and I.C.1 d.

Operators are instructed not to rely 79-05A (Item 4.d)

PWR and CWR ld/lc Refer to Action Plan on level indication alone in 79-06A (Item 7.d)

Items I.C.1, I.A.3.1, and evaluating plant conditions.79-068 (Item 6.d)

II.F.2 79-08 (Item 5.b)

T'5.

Safety-related valve position.

73-35&05A (Item 5)

PWR and BWR id/ld Refer to Action Plan O'

Items I.C.2 and I.C.6 a.

Review all valve positions and 79-06A (Item 8) positioning requirements and positive 79-06B (Item 7) controls and all related test and 79-008 (Item 6) maintenance procedures to assure proper ESF functioning, if required.

b.

Verify that AFW valves are in open 79-05A (Item 5)

B&W id/ld Refer to Action Plan position..See Requirement 8 below.

Items I.C.2 and I.C.6 6.

Review containment isolation initiation 79/05A (Item 6)

PWR and BWR ld/ld Refer to Action Plan design and procedures. Assure isolation 79-06A (Item 4)

Item II.E.4.2 of all lines that do not degrade safety 79-06B (Item 3) features or coolin0 capability upon 79-08 (Item 2) automatic initiation of SI.

7.

Implement positive position controls on 79-05A (Item 7)

B&W id/lc Refer to Action Plan valves that could compromise or defeat Item II.E.1.1 AFW flow.

a TABLE C.1 (continued)

CP/ML Requirement Source for Category Requirement Operating Reactors Applicability Assignment Consnents 8.

Immediately implement procedures that 79-05A (Item 8)

B&W id/lc Refer to Action Plan assure two independent 100% AFW flow paths, Item II.E.1.1 or specify explicitly LCO with reduced AFW capacity.

9.

Peview prncedures to assure that radio-79-05A (Item 9)

PWR and BWR ld/lc Refer to Action Plan active liquids and gases are not trans-79-06A (Item 9)

Item II.E.4.2 ferred out of containment inadvertently 79-068 (Item 8) especially upon ESF reset).

L'..t all 79-08 (Item 7) applicable systems and interl.scks.

m 10.

Review and modify (as required)79-05A (Item 10)

PWR and BWR Id/lc Refer to Action Plan procedures for removing safety-79-06A (Item 10)

Items I.C.2 and I.C.6 related systems from service (and 79-068 (Item 9) restoring to service) to assure 79-08 (Item 8) operability status is known.

11.

Make all operating and maintenance 79-05A (Item 11)

PWR and BNR ld/lc Refer to Action Plan personnel aware of the seriousness79-06A (Item 1.a)

Items I.A.3.1 and and consequences of the erroneous79-06B (Item 1.a)

I.A.2.2 actions taken leading up to, and in 79-08 (Item 1.a) early phases of, the THI-2 accident.

12.

One hour notification requirement, and 79 -05B (Item 6)

PWR and BWR ld/lc Refer to Action Plan continuous communications channel.79-06A (Item 11)

Items I.E.6 and-79-06B (Item 10)

III.A.3.3 79-08 (Item 9)

TABLE C.1 (continued)

CP/ML Requirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 13.

Propose Techrical Specification changes79-058 (Item 7)

PWR and BWR la /lc reflecting implementation of all Bulletin 79-06A & Rev. 1 items, as required.

(Item 13)79-06B (Item 12) 79-08 (Item 11) 14.

Review operating modes and procedures79-06A (Item 12)

W, CE GE ld/lc Refer to Action Plan to deal with significant amounts of 79-06B (Item 11)

Items II.B.4, II.B.7, hydrogen.

79-08 (Item 10)

II.E.4.1 and II.F.1 T

U 15.

For facilit.ies with non-automatic AFW 79-06A (Item 5)

W & CE ld/lc Refer to Action Plan initiation, provide dedicated ccerator 79-068 (Item 4)

Item II.E.1.2 in continuous communication wi.h CR to operate AFW.

16.

Implement (immediately) procedures that 79-06A (Item 6)

W & CE Id/lc Refer to Action Plan identify PRZ PORV "Open" indications and 79-068 (Item 5)

Items I.C.1 and II.D.3 that direct operator to close manually at " RESET" setpoint.

17.

Trip PZR Level Bistable so that PZR Lo 79-06A & Rev. 1 W

Ic/lc Press. (rather than PZR Lo Press. and PZR (Item 3)

M Level coincidence) will initiate safety injection. For test, reset Lo Level bistable.

18.

Develop procedures and train operators on 79-05B (Item 1)

B&W id/lc Refer to Action Plan methods of establishing and maintaining Items I.C.1 and I.G.1 natural circulation.

l

TABLE C.1 (continued)

CP/ML Requirement Source for Cateoory Requirement Operating Reactors Applicability Assignment Canments

19. Describe design and procedure modifications79-05B (Item 3)

B&W id/lc Refer to Action Plan (based on analysis) to reduce likelihood Item II.E.5 of automatic PZR PORV actuation in transients.

20. : Provide procedures and training to 79-05B (Item 4).

B&W 4/lc Refer to Appendix 0 operators for prompt manual reactor trip for LOFW, TT, MSIV closure, LOOP, LOSG Level, & Lo PZR Level.

21.

Provide automatic safety grade anticipatory 79-05B (Item 5)

B&W 4/lc Refer to Appendix D reactor tr,ip for LOFW, TT, or significant as g

decrease in SG level.

22.

Describe automatic and manual actions 79-08 (Item 3)

DWR 4/lc-Refer to Appendix D for proper functioning of auxiliary heat removal systems when FW system not operable.

23.

Describe uses and types of RV level 79-08 (Item 4)

BWR 4/lc Refer to Appendix D indication for automatic and manual initiation safety systems.

Also, describe alternative instrumentation.

24.

Perform LOCA analyses for a range of 79-05C (short-PWR Id/lc Refer to Action Plan small-break sizes and a range of term Item 2)

Item I.C.l time lapses between reactor trip 79-06C (short-and RCP trip.

term Item 2)

l I

l TABLE C.1 (continued) l LP/ML Requirement l

Source for.

Category Requirement Operating Reactors Applicability Assignment Comments l

25.

Develop operator action guidelines (based 79-05C'(short-PWR ld/ld Refer to Action Plan l

on analyses in Requirement-24 above).

term Item 3)

Item I.C.1 79-06C (short-term Item 3) 26.

Revise emergency procedures and train RO's79-05C (short-PWR Id /lc Refer to Action Plan and SRO's based on guidelines developed in term Item 4)

I.C.1, I.A.3.1, and Requirement 25 above.79-06C (short-I.G.1 l

y term Item 4) l 27.

Provide analyses end develop guidelines79-05C (short-PWR Id/ld Refer to Action Plan l

and proceddres for inadequate core term Item 5)

Items I.C.1 and II.F.2 cooling conditions. Also, define RCP 79-06C (short-restart criteria.

term Item 5) 28.

Provide design that.will assure automatic NUREG-0623 PWR Id /ld Refer to Action Plan RCP cria for all circumstances where Item II.K.3.5 requirea.

-TABLE C.2 REQUIREMENTS FOR NEW B&W PLANTS DERIVED FROM COMMISSION ORDERS ON OPERATING B&W PLANTS Ye#"uhrement q

Category Comments Requirement Source Applicability Assignment.

1.

Upgrade timeliness and reliability Commission Order

.B&W id /lc Refer to Action Plan of AFW system.

Item II.E.1 2.

Procedures and training to initiate Commission Order B&W 4 /lc Refer to Appendix D and control AFW independent of integrated control system.

i' 3.

Hard-wired control grade anticipatory Commission Order B&W id /lc Refer to Action Plan El reactor trips.

Item II.K.2.10 4.

Small-break LOCA analysis, procedures, Commission Order B&W Id /lc Refer to Action Plan and operator training.

Items I.A.3.1 and I.C.1 5.

Complete THI-2 simulator training for Commission Order B&W ld /lc Refer to Action Plan all operators.

Item I.A.2.6 6.

Reevaluate analysis for dual-level Commission Order Davis-Besse 1 lc /lc setpoint control.

7.

Reevaluate transient of September 24, Commission Order Davis-Besse 1 lc /lc 1977.

Re e 8.

Continued upgrading of AFW system.

Commission Order B&W id /lc m II E

TABLE C.2 (continued)

CP

~

Req /ML uirement t

r Requirement Source Applicability 9.

Analysis and upgrading of integrated Conunission Order B&W 4 /lc Refer to Appendix D control system.

10.

Hard-wired safety grade anticipatory Conunission Order B&W 4/lc Refer to Appendix 0 reactor trips.

11.

Operator training and drilling.

Commission Orderi B&W id/lc Refer to Action Plan Items I. A.3.1, I. A.2.2 I.A.2.5, and I.G 1 cn b

12.

Transient analysis and procedures for Commission Order B&W Id/lc Refer to Action Plan management of small breaks.

Item I.C.1 13.

Thermal-mechanical report -- effect Letter,'D. Ross to B&W 3/lc Refer to Appendix D of HPI on vessel integrity for small-B&W operating plants, break LOCA with no AFW.

8/21/79 14.

Demonstrate that predicted lift Letter, D. Ross to B&W 3/lc Refer to Appendix D frequency of PORVs and SVs is B&W operating plants, acceptable.

8/21/79 15.

Analysis of effects of slug flow on Letter, D. Ross to B&W 3/lc Refer to Appendix D once-through steam generator tubes B&W operating plants, af ter primary system voiding.

8/21/79

TABLE C.2 (continued)

LP/ PIL Requirement Requirement Source Applicability As n n Comments 16.

Impact of RCP seal d* mage following Letter, D. Ross to B&W 3/1'c Refer to ' Appendix D small-break LOCA with loss of offsite B&W operating power.

plants, 8/21/79 17.

Analysis of potential voiding ia Letter, R. Reid All B&W ld/lc Refer to Action Plan RCS during anticipated transients, to all B&W operating Item I.C.1 plants 1/9/80 l

18. Analysis of loss of feedwater and Letter, D. Ross to All B&W id/ic Refer to Action Plan

. T' other anticipated transients.

B&W operating plants, Iten I. Col M

8/21/79 19.

Benchmark analysis of sequential Letter, D. Ross to All B&W ld/lc Refer to Action Plan AFW flow to once-through steam B&W operating plants, Item I.C.I generator.

8/21/79 20.

Analysis of system response to small-Letter, D. Ross to All B&W id/lc Refer to Action Plan break LOCA that causes system pressure B&W operating plants Item I.C.l to exceed PORV setpoint.

8/21/79 21.

LOFT 3-1 predictions.

Letter, D. Ross-to All B&W

.le/l c B&W operating plants, 8/21/79

TABLE C.3 FINAL RECOMMEN0ATIONS OF BULLETINS AND ORDERS TASK FORCE CP/ML Requirement Requirement Source Applicability Category Assionment Coments 1.

Install automatic PORV isolation NUREG-0565(2.1.2.a)

PWR Id/1d Refer to Action Plan system and perform operational NUREG-0611(3.2.4.e Item II.K.3.2-test.

3.2.4.f)

NUREG-0635(3.2.4.a)

'(3.2.4.b) 2.

Report on overall safety effect NUREG-0565 (2.1;2.d) 3/3

.Refef to Appendix D -

of PORV isolation system.

NUREG-0611(3.2.4.g, PWRs i"

13.2.4.i)

C' NUREG-0635(3.2.4.c) 3.

Report safety and relief valve NUREG-0565(2.1.2.c, All 2/2 failures promptly and challenges 2.1.2.e) annually.

NUREG-0611(3.2.4.h, 3.2.4.j)

NUREG-0626(8.14)

NUREG-0635(3.2.4.d) l l

4.

Review and upgrade reliability NUREG-0565(2.-3.2.b)

All lb/lb Refer to Action Plan I

and redundancy of non-safety NUREG-0611(3.2.2.b)

Items II.C.1, II.Cs2, equipment for small-break LOCA NUREG-0626 (B.12, and II.C.3 mitigation.

NUREG-0635(3.2.2.b) l 5.

Continue to study need for NUREG-056F(2.3.2.a)

PWR 4/4~

Refer to Appendix D l

C.1.4.c and need for auto-NUREG-0611(3.2.2.a) matic trip of RCPs, then NUREG-0635(3.2.2.a) modify procedures or designs NUREG-0623 ds appropriate.

j

1 TABLE C 3 (continued)

CP/t'L Requirement Category Requirement Source Applicability Assignment Consnents 6.

Instrumentation to verify NUREG-0565(2.6.2.b)

PWR Id/ld Refer to Action Plan Items natural circulation.

NUREG-0611(3.2.3.b)

I C,1, II.F.2, and II.F.3 NUREG-0635(3.2.3.b) 7.

Evaluation of PORV opening NUREG-0565(2.1.2.b)

B&W Id/lc Refer to Action Plan Item probability during overpressure II,K.2.14 transient.

8.

Further staff consideration of NUREG-0565(2.5.2.a)

PWR ld/ld Refer to Action Plan Items a

need for diverse decay heat NUREG-0635 (4.2.5.,

II,C,1 and II.E,3.3 removal method independent App. VIII) o of SG's NUREG-0611 (4.2.5, App. VIII) 9.

Proportional integral Derivative NUREG-0611(3.2.4.b)

W Ic/2 controller modification.

10.

Anticipatory trip modifcation NUREG-0611(3.2.4.c)

W Ic/2 proposed by some licensees to confine range of use to high power levels, 11.

Control use of PORV supplied NUREG-0611(3.2.4.d)

All 4/4 Refer to Appendix D by Control Components Inc. until further review complete.

12.

Confirm existence of anticipa-NUREG-0611(3.2.4.a)

W lc/2 tory trio upon turbine trip.

TABLE C.3 (continued)

CP/ML Requirement Category Requirement Source Applicability Assignment Comments 13.

Separation of HPCI and RCIC NUREG-0626(A.1)

GE 3/lc Refer to Appendix D system initiation levels.

Analysis and implementation.

14.

Isolation of isolation NUREG-0626(A.2)

GE plants lc/lc condensers on high radiation.

with isolation

. condenser co

$3 15.

Modify break detection logic NUREG-0626(A.3)

GE 2/lc to prevent spurious isolation of HPCI and RCIC systems, i

i 16.

Reduction of challenges and NUREG-0626(A.4)

GE 3/lc Refer to Appendix D

' failures of relief valves -

l feasibility study and system l

modification.

17.

Report on outage of ECC NUREG-0626(A.6)

GE la/lc systems - licensee report and proposed technical specification changes.

18.

Modification of ADS logic NUREG-0626(A.7)

GE 3/lc Refer to Appendix D feasibility study and modifica-tion for increased diversity for some event sequences.

TABLE C.3 (continued)

CP/ML Requirement Category Requirement Source Applicabili ty Assignment Comments 19.

Interlock on recirculation NUREG-0626(A.8)

GE Non-Jet lc/lc pump loops.

Pump ors 20.

Loss of service water for NUREG-0626(A.9)

Big Rock Ic/lc Big Rock Point.

Point 21.

Restart of core spray and LPCI NUREG-0626(A.10)

GE 3/lc Refer to Appendix D systems on low level - design as gj and modification.

22. Automatic switchover of RCIC NUREG-0626(B.1)

GE Ic/lc system suction - verify procedures and modify design.

23.

Central water level recording.

NUREG-0626(B.2)

GE 4/lc Refer to Appendix D 24.

Confirm adequacy of space cool-NUREG-0626(B.3)

GE 3 /lc Refer to Appendix D ing for HPCI and RCIC systems.

25.

Effect of loss of AC power on NUREG-0626(B.4)

GE 3/lc Refer to hppendix D pump seals.

TABLE C.3 (continued) t,?/MLJ Requirement i

Category Requirement Source Applicability Assignment-Comments 26.

Study effect on RHR reliability NUREG-0626(B.5)

GE Id/lc.

Refer to Action Plan Item II.E.2.1 of its use for fuel pool cooling.

27.

Provide common reference level NUREG-0626(B.6)

GE 2/lc s

for vessel level instrumentation.

28.

Study and verify qualification NUREG-0626(B.7)

GE 3/lc Refer to Appendix D of accumulators on ADS valves.

?'

w 29.

Study to demonstrate perform-NUREG-0626(B.13)

GE Isolation ic/lc ance of isolation condensers Condenser ors with non-condensibles.

30.

Revised small-break LOCA methods NUREG-0565(2.2.2.a)

All 3/lc Refer to Appendix D to show compliance with 10 CFR NUREG-0611(3.2.1.a) 50, Appendix K.

NUREG-0626(A.12)

NUREG-0635(3.2.1.a) l (3.2.5.a) l l

31.

Plant-specific calculations to NUREG-0565(2.2.2.b)

All 3/lc Refer to Appendix D show compliance with 10 CFR NUREG-0611(3.2.1.b) 50.46.

NUREG-0626(A.13, B.10)

NUREG-0635(3.2.1.b) 1 1

- g3

l TABLE C.3 (continued)

CP/ML Reouirement Category Requirement Source Applicability Assignment Comments 32.

Provide experimental verifica-NUREG-0565(2.6.2.a)

PWR lb/lb Refer to Action Plan II.E.2.2 tion of two phase natural NUREG-0611(3.2.3.a) circulation models.

NUREG-0635(3.2.3.a) 33.

Evaluate elimination of PORV NUREG-0565(3.5)

PWR lb/lb Refer to Action Plan Item function.

NUREG-0611(3.2.4.k)

I? ^.1 NUREG-0635(3.2.4.e) co 34.

RELAP-4 model development.

NUREG-0511(3.2.5)

PWR lb/lb Refer to Action Plan Item 3:

NUREG-0635(3.2.5)

II.E.2.2 35.

Evaluation of effects of core NUREG-0565(2.2.2.c)

B&W ld/lc Refer to Action Plan Item flood tank injection on small-I.C.l break LOCAs.

36.

Additional staff audit calcula-NUREG-0565(2.4.2.a)

B&W lb/lc Refer to Action Plan Item tions of B&W small-break LOCA I.C.1 analyses.

37.

Analysis of B&W plant response NUREG-0565(2.6.2.c)

B&W Id/lc Refer to Action Plan Item to isolated small-break I.C.1 LOCA.

38.

Analysis of plant response to NUREG-0565(2.6.2.d>

B&W Id /lc Refer to Action Plan Item a small-break LOCA in the I.C.1 pressurizer spray line.

TABLE C.3 (continued)

CP/ML Requirement Category Requirement Source Applicability Assignment Comments i.

39.

Evaluation of effects of water NUREG-0565(2.6.2.e)

B&W ld/lc Refer to Action Plan Item slugs in piping caused by HPI I.C.l and CFT flows.

40.

Evaluation of RCP seal damage NUREG-0565(2.6.2.f)

B&W id/lc Refer to Action Plan Item and leakage during a small-II.K.2.16 break LOCA.

41.

Submit p.edictions for LOFT Test NUREG-0565(2.6.2.g)

B&W id/lc Refer to Action Plan Item L3-6 with.RCPs running. I.C.l m 42. Submit requested information NUREG-0565(2.6.2.h) B&W id/lc Refer to Action Plan Item on the effects of non-I.C.1 condensible gases. l l 43. Evaluation of mechanical effects NUREG-0565(2.6.2.i) B&W id/lc Refer to Action Plan Item of slug flow on steam generator II.K 2.15 tubes. 44. Evaluation of anticipated NUREG-0626(A.14) GE 3/lc Refer to Appendix D transients with single failure to verify no significant fuel failure. 45. Evaluate depressurization with NUREG-0626(A.15) GE 3 /lc Refer to Appendix D other than full ADS. .J

TABLE C.3 (continued) CP/ML Requirement Category Requirement Source Applicability Assignment Comments 46. Response to list of concerns NUREG-0626(A.17) GE 4/lc Refer to Appendit D from ACRS consultant. 47. Test program for small-break NUREG-0626(B.9) GE ld/lc Refer to Action Plan LOCA model verification pretest Items I.C.1, and II.E.2.2 prediction, test program and model verification. ,48. Assess change in safety NUREG-0626(B.15) GE Id/lc Refer to Action Plan Items II.C.1 and II.C.2 a reliability as result of imple-menting B&OTF recommendations. m 49. Review of procedures (NEC). NUREG-0611(3.4.1) W. CE lb/lb Refer to Action Plan Items. NUREG-0635(3.4.1) I.C.8 and I.C.9 50. Review of procedures NUREG-0611(3.4.2) W, CE ld/lc Refer to Action Plan Items (NSSS vendors) NUREG-0635(3.4.2) I.C,7 and I.C.9 51. Symptom-based emergency NUREG-0611(3.4.3) W. CE ld/lc Refer to Action Plan Item procedures. NUREG-0626(B.8) GE I C.9 NUREG-0635(3.4.3) 52. Operator awareness of revised NUREG-0626(A.11) GE ld/lc Refer to Action Plan Items emergency procedures. I,B.1, I.C,2, and I.C.5

- ~. _. h t TABLE C.3 (continued) CP/ML Requirement Category Requirement Source Applicability-Assignment Comments i 53. Two operators in control room. NUREG-0626(A.16) GE ld/lc Refer to Action Plan Item I.A.l.3 i

54. Simulator upgrade for small-NUREG-0565(2.3.2.c)

All ld/lc Refer to Action Plan break LOCAs. NUREG-0611(3.3.1.b) Item I.A.4.1 NUREG-0626(B.11) 'NUREG-0635(3.3.1.b) i:

55. Operator monitoring of control NUREG-0611(3.5.1)

W, CE Id/lc Refer to Action Plan Items i tj board. NUREG-0635(3.5.1) I.C.1, I.D.2 and I.D.3 56. Simulator training requirements. NUREG-0611(3.3.1.a) W, CE Id/lc Refer to Action Plan NUREG-0635(3.3.1.a) Items I.A.3.1,- I.A.3.3, and I.A.2.6 l l 57. Identify water sources NUREG-0626(A.5) GE Id /lc Refer to Action Plan Item i i prior to manual I.C.1 activation of ADS i-f i I I , ~

I APPENDIX C ACTION PLAN ITEMS APPLICABLE TO PENDING CONSTRUCTION PERMIT AND MANUFACTURING LICENSE APPLICATIONS CP ML I. A.1.1 Shift Technical Advisor X I.A.1.2 Shift Supervisor Administrative Duties X I.A.1.3 Shift Manning X I.A.2.5 Revise Scope and Criteria for Licensing Exams X I.A.4.2 Long-Term Training Simulator Upgrade X I.B.1.1 Organization and Management Long-Term Improvements X X I.C.1 Short-Term Training Simulator Upgrade X X I.C.2 Shift and Relief Turnover Procedures X I.C.3 Shift Supervisor Responsibilities X I.C.4 Control Room Access X I.C.5 Procedures for Feedback of Operating Experience X X [ I.C.6 Procedures for Verification of Correct Performance of Operating Activities X I.C.7 NSSS Vendor Review of Procedures X I.C.9 Long-Term Program Plan for Upgrading Procedures X I. D.1 Control Room Design Reviews X X I.D.2 Plant Safety Parameter Display Console X X I.D.3 Safety System Status Monitoring X X I.D.4 Control Room Design Standard X X I.E.4 Coordination of Licensee, Industry, and Regulatory Programs X X I. F.1 Expand QA List X X I.F.2 Develop More Detailed Criteria X X II.A.2 Site Evaluation of Existing Facilities X II.B.1 Reactor Coolant System Vents X X C-1

~. 2 CP ML l 1 Plant Shielding to Provide Access to Vital II.B.2 Areas and Protect Safety Equipment From Post-Accident Operation X X Post-Accident Sampling X X II.B.3 Training for Mitigating Core Damage X II.B.4 Rulemaking Proceeding on Degraded Core Accidents X X 'II.B.8 Reliability Engineering X X II.C.4 Testing Requirements X X II.D.1 Research on Relief and Safety Valve Test II.D.2 Requirements X X Post Accident Sampling X X II.D.3 II.E.1.1 Auxiliary Feedwater System Evaluation X X II.E.1.2 Auxiliary Feedwater System Automatic Initiation and Flow Indication X X II.E.2.1 Reliance on ECCS X Uncertainties in Performance Predictions X II.E.2.3 Reliability of Power Supplies for Natural II.E.3.1 Circulation X X II.E.4.1 Dedicated Penetrations X X Isolation Dependability X X II.E.4.2 Integrity Check X II.E.4.3 II.E.4.4 Purging X X Design Evaluation X II.E.5.1 B&W Reactor Transient Response Task Force X II.E.5.2 I I'. F.1 Additional Accident Monitoring Instrumentation X X Identification and Recovery from Conditions II.F.2 Leading-to Inadequate Core Cooling X X Instrumentation for Monitoring Accident II.F.3 Conditions (Reg. Guide 1.97) X X Power Supplies for Pressurizer Relief Valves, II.G.1 Block Valves, and Level Indicators X X Organization and Staffing to Oversee Design II.J.3.1 and Construction X X C-2

4 CP ML II.K.1.20 Provide Procedures and Training to Operators for Prompt Manual Reactor Trip for LOFW, TT, MSIV Closure, LOOP, LOSG Level, and Low Pressurizer Level X II.K.1.21 Provide Automatic Safety-Grade Anticipatory Reactor Trip for LOFW, TT, or Significant Decrease in SG Level X II.K.1.22 Describe Automatic and Manual Actions for Proper Functioning of Auxiliary Heat Removal Systems when FW System is not Operable X II.K.1.23 Describe Uses and Types of RV Level Indica-tion for Automatic and Manual Initiation of Safety Systems. Also Describe Alternative Instrumentation. X II.K.2.2 Procedures and Training to Initiate and Control AFW System Independent of Integrated Control System X II.K.2.9 Analysis and Upgrading of Integrated Control System X II.K.2.10 Hard-Wired Safety-Grade Anticipatory Reactor Trips X II.K.2.13 Thermal-Mechanical Report. Effect of HPI on Vessel Integrity for Small-Break LOCA with no AFW X II.K.2.14 Demonstrate that Predicted Lift Frequency of PORVs and SVs is Acceptable X II.K.2.15 Analysis of Effects of Slug Flow on Once-Through Steam Generator Tubes After Primary System Voiding X II.K.2.16 Impact of RCP Seal Damage Following Small-Break LOCA with Loss of Offsite Power X II.K.3.2 Report on Overall Safety Effect of PORV Isolation System X X ^ C-3

I CP ML Report Safety and Relief Valve Failures II.K.3.3 Promptly and Challenges Annually X X Continue to Study Need for Trip of RCPs. II.K.3.5 Modify Procedures or Designs as Appropriate X X Proportional Integral Derivative Controller II.K.3.9 Modification X Anticipatory Trip Modification by Some II.K.3.10 Licensees to Confine Range of 'Jse to High Power Levels X II.K.3.11 Control Use of PORV Supplied by Control Components, Inc. Until Further Review is Completed X X II.K.3.13 Separation of HPCI and R'IC Sys.em Initiation Levels. Analysis and Implementation X II.K.3.15 Modify Break Detection Logic to Prevent Spurious Isolation of HPIC and RCIC Systems X Reduction of Challenges and Failures of II.K.3.16 Relief Valves. Feasibility Study and System Modification. X Modification of ADS Logic. Feasibility II.K.3.18 Study and Modification for Increased Diversity for Some Event Sequences X II.K.3.21 Restart of Core Spray and LPCI Systems on Low Level. Design and Modification. X II.K.3.23 Central Water Level Recording X Confirm Adequacy of Space Cooling for HPCI II.K.3.24 and RCIC Systems X Effect of Loss of AC Power on Pump Seals X II.K.3.25 II.K.3.27 Provide Common Reference Level for Vessel Level Instrumentation X Study and Verify Qualification of Accumulators II.K.3.28 on ADS Valves X II.K.3.30 Revise Small-Break LOCA Methods to Show Compliance with 10 CFR 50, Appendix K X X C-4

i l i CP ML Plant Specific Calculations to Show II.K.3.31 Compliance with 10 CFR 50.46 X X Evaluation of Anticipated Transients II.K.3.44 with Single Failure to Verify no Significant Fuel Failure X Evaluate Depressurization with Other Than II.K.3.45 Full ADS X Response to List of Concerns From ACRS II.K.3.46 Consultant X III.A.1.1 Upgrade Emergency Preparedness X III.A.1.2 Upgrade License Emergency Support Facilities X X Maintain Supplies of Thyroid Blocking Agent III.A.I.3 (Potassium Iodide) X III.A.2.1 Amend 10 CFR Part 50 and 10 CFR Part 50, Appendix E X III.A.2.2 Development of Guidance and Criteria X X III.A.3.3 Communications X III.A.3.5 Training, Drills, and Tests X III.D.1.1 Primary Coolant Sources Outside the Containment Structure X X III.D.1.2 Radioactive Gas Management X .X III.D.1.3 Ventilation System and Radiciodine Ads 9rber Criteria X X III.D.2.3 Liquid Pathway Radiological Control X X III.D.2.4 Offsite Dose Measurements X III.D.2.5 Offsite Dose Calculation Manual X III.D.3.1 Radiation Protection Plans X X III.D.3.3 In-Plant Radiation Monitoring X X III.D.3.4 Control Room Habitability X X C-5

I APPENDIX D INFORMATION REQUIREMENTS FOR TMI-2 ACTION PLAN ITEMS IN CATEGORIES 3, 4, AND 5 I.A.4.2 LONG-TERM TRAINING SIMULATOR UPGRADE Applicants shall describe their program for providing simulator capability for their plants. In addition, they shall describe how they will assure that their proposed simulator will correctly model their control room. Applicants shall provide sufficient information to permit the NRC staff to verify that they will have the necessary simulator capability to carry out the actions described in this Action Plan item as well as Action Plan Item II.K.3.54. Applicants shall submit, prior to the issuance of construction permits, a general discussion of how the requirements will be met. Sufficient details shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. l I. B. l.1 ORGANIZATION AND MANAGEMENT LONG-TERM IMPROVEMENTS l Applicants shall define the operations titles, management positions and organi-zational elements in terms of an applicant constructing a plant. Applicants shall describe how the organization will function to assure that the plant is constructed in accordance with NRC requirements. The current criteria for Utility Management and Technical Competence shall be used by applicants as guidelines in completing this activity. Applicants shall alsu address the manner by which they will assure close integration of the utility-architect r engineer and nuclear steam supply system vendor in their program for design, construction, and operation of their facilities. Applicants shall submit detailed information in order to provide reasonable assurance that the require-ment is implemented properly prior to the issuance of the construction permits or manufacturing license. D-1

e I. C.1 SHORT-TERM ACCIDENT ANALYSIS AND PROCEDURES REVISION Applicants shall perform analyses which will then be used to develop emergency operational procedures for handling small break loss-of-coolant accidents. Applicants shall develop procedures to assist the plant operating staff to recognize and prevent impending core uncovering, and recover from a condition in which the core has experienced inadequate core cooling. Applicants shall perform analysis of transients and accidents, using guidance provided by the NRC staff, and develop emergency procedures consistent with the actions neces-sary to cope with the transients and accidents analyzed. This effort shall include the development of procedures for operating with natural circulation conditions. Applicants shall submit, prior to the issuance of construction permits or manufacturing license, a general discussion of how the requirements will be met, including the program for review of the procedures by the nuclear steam supply system and architect-engineer. Sufficient detail shall be pre-sented to provide reasonable assurance that the requirements will be imple-mented properly prior to the issuance of operating licenses. I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE Applicants shall submit a description of their programs for the evaluation of operating experience and describe how they will assure that important industry experiences originating from both within and outside the construction organiza-tion are continually provided to those designing and constructing the plant. Applicants shall submit a general discussion of how the requirements will be met. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of construction permits or manufacturing license. I I.C.9 LONG-TERM PROGRAM PLAN FOR UPGRADING 0F PROCEDURES Applicants shall describe their program plan which is to begin during construc-tion and follow into operation for integrating and expanding current efforts in the area of plant procedures. The scope of the program should include emergency procedures, reliability analysis, human factors engineering, crisis l management and operator training. Applicants shall also describe how their j D-2 l

program will be coordinated with INP0 activities. Applicants will submit, prior to the issuance of construction permits, a general discussion of how the requirements will be met. Sufficient detoil shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. I. D.1 CONTROL ROOM DEStGN REVIEWS Applicants shall describe their program for reviewing their control room designs to identify and correct design deficiencies as described in this Action Plan item. Applicants shall, to the extent possible, provide prelim-inary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. I.O.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE l Applicants shall describe how they intend to meet the staff criteria for the plant safety parameter display console. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. I.D.3 SAFETY SYSTEM STATUS MONITORING Applicants shall describe how their design conforms to Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety D-3

1 Systems." Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, appli-cants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. I.D.4 CONTROL ROOM DESIGN STANDARD Applicants shall describe the extent to which their control room design conforms to IEEE 566 and IEEE 567 which are scheduled to be amended later this year. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Wnere new designs are involved, appli-cants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the' requirements will be implemented properly prior to the issuance of operating licenses. I.E.4 COORDINATION OF LICENSEE, INDUSTRY AND REGULATORY PROGRAMS Applicants shall, in conjunction with Action Plan Item I.C.5, provide a description of their program to evaluate experience both at their own plant and similar plants and factor this experience, as appropriate, into the design and construction of their plant. In addition, the program shall describe how these activities will be factored into the operation of the plant. Applicants shall submit, prior to the issuance of construction permits or manufacturing license, a general explanation of how the requirements will be met. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly. l D-4

l [ l t I.F.1 EXPAND QA LIST Applicants shall describe their program to expand their QA lists as a result of the accident at TMI-2. Prior to issuance of the construction permits or manufacturing license, applicants shall provide a commitment to apply their QA program to items and activities affecting safety as defined by Regulatory Guide 1.29 and Appendix A to 10 CFR Part 50, and shall provide a revised QA program that includes all such items and activities. I.F.2 DEVELOP MORE DETAILED QA CRITERIA Applicants shall describe the changes to their QA program that have arisen as a result of their review of the accident at THI-2. In addition, applicants shall address the appropriate matters discussed in this Action Plan item and the extent to which they have been considered in their QA program. Applicants shall submit, prior to the issuance of the construction permits or manufactur-ing license, a revised description of their QA program that includes consideration of these matters. II.A.2 SITE EVALUATION OF EXISTING FACILITIES The Commission has already established a transition policy for construction permit applicants. Applicants will be asked to compare their sites with the recommendations of NUREG-0625, as modified by comments from the NRC's Office of Policy Evaluation and Advisory Committee on Reactor Safeguards. At such time as the proposed rule on siting is issued for comment (scheduled for October 1980), applicants will be required to assess their sites against the criteria contained in the proposed rule. II. B.1 REACTOR COOLANT SYSTEM VENTS Applicants shall (1) modify their plant designs as necessary to provide high point reactor coolant system and reactor vessel head vents that can be remotely operated from the control room, and (2) demonstrate by analysis that direct venting will not result in violation of combustible gas concentration limits. Applicants shall, to the extent possible, provide preliminary design information D-5

at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specify-ing the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feas-ible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.B.2 PLJ,4T SHIELDING TO PROVIDE ACCESS TO VITAL AREAS AND PROTECT SAFETY EQUIPMENT FOR POST-ACCIDENT OPERATION Applicants shall (1) perform radiation and shielding design reviews of spaces around systems that may contain highly radioactive fluids and (2) implement plant design modifications necessary to permit adequate access to vital areas and protect safety equipment. Applicants shall, to the extent possible, pro-vide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.B.3 POST-ACCIDENT SAMPLING Applicants shall (1) review the reactor coolant and containment atmosphere sampling system designs and the radiological spectrum and chemical analysis facility designs, and (2) modify their plant designs as necessary to meet the requirements. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, appli-cants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Ahlicants shall also demonstrate that the design D-6

concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.B.8 MJLEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS Applicants shall describe the degree to which their designs conform to the proposed interim rule. Applicants shall also provide reasonable assurance to the extent practicable and taking into account the present state of the art of this technology that issuance of construction permits or the manufacturing license will not foreclose or preclude the modification of the facilities to accommodate potential requirements that may result from the rulemaking proceed-ings. These potential requirements include such features as filtered vented containment, molten core retention, and hydrogen control systems. Special attention should be given to those facility designs with small containment volu:::es, i.e., ice condenser and Mark III containment designs. Prior to issuance of a construction permit or manufacturing license, applicants will also be required to submit their evaluation of the additional features, both preventive and mitigative; they propose to include at their facilities that have the potential for significant risk reduction. II.C.4 RELIABILITY ENGINEERING Applicants shall perform simplified system reliability analyses for the following systems: subcriticality systems, emergency feedwater systems (PWRs), reactor core isolation cooling system, (BWRs), emergency core cooling system injection and recirculation systems, shutdown cooling system, containment cooling and spray systems, safety features actuation systems, and auxiliary systems upon which these depend (alternating and direct current, compressed air, essential service water or cooling systems, and heating, ventilating and air conditioning systems). These analyses shall use event-tree and fault-tree logic techniques to identify design weaknesses and possible system modifications that could be made to improve the capability and reliability of the above systems under various transient and loss of coolant accident events. Particular emphasis shall be given to determining potential failures that could result from human a:rrors, common causes, single point vulnerabilities, and test and maintenance cutages. D-7

Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program j to assure that the results of such studies are factored into the final designs. II.D.1 TESTING REQUIREMENTS Applicants and their agents shall plan and carry out the test program and model development. Consideration of anticipated transient without scram (ATWS) conditions shall be included in the test planning. Actual testing under ATWS conditions may not be carried out until subsequent phases of the test program are developed. Applicants shall submit, prior to the issuance of the construction permits or manufacturing license, a general explanation of how tL requirements will be met. Sufficient detail should be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.D.2 RESEARCH ON RELIEF AND SAFETY VALVE TEST REQUIREMENTS Applicants shall perform studies to (1) demonstrate the applicability of the generic tests to their particular plants and modify their plant design as necessary. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final design. II.D.2 RESEARCH ON RELIEF AND SAFETY VALVE TEST REQUIREMENTS Applicants shall (1) demonstrate the ap,i h 4 Hity of the generic tests to their particular plants and (2) modify their plant designs as necessary. Applicants shall commit, prior to the issuance of the construction permits or manufacturing license, to comply with the requirements and shall submit within six months following the completion of the generic tests or the issuance of construction permits, whichever is later, a detailed explanation of how the requirements resulting from the tests will be met. Sufficient detail should be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. D-8

l II.D.3 RELIEF AND SAFETY VALVE POSITION INDICATION Applicants shall modify their plant designs as necessary to provide direct indication of relief and safety valve position in the control room. Appli-cants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.D.1.1 AUXILIARY FEEDWATER SYSTEM EVALUATION Applicants shall perform a reevaluation of their proposed auxiliary feedwater (AFW) system. This reevaluation shall include the following: (1) Performance of simplified auxiliary feedwater system reliability analyses using event-tree and fault-trae logic techniques to determine the potential for AFW system failure unde various loss of main feedwater transient conditions, with par-ticular emphasis being given to determining potential failures that could result from hLian errors, common causes, single point vulnerabilities, and test and maintenance outages. The results of this evaluation shall be compared with the results of the NRC staff's generic AFW system evaluation published in Appendix III to NUREG-0611 and Appendix III to NUREG-0635. Applicants with plants with AFW systems with relatively low reliabilities shall submit proposed design changes and/or procedural actions which will improve the relative reliability of the AFW system to above average. Applicants whose plant designs do not include high head high pressure injection system pumps for use in the feed and bleed mode of decay heat removal in case of AFW system failure shall assure that their AFW system has a very high reliability relative to those AFW systems evaluated by the NRC and staff and reported in NUREG-0611 and NUREG-0635 respectively. D-9

(2) Completion of a deterministic review of the AFW system using the acceptance criteria of Standard Review Plan Section 10.4.9 as principal guidance. This requirement applies to those plants where the Standard Review Plan was not used as criteria during the NRC staff's CP review. (3) Reevaluation of the AFW system finw design bases and criteria. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.E.1.2 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION Applicants with PWR plants which have manually initiated auxiliary feedwater (AFW) systems shall submit (1) proposed safety grade designs which meet the requirements specified in Sections 2.1.7.a and 2.1.7.b of NUREG-0578, and (2) analyses of a potential unreviewed safety issue relating to automatic AFW system initiation with a postulated main steam line break inside contain-ment and its effect on containment pressure design capability and return to reactor power. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, appli-cants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. Applicants with PWR plants which have automatically initiated AFW systercs shall provide information in sufficient detail to provide reasonable assurance that their designs are safety grade and meet the requirements specified in Sections 2.1.7.a and 2.1.7.b of NUREG-0578. l D-10

II.E.2.3 UNCERTAINTIES IN PERFORMANCE PREDICTIONS Applicants shall (1) demonstrate the applicability of the generic evaluations to their particular plants and (2) modify their analysis methods as necessary. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.E.3.1 RELIABILITY OF POWER SUPPLIES FOR NATURAL CIRCULATION Applicants shall (1) upgrade the power supplies for the pressurizer heaters and associated motive and control power interfaces to meet the applicable requirements specified in Section 2.1.1 of NUREG-0578 and (2) establish pro-cedures and training for maintaining the reactor coolant system at hot standby conditions with only onsite power available. Applicants shall, to the extent pssible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Wher's new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feas-ible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.E.4.1 DEDICATED PENETRATIONS Applicants for plant designs with external hydrogen recombiners shall modify their applications as necessary to include redundant dedicated containment penetrations so that the recombiner systems can be connected to the contain-ment atmosphere without violating single-failure criteria, such as having to open large containment purging ducts or otherwise jeopardizing the containment function. Applicants shall submit, prior to the issuance of construction permits or the manufacturing license, a detailed explanation of how the D-11

requirements will be met in order to provide reasonable assurance that the requirements will be implemented properly. II.E.4.2 ISOLATION DEPENDABILITY Applicants shall evaluate their plant designs for isolation dependability and for purge valve closure on a high airborne radiation signal and shall modify their plant designs as needed. Applicants shall review their containment pressure setpoint and reduce it as necessary. Applicants shall also modify their plant designs as necessary to include high-radiation isolation signal circuitry. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.E.4.4 PURGING Applicants shall (1) address restricted purging and justification of any unrestricted purging, (2) evaluate the performance of purging and venting isolation valves against accident pressure, (3) address the interim NRC guid. ance on valve operability and (4) adopt procedures and restrictions consistent with-the revised requirements. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a ~ general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically D-12

feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.E.5.1 DESIGN EVALUATION Applicants with B&W-designed reactors shall (1) identify the most severe overcooling events (considering both anticipated transients and accidents) that could occur at the facilities, (2) show, in view of the arrival rate for these events, that the design criterion for the number of actuation cycles of the emergency core cooling system and raactor protection system is adequate, (3) recommend changes to systems or procedures that would reduce primary system sensitivity. Applicants with B&W-designed reactors shall, to the extent pos-sible, provide preliminary design information at a level consistent with that normally required at che construction permit stage of review. Where new designs are invntved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demon-strate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.E.5.2 B&W REACTOR TRANSIENT RESPONSE TASK FORCE Applicants with B&W-designed reactors shall address the additional licensing requirements resulting from this action plan item when issued. Applicants with B&W-designed reactors shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, appli-cants shall provids a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. D-13

II.F.1 ADDITIONAL ACCIDENT M0MITORING INSTRUMENTATION Applicants shall comply with the requirements addressed to construction permit applicants in NRR letters dated October 10, 1979 and November 9, 1979. Appli-cants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.F.2 IDENTIFIC4 TION OF AND RECOVERY FROM CONDITIONS LEADING TO INADEQUATE CORE COOLING Applicants shall describe their program for deve?oping and implementing procedures to be used by the reactor operators to detect and recover from conditions leading to inadequate core cooling. Applicants with PWR plants shall incorporate in their plant designs a primary coolant saturation meter and all applicants shall incorporate in their plant designs instrumentation to detect conditions with a potential that may lead to inadequate core cooling. Any additional equipment that could be used to indicate inadequate core cooling shall be incorporated in the plant designs. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit plicants shall provide a stage of review. Where new designs are involvec' general discussion of their approach to meeting i requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. D-14 j l

II.F.3 INSTRUMENTATION FOR MONITORING ACCIDENT CONDITIONS (REG. GUIDE 1.97) Applicants shall provide in their facility design instrumentation to monitor plant variables and systems during and following an accident in accordance with design bases that are defined, and criteria and requirements that are specified in Regulatory Guide 1.97 (to be issued in final form by August 1980). Designs are already established for much of the instrumentation that will be required; some of the requirements, however, may involve state-of-the-art designs or designs which have yet to be developed. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.G.1 POWER SUPPLIES FOR PRESSURIZER RELIEF VALVES, BLOCK VALVES, AND LEVEL INDICATION Applicants with PWR plants shall upgrade the power supplies for the pressurizer relief valves, Slock valves, and pressurizer level indicators to meet the applicable requirements specified in Section 2.1.1 of NUREG-0578. Applicants with PWR plants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construc-tion permit stage of review. Where new designs are involved, applicants shall provide a gener:41 discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is tech-nically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. D-15

II.J.3.1 ORGANIZATION AND STAFFING TO OVERSEE DESIGN AND CONSTRUCTION Applicants shall describe their program for improving the oversight of design, j construction, and modification activities so that they will gain the critical expertise necessary for the safe operation of the plant. Specific items to be addressed include (1) the technical resources to oversee the desigii and construction of the plant considering the number of people, expertise, comp-etency, and scope of work to be performed and (2) the degree of management and technical control to be exercised by the utility during design and con-struction, including the preparation and implementation of procedures necessary to guide the effort. Applicants shall submit detailed information in order to provide reasonable assurance that the requirements will be implemented properly prior to issuance of the construction permits or manufacturing license. II.K.1.20 PROVIDE PROCEDURES AND TRAINING TO OPERATORS FOR PROMPT MANUAL REACTOR TRIP FOR LOFW TT, MSIV CLOSURE, LOSG LEVEL, & LO PZR LEVEL Applicants with B&W-designed plants shall address the requirements set forth in action item 4 of IE Bulletin 79-05B. A general explanation of how these requirements will be met is required prior to issuance of the construction permits. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly. II.K.1.21 PROVIDE AUTOMATIC SAFETY-GRADE ANTICIPATORY REACTOR TRIP FOR LOFW, TT, OR SIGNIFICANT DECREASE IN SG LEVEL Applicants with B&W-designed plants shall address the requirements set forth in Action Item 5 of IE Bulletin 79-058. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demon-strate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. D-16

1 II.K.1.22 DESCRIBE AUTOMATIC AND MANUAL ACTIONS FOR PROPER FUNCTIONING OF AUXILIARY HEAT REMOVAL SYSTEMS WHEN FW SYSTEM NOT OPERABLE Applicants with B&W plants shall address the requirements set forth in. action item 3 of IE Bulletin 79-80. A general explanation of how these requirements will be met is required prior to issuance of the construction permits. Suf-ficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly. II.K.1.23 DESCRIBE USES AND TYPES OF RV LEVEL INDICATION FOR AUTOMATIC AND MANUAL INTERACTION OF SAFETY SYSTEMS. ALSO DESCRIBE ALTERNATIVE INSTRUMENTATION Applicants with BWR plants shall address the requirements set forth in action item 4 of IE Bulletin 79-08. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.K.2.2 PROCEDURES AND TRAINING TO INITIATE AND CONTROL AFW INDEPENDENT OF INTEGRATED CONTROL SYSTEM Applicants with B&W-designed plants shall address the requirements set forth in the Commission Orders regarding procedu..s and training to initiate and control auxiliary feedwater independent of the integrated control system. A general explanation of how these requirements will be met is required prior to issuance of the construction permits. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly. 0-17

II.K.2.9 ANALYSIS AND UPGRADING OF INTEGRATED CONTROL SYSTEM Applicants with B&W-designed plants shall address the requirements set forth in the Commission Orders regarding the analysis and upgrading of the integrated control system. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, appli-cants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and.the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.K.2.10 HARD-WIRED SAFETY-GRADE ANTICIPATORY REACTOR TRIPS Applicants with B&W-designed plants shall address the requirements set forth ir the Commission Orders regarding hard-wired, safety grade anticipatory reactor trips. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the i construction permit stage of review. Where new designs are involved, appli-cants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.K.2.13 THERMAL-MECHANICAL REPORT - EFFECT OF HPI ON VESSEL INTEGRITY FOR SMALL-BREAK LOCA WITH NO AFW Applicants with B&W-designed plants shall address the requirements set forth in the Commission Orders regarding a thermal-mechanical report on the effect of high pressure injection on vessel integrity for the case of a small-break loss-of-coolant-accident with no auxiliary feedwater. Applicants with 4 B&W-designed plants shall provide sufficient information to describe the D-18

t i nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.2.14 DEMONSTRATE THAT PREDICTED LIFT FREQUENCY OF PORVs AND SVs IS ACCEPTABLE 1 Applicants with B&W-designed plants shall address the requirements set forth in the Commission Orders regarding demonstration that the predicted lift frequency of power operated relief valves and safety valves is acceptable. I Applicants with B&W-designed plants shall provide sufficient information to describe the nature of the. studies, how they are to be conducted, the com-pletion dates, and the program to assure that the results of such studies are factored into the final designs. i II.K.2.15 ANALYSIS OF EFFECTS OF SLUG FLOW ON ONCE-THROUGH STEAM GENERATOR TUBES AFTER PRIMARY SYSTEM VOIDING Applicants with B&W-designed plants shall address the requirements set forth in the Commission Orders regarding analysis of the effects of slug flow on once-through steam generator tubes after primary system voiding. Applicants i with B&W-designed plants shall provide sufficient information to describe the nature of the studies, hew they are to be conducted, the completion dates, and j the program to assure that the results of such studies are factored into the j final designs. II.K.2.16 IMPACT OF RCP SEAL DAMAGE FOLLOWING SMALL-BREAK LOCA hITH LOSS OF 4 0FFSITE POWER i Applicants with B&W-designed plants shall address the requirements set forth in the Commission Orders regarding the impact of reactor coolant pump seal damage following a small break loss-of-coolant accident with loss of offsite l power. Applicants with B&W-designed plants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the com-pletion dates, and the program to assure that the results of such studies are factored into the final designs. D-19

II.K.3.2 REPORT ON OVERALL SAFETY EFFECT OF PORV ISOLATION SYSTEM Applicants with PWR plants shall address the requirements set forth in Items 3.2.4.e and 3.2.4.f of NUREG-0611. Applicants with PWR plants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.5 CONTINUE TO STUDY NEED FOR C.1.4.c AND NEED FOR AUTOMATIC TRIP OF RCPs, THEN MODIFY PROCEDURES OR DESIGNS AS APPROPRIATE Applicants with PWR plants shall address the requirements set forth in Item 3.2.2.a of NUREG-0611. Applicants with PWR plants shall submit, prior to the issuance of construction permits or manufacturing license, a general discussion of how the requirements will be met. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. ~ II.K.3.11 CONTROL USE OF PORV SUPPLIED BY CONTROL COMPONENTS, INC. UNTIL FURTHER REVIEW COMPLETE Applicants with PWR plants shall address the applicable requirements set forth in Item 3.2.4.d of NUREG-0611. Applica'nts with PWR plants shall submit, prior to the issuance of construction permits or manufacturing license, a gen-eral discussion of how the requirements will be met. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. II.K.3.13 SEPARATION OF HPCI AND RCIC SYSTEM INITIATION LEVELS - ANALYSIS AND IMPLEMENTATION Applicants with BWR plants shall address the requirements set forth in Item A.1 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. D-20

i l II.K.3.16 MODIFICATION OF ADS LOGIC - FEASIBILITY STUDY AND MODIFICATION FOR INCREASED DIVERSITY FOR SOME EVENT SEQUENCES Applicants with BWR plants shall address the requirements set forth in Item A.4 of NUREG-0626. Applicants shall provide sufficient information to descrifie the nature of the studies, how they are to be conducted, the completion dabs, and the program to assure that the results of such studies are factored 'ito the final designs. II.K.3.18 MODIFICATION OF ADS LOGIC - FEASIBILITY STUDY AND MODIFICATION FOR INCREASED DIVERSITY FOR SOME EVENT SEQUENCES Applicants with BWR plants shall address the requirements set forth in Item A.7 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.21 RESTART OF CORE SPRAY AND LPCI SYSTEMS ON LOW LEVEL - DESIGN AND MODIFICA110N Applicants with BWR plants shall address the requirements set forth in Item A.10 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.23 ENTRAL WATER LEVEL RECORDING Applicants with BWR plants shall address the requirements set forth in Item B.2 of NUREG-0626. Applicants shall implement design modifications as necessary to meet the requirements. Applicants shall submit, prior to the issuance of construction permits, a general explanation of how the requirements will be met. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. D-21

l II.K.3.24 CONFIRM ADEQUACY OF SPACE COOLING FOR HPCI AND RCIC SYSTEMS Applicants with BWR plants shall address the requirements set forth in Item B.3 -of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.25 EFFECT OF LOSS OF AC POWER ON PUMP SEALS Applicants with BWR plants shall address the requirements set forth in Item B.4 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.28 STUDY AND VERIFY Q'JALIFICATION OF ACCUMULATORS ON ADS VALVES Applicants with BWR plants shall address the requirements set forth in Item B.7 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.30 REVISED SMALL-BREAK LOCA METHODS TO SHOW COMPLIANCE WITH 10 CFR 50, APPENDIX K Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the Action Plan Item II.K.3.31. II.K.3.31 PLANT-SPECIFIC CALCULATIONS TO SHOW COMPLIANCE WITH 10 CFR 50.46 Applicants shall address the applicable requirements of this Action Plan item which is contingent upon the completion of the applicable requirements of D-22 i

Action Plan Item II.K.3.30. In addition, applicants shall address the requirements set forth in Item B.10 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to-be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.44 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE FAILURE TO VERIFY NO SIGNIFICANT FUEL FAILURE Applicants with BWR plants shall address the requirements set forth in Item A.14 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.45 EVALUATE DEPRESSURIZATION WITH OTHER THAN FULL-ADS Applicants with BWR plants shall address the requirements set' forth in Item A.15 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. II.K.3.46 RESPONSE TO LIST OF CONCERNS FROM ACRS CONSULTANT Applicants with BWR plants shall address the requirements set forth in Item A.17 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs. It is noted that the General Electric Company has provided a generic response to the Michelson concerns as they relate to BWRs by letter dated February 21, 1980. Applicants may elect to incorporate by reference the General Electric Company generic response for their own plant designs. Applicants electing this option will, however, include in their responses an assessment of the applicability and adequacy of the General Electric Company generic response to their own plant designs. D-23

III.A.1.1 UPGRADE EMERGENCYJ REPAREDNESS Applicants shall submit, prior to the issuance of construction permits, a discussion of preliminary plans for coping with emergencies addressing the actions described in SECY 79-450 and NUREG-0654. The seven actions to be addressed are described in Section C.1.a of this Action Plan Item. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly. III.A.1.2 UPGRADE LICENSEE EMERGENCY SUPPORT FACILITIES Applicants shall comply with the requirements relating to the Technical Support Center, Operational Support Center, the Emergency Operations Facility. Applicants shall, to the extent possible, provide preliminary design informa-tion at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is tech-nically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. III.A.1.1 UPGRADE EMEP.GENCY PREPAREDNESS Applicants shall submit, prior to the issuance of construction permits, a discussion of preliminary plans for coping with emergencies, addressing the guidance and criteria described in this Action Plan Item as they apply to construction permit applications. Sufficient detail shall Se presented to I provide reasonable assurance that the requirements will be implemented properly. III.A.1.2 UPGRADE LICENSEE EMERGENCY SUPPORT FACILITIES Applicants shall address the requirements for a Technical Support Center, Operational Support Center and Emergency Operations Facility. Applicants D-24

i shall, to the extent possible, provide prelimi_ nary design information at a l level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Appli-cants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. III.A.2.1 AMEND 10 CFR 50 AND 10 CFR 50, APPENDIX E Applicants shall submit, prior to the issuance of construction permits, a discussion of preliminary plans for coping with emergencies addressing the amended rule as it applies to construction permit applications. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly. III.A.2.2 DEVELOPMENT OF GUIDANCE AND CRITERIA Applicants shall submit, prior to the issuance of construction permits or manufacturing license, a discussion of preliminary plans for coping with emergancies addressing the guidance and criteria described in Section B.2.a of this Action Plan item as they apply to the respective applications. Suf-ficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly. III.A.3.3 COMMUNICATIONS Applicants shall include provisions in their designs for prompt notification of, and reliable continuous communications with, the NRC Operations Center in the event of accidents and other emergencies. This will involve installa-tion of dedicated telephone lines and possibly a high-frequency radio backup system. D-25

Applicants will submit prior to the issuance of construction permits a general discussion of how requirements in this regard will be met. Sufficient detail shall be presented to provide reasonable assurance that those requirements will be implemented properly prior to the issuance of operating licenses. III.D.1.1 PRIMARY COOLANT SOURCES OUTSIDE THE CONTAINMENT STRUCTURE NRC is studying the need for improved acceptance criteria for systems outside containment that contain (or might contain) radioactive material either during normal operations or following an accident. These studies are to be completed in early 1981, and these matters will be included in the degraded-core rulemaking proceeding. Applicants shall review the designs of such systems outside containment, and their provisions for leakage control and detection, overpressurization design, discharge points for waste gas venting systems, etc., with the goal of minimiz-ing the possibility of exposure to workers and public during normal operations and in the event of an accident. In this regard, applicants shall submit, prior to the issuance of construction permits, a general discussion of their approach to minimizing leakage from such systems outside containment, in sufficient detail to provide reasonable assurance that this objective will be met satisfacto'rily prior to issuance of operating licenses. III.D.1.2 RADI0 ACTIVE GAS MANAGEMENT Applicants shall address the feasibility of alternative methods / designs for handling large quantities of radioactive noble gases following an accident. Applicants shall submit, prior to issuance of construction permits, a general discussion of their approach to implementing such alternative methods / designs, and information sufficient to provide reasonable assurance that the implerrenta-tion of such alternative methods / designs for handling large amounts of radioactive gases will not be precluded by issuance of construction permits or manufacturing license. D-26

III.D.1.3 VENTILATION SYSTEM AND RADI0 IODINE ADSORBER CRITERIA Applicants shall review their designs to assure that adequate provisions are cade in their facilities for filtration of radioactivity in ventilation exhausts, and that acceptable collection efficiencies of radioiodine adsorb-crs are maintained during accident conditions. NRR will issue requirements for charcoal adsorber upgrading and surveillance testing and may issue requirements for overall improvement of ventilation filtration system designs (e.g., damper design) based on the results of ongoing and planned studies. Applicants shall submit, prior to issuance of construction permits or the manufacturing license, a general discussion of their approach to providing adequate filtration in ventilation exhausts (including surveillance testing) in order to provide reasonable assurance that NRC criteria / requirements in existence prior.to the issuance of construction permits will be implemented properly prior to issuance of operating licenses. Applicants shall also provide information regarding possible improved filtration methods / designs in sufficient detail to provide reasonable assurance that NRC criteria / require-ments in existence prior to issuance of construction permits or manufacturing license will be implemented properly prior to issuance of operating licenses. III.D.3.1 RADIATION PROTECTION PLANS Applicants shall address Action Plan requirements regarding radiation protection plans which will keep exposures to workers as low as reasonably achievable during both normal operation and accident conditions, and which would allow plant workers to take effective action to control the course and consequences of an accident. A general explanation of how the requirements will be met is required prior to issuance of the construction permits or the c:nufacturing license. Sufficient detail shall be presented to provide reasonable assurance that the requirements will be imple; rented properly. III.D.3.3 IN-PLANT RADIATION MONITORING Applicants shall review their designs to assure that provisions for monitoring inplant radiation and airborne radioactivity are appropriate for a broad range D-27

l of routine and emergency conditions. Applicants shall, to the extent possible, i provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. III.D.3.4 CONTROL ROOM HABITABILITY Applicants shall review the design of their facilities for conformance to requirements stated in the Action Plan. NRC will consider possible new cri-teria to preclude control room contamination via potential internal pathways indicated by the TMI-2 experience. Applicants shall address prior to issuance of the construction permits or manutacturing license, how they will implement the existing requirements set forth in this Action Plan item. Applicants shall also address the extent tc which improvements have been made to prevent control room contamination via pathways not,previously considered. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the t supporting design bases and criteria. Applicants shall also demonstrate that the design concept is ter.iinically feasible and within the state of the art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. 0-28 -}}