ML20009F405
| ML20009F405 | |
| Person / Time | |
|---|---|
| Issue date: | 06/30/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20009F403 | List: |
| References | |
| NUREG-0718, NUREG-0718-R01, NUREG-718, NUREG-718-R1, NUDOCS 8107310111 | |
| Download: ML20009F405 (64) | |
Text
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NUREG-0718 Rev.1 Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License l
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[J.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation
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NUREG-0718 Rev.1
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Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License ateYu shed un 1 Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Weshington, D.C. 20565 p
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ABSTRACT The 1MI-2 Action Plan, NUREG-0660, does not specifically address requirements l
for construction permit and manufacturing license applications.
There are currently pending six construction permit applications for eleven units with light water reactors and one manufacturing license application for eight floating nuclear plants.
Staff review of these applications had been suspended since the 1MI-2 accident pending the formulation of a policy to appropriately reflect the lessons learned from the accident.
The Comcission is considering a new rule which will state the TMI-related requirements.to be applied to these applications.
NUREG-0718 was issued, and has now been revised, to provide guidance that the NRC staff believes should be followed to account for the lessons learned from the TMI-2 accident.
NUREG-0718 is not a substitute for.the regulations, and compliance is not a requirement.
However, an approach or method different from the guidance contained herein will be accepted only if the substitute approach or method provides an equivalent basis for meeting the requirements.
iii
TABLE OF CONTENTS Section Title Page Abstract...
iii I
Introduction...................................... '..
1 II Assessment of TMI-2 Action Plan for Pending CP and ML Applications......................................
2 Appendix A Requirement Category Assignments for Pending Construction Permit and Manufacturing License Applications..........
A-1 Appendix B Information Requirements for TMI-2 Action Plan Items in Categories 3, 4, and 5.........
B-1 v
l I
I.
INTRODUCTION After the accident at'Inree 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.
Furthermore, the Commission decided that power reactor iicensing 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 recommendations 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 Lessoas Learned Task Force led to NUREG-0578, "TMI-2 Lessons Learned Task Force Statis Report and Short-Term Recommendations" and NUREG-0585, "TMI-2 Lessons Learned Task Force Final Report." Following release of the report of the Presi-dential Comm.ssion, the Commission provided a preliminary set of responses to the recommendations in that report.
This response provided broad 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 Special Inquiry Group Report was received, which had the benefit of review by panels of outside con 3ultants representing a cross section of technical and public views.
This report provided additional reccmmendations.
NUREG-0660, "NRC Action Plan Demloped 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 investi-gations of the accident.
In developing the Action Plan, the various recommend-ations 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 immediately 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 immediately after the the accident, the first report of the Lessons Learned Task Force issued in July 1979, the recommendations of the Emergency Preparedness Task Force, and the NRC staff and Commission.
Before these immediate actions were applied to operating plants, they were approved by the Commission.
Most of the required immediate actions have already been taken by licensees.
On February 7, 1980, based on its review of initial drafts 30f the Action Plan, the Commission approved a listing of near-term operating license (NT0L) require-ments as being necessary, but not necessarily sufficient, TMI-related requirements for granting new operating licenses.
The fuel load requirements on the NTOL list were used by the Commission in granting operating licenses for three plants, with limited authorizations for fuel loading and low power testing. -
On May 15, 1980, after review of the last version of the Action Plan, the Commission approved a list of " Requirements for New Operating Licenses," now contained in NUREG-0694, which the staff recommended for imposition on current operating license applications.
That list was recast from the previous NT0L list and sets forth the TMI-related requirements and actions for new operating licenses.
In a Statenent of Policy issuad on June 16, 1980, the Commission concluded that the 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 applica-tions should be measured against the regulations, as augmented by these require-
- rents, Subsequently, the staff incorporated all of the THI-related items for operating reactor licensees and operating license applicants in one document, NUREG-0737, which was reviewed and approved by the Commission on October 28, 1980.
This report was issued by letter on October 31, 1980; and the Commission issued a Statement of Policy on December 16, 1980, adopting NUREG-0737 in piace of NUREG-0694 for operating license apr lications.
The THI-2 Action Plan, NUREG-0660 does not specifically address requirements for constre 1 permits (CP) or manufacturing license (ML) applications.
There are current pending six CP applications for eleven units with light water reactors ano one ML application for eight floating nuclear plants.
The NRC staff review of these applications had been suspended since the TMI-2 accident pending the formulation 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 of TMI-2 related requirements for these applications.
Those requirements are describad in NUREG-0718, which was issued in & aft for public comment in August 1980, and in the final report dated Marcn 1981.
Subsequently, some revisions in 1.he requirements have been made by the staff in the course of develcping a praposed rule for the pending applications.
The items in this document include the revisions.
II. ASSESSMENT OF TMI-2 ACTION PLAN FOR PtNDING CP AND ML APPLICATIONS In order to assess the extent to which the TMI-2 Action Plan should be implemented on the seven pending CP and ML applications, the staff developed five requirement categories.
Each of the TMI ~ Action Plan requirements was carefully evaluated and then assigned to one of these five categories.
A discussion of each of the requirement categories follows.
Category 1 A requirement of a type not applicable to the pending CP or 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.
't does not apply to plants of the type now pending; d.
.t has been (or will be) superseded by another requirement in the A.. ion Plan or in the regulations; e.
It has already been completed. ________ -
Category 2 A requirement of the type customarily left for the operating license stage.
Category 3 Studies (and other research and development activities to provide design l
development information) of the type customarily left for review at the operating license stage.
However, to satisfy 50.35(a)(3) the staff believes that items in this category should be corapleted 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 conCucted, the completion dates, and a program to assure that the results of such studies are factored into the final design.
Category 4 A requirement to demonstrate that any additional design, development and imple-men'.ation necessary to satisfy the requirement (or to satisfy the goals of the trok whose requirements are to be developed in the future) will be satisfactorily cc..jpleted by the operating license stage.
This is the type of information customarily required at the construction permit stage to satisfy 50.35(a)(2),
or to satisfy ALAB-444 with respect to generic issues.
Category 5 A requirement for information of the typ ~ntomarily reviewed at the preliuinary design stage for the following types of items:
l Items for which the required information should be sufficient to demonstrate a.
that the requirement has been satisfied by the application.
This is the.
kind of information and degree of detail customarily provided at the prelim-inary design stage with respect tc 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 i
kind of information and degree of detail customarily provided at the pre-l liminary design stage with rasnect.to the preliminary design of the facility to satisfy 50.34(a)(3)(4), etc.
Tables 1, C.1, C.2, and C.3 from NUREG-0660 list each of the TMI-2 Action Plan requirements.
Appendix A of this report is a reprint of these tables with the NRC staff's category assignments for the pending CP and ML applications.
l Appendix B provides a description of the specific information to be provided by CP and ML applicants for each of the Action Plan requirements assigned to Categories 3, 4, and 5.
APPENDIX A REQUIREMENT CATEGORY ASSIGNMENTS FOR PENDING CONSTRUCTION PERMIl AND MANUFACTURING LICENSE APPLICATIONS l
TABLE 1 - PRIORlTIES 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 2.
Shift Supervisor Admin. Duties 2/lc 3.
Shift Manning 2/1/c 4.
Long-Term Upgrading Ib/lb Refer to Action Plan Item I.B.1.1 I.A.2 Training and Qualifications of Operating Personnel 1.
Immediate Upgrading of Operating Id/lc Refer to Action Plan Item I.B.1.1 and Senior Operator Train'ng and Qualifications 2.
Training and Qualifications Id/lc Refer to Action Plan Item I.B.1.1 of Operations Personcel 3.
Administration of Training lb/lc Programs for Licensed Operators
u TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML 4.
NRR Participation in IE lb/lc Inspector Training 5.
Plant Drills 2/lc 6.
Long-Term Upgrading of Training Id/lc Refer to Action Plan Item I.B.1.1 and Qualifications 7.
Accreditation of Training Ib/lc Institutions T
I.A.3 Licensing and Requalification N
of Operating Personnel 1.
Revise Scope and Criteria 2/lc Refer to Action Plan Item I.A.3.2 l
for Licensing Exams 2.
Operator Licensing Program Ib/lc Changes 3.
Requirements for Operator lb/lc Fitness 4.
Licensing of Additional Ib/lc Operations Personnel 5.
Establish Statement of Under-Ib/lc standing with INP0 and DOE I.A.4 Simulator Use and Development 1.
Initial Simulator Improvement ld/lc Refer to Action Plan Item I.A.4.2
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML l
2.
long-Term Training Simulator 4/lc Refer to Appendix B Upgrade 3.
Feasibility Study of Procure-lb/lb l
ment of NRC Training Simulator l
l 4.
Feasibility Study of NRC lb/lb Engineering Computer I.B Support Personnel I.B.1 Management for Operations 1.
Organization and Management 2/la Long-Term Improvements 2.
Evaluation of Organization Id/lc Refer to Action Plan Item I.B.1.1 and Management Improvements of NT0L Applicants 3.
Loss of Safety Function Ib/lc ll I.B.2 Inspection of Operating Reactors f
1.
Revise IE Inspection Program lb/lc 2.
Resident Inspector at Operating Ib/lb Reactors 3.
Regional Evaluations Ib/lb 4.
Overview of Licensee Performance Ib/lb
TABLE 1 (Conticued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML I.C Operating Procedures 1.
Short-Term Accident Analysis 2/lc and Procedures Revision 2.
Shift and Relief Turnover 2/lc Procedures 3.
Shift Supervisor Responsibilities 2/lc 4.
Control Room Access 2/lc T
5.
Procedures for Feedback of 5/5 Refer to Appendix B Operating Experience 6.
Procedures for Verification 2/lc of Correct Performance of Operating Activities
- 7. 'NSSS Vendor Review of Procedures 2/lc 8.
Pilot Monitoring of Selected 1b/lc Emergency Procedures for NT0L Applicants 9.
Long-Term Program Plan for 4/lc Refer to Appendix B Upgrading _of Procedures I. D Control Room Design 1.
Control Room Design Reviews 4/4 Refer.to Appendix B 2.
Plant Safety Parameter Display 4/4 Refer.to Appendix B Console
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML 3.
Safety System Status Monitoring 4/4 Refer to Appendix B 4.
Control Room Design Standard lb/lb 5.
Improved Control Room Ib/lb Instrumentation Research 6.
Technology Transfer Conference Ib/lb I.E Analysis and Dissemination of Operating Experience E
1.
Office for Analysis and Evalua-Ib/lb tion of Operation Data 2.
Program Office Operational Data Ib/lb Activities 3.
Operational Safety Data Analysis Ib/lb
.4.
Coordination of Licensee, Industry, Id/1d and Regulatory Programs 5.
Nuclear Plant Reliability Data Ib/lb System 6.
Reporting Requirements 1b/lb 7.
Foreign Sources 1b/lb 8.
Human Error Rate Analysis 1b/lb
c TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML I. F Quality Assurance 1.
Expand QA List 5/5 Refer to Appendix B 2.
Develop More Detailed QA Criteria 5/5 Refer to Appendix B 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 Ib/lc 2.
Site Evaluation of Existing Id/lc Refer to Appendix B Facilities i
1 II.B Consideration of Degraded or i
Melted Cores in Safety Review 1.
Reactor Coolant System Vents 4/4 Refer to Appendix B 2.
Plant Shielding to Provide 4/4 Refer to Appendix B Access to Vital Areas and Protect Safety Equipment for Post-accident Operation 3.
Post-accident Sampling 4/4 Refer to Appendix B
TABLE 1 (Continued)
REQUIREKENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML 4.
Training for Mitigating Core 2/lc Damage 5.
Research on Phenomena Associated Ib/lb with Core Degradation and Fuel Melting 6.
Risk Reduction for Operating Ic/lc Reactors at Sites.with High Population Densities
[4 7.
Analysis of Hydrogen Control ld/1d Refer to Action Plan Item II.B.8 8.
Rulemaking Proceeding on Degraded 5/5
-Refer to Appendix B Core Accidents II.C Reliability Engineering and Risk Assessment 1.
Interim Reliability Evaluation Ib/lb
~ Program (IREP) 2.
Continuation of IREP lb/lb 3.
Systems Interaction Ic/lc 4.
Reliability Engineering Id/1d
~II.D Reactor Coolant System Relief and Safety Valves 1.
Testing Requirements 4/4 Refer to Appendix B
____ ~ _ _ _ -. _ -.
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION IlEM ASSIGNMENT COMMENTS CP/ML 2.
Research on Relief and Safety Ib/lb Valve Test Requirements 3.
Relief and Safety Valve Position 4/4 Refer to Appendix B Indication II.E
System Design
II.E.1 Auxiliary Feedwater System 1.
Auxiliary Feedwater System 3/3 Refer to Appendix B
'E Evaluation co i
2.
Auxiliary Feedwater System 4/4 Refer to Appendix B Automatic Initiation and Flow Indication 3.
Update Standard Review Plan Ib/lb and Develop Regulatory Guide II.E.2 Emergency Core Cooling System 1.
Reliance on ECCS 2/lb 2.
Research on Small Break LOCAs Ib/lb and Anomalous Transients 3.
Uncertainties in Performance Ib/lb Predictions t
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML II.E.3 Decay Heat Removal 1.
Reliability of Power Supplies 4/4 Refer to Appendix B for Natural Circulation 2.
Systems Reliability lb/lb 3.
Coordinated Study of Id/1d Refer to Action Plan II.C.4 Shutdown Heat Removal 4.
Alternate Concepts Research Ib/lb
?"
5.
Regulatory Guide 1b/lb e
II.E.4 Containment Design 1.
Dedicated Penetrations 5/5 Refer to Appendix B 2.
Isolation Dependability 4/4 Refer to Appendix 8 3.
Integrity Check 2/lc l
4.
Purging 4/4 Refer to Appendix B 1
II.E.5 Desi0n Sensitivity of B&W Reactors 1.
Design Evaluation 4/lc Refer to Appendix B I
- 2.
B&W Reactor Transient Response Id/lc Deleted - future requirements will be established if necessary Task Force 1
._. =
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML II.E.6 In-situ Testing of Valves 1.
Test Adequacy Study Ib/lb II.F Instrumentation and Control 1.
Additional Accident Monitoring 4/4 Refer to Appendix B Instrumentation 2.
Identification of and Recovery 4/4 Refer to Appendix B from Conditions Leading to T
Inadequate Core Cooling l
3.
Instrumentation for Monitoring 4/4 Refer to Appendix B Accidtat Conditions (Reg.
Guide 1.17) 4.
Study of lontrol and. Protective Ib/lb Actier i;ign Requirements 5.
Cla_
'i'_*.ior of Instrumentation, Ib/lb conti.'. ano E'ectric Equipment II.G Electrical Power 1.
Power Supplies for Pressurized
.4/4 Refer to Appendix B Relief Valses, Block-Valves, i
and Level Indicators II.H TMI-2 Cleanup and Examination l
1.
Maintain Safety of TMI-2 and Ib/lb j
Minimize Environmental Impact
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML 2.
Obtain Technical Data on the Ib/lb Conditions Inside the TMI-2 Containment Structure 3.
Evaluate and Feedback Infor-lb/lb mation Obtained from TMI 4.
Determine Impact of TMI on Ib/lb Socioeconomic and Real Property Values T
II.J General Implications of TMI for U
Design.and Construction Activities II.J.1 Vendor Inspection Program 1.
Establish a Priority System for Ib/lb Conducting Vendor Inspections 2.
Modify Existing Vendor Inspection Ib/lb
-Program i
3.
Increase Regulatory Control lb/10 Over Present-Non-licenses 4.
Assign Residents Inspectors to Ib/lb Vendors and Architect-Engineers II.J.2 Construction Inspection Program 1.
Reorient Inspection Program lb/lb'
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML 2.
Increase Emphasis on Independent Ib/lb Measurem':nt in the Construction Inspection Program 3.
Assign Resident Inspectors to Ib/lb l
all Construction Sites l
II.J.3 Management for Design and Construction 1.
Organization and Staffing to 5/5 Refer to Appendix B T
Oversee Design and Construction M
2.
Issue Regulatory Guide Ib/lb II.J.4 Revise Deficiency Reporting Requirements 1.
Revise Deficiency Reporting lb/lb Requirements II.K Measures to Mitigate Small-8reak LOCAs and Loss of Feedwater Accidents s
1.
IE Bulletins See Table 1A 2.
Commission Orders on B&W plants See Table 18 3.
Final Recommendations of B&O See Table IC Task Force
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML III.
Emergency Preparations and Radiation Effects III.A NRC and Licensee Preparednesc III.A.1 Improve Licensee Emergency Preparedness - Short-term 1.
Upgrade Emergency Preparedness Ib/lb 2.
Upgrade Licensee Emergency 4/4 Refer to Appendix B;
{
Support Facilities w
3.
Maintain Supplies of Thyroid 2/lc Refer to NUREG-0654, Rev. 1; Stockpiling Blocking Agent (Potassium for the general public is under consideration Iodide)
III.A.2 Improving Licensee Emergency Preparedness - Long-term 1.
Amend 10 CFR 50 and 10 CFR 50, Id/1d Complete.
Refer to amended regulations.
Appendix E 2.
Development of Guidance and le/lb Criteria III.A.3 Improving NRC Emergency Preparedness 1.
NRC Role in Responding to lb/lb Nuclear Emergencies 2.
Improve Operations Centers 1b/lb
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML 3.
Communications ld/lc Refer to Appendix B 4.
Nuclear Data Link lb/lb 5.
Train.ob, Drills, and Tests 2/lc 6.
Interaction of NRC with Other lb/lb Agencies T"
III.8 Emergency Preparedness of State and
}l Local Governments 1.
Transfer of Responsibilities Ib/lb to FEMA 2.
Implementation of NRC's and Ib/lb FEMA's Responsibilities III.C Public Information 1.
Have Information Available for Ib/lb i
the News Media and the Public 2.
The Office of Public Affairs Ib/lb will Develop Agency Policy and Provide Training for Inter-facing with the News Media and Other Interested Parties l
l
_o
TABLE 1 (Continued) l I
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML
,III.D Radiation Protection III.D.1 Radiation Source Control 1.
Primary Coolant Sources 4/4 Refer to Appendix B Outside the Containment Structure L
2.
Radioactive Gas Managemeat Ib/lb 3.
Ventilation System and Ib/lb j[
Radioicdine Adsorber Criteria v
4.
Radwaste System Design Features 1b/lb to Aid in Accident Recovery and Decontamination III.D.2 Public Radiation Protection Ieprovement 1.
Radiological Monitoring of Ib/lb Effluents 2.
Radioiodine, Carbon-14, and lb/lb Tritium Pathway Dose Analysis 3.
Liquid Pathway Radiological 2/lb Control 4.
Offsite Dose Measurements 2/lc 5.
Offsite Dose Calculation Manual 2/lb i
~_ _
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML 6.
Independent Radiological Ibfl.h Measurements 4
III.D.3 Worker Radia^. ion Protection Improvements 1.
Radiation Protection Plans 2/2 2.
Health. Physics Improvements Ib/lb 3.
Inplant Radiation Monitoring 4/4 Refer to Appendix B
$(
4.
Control Room Habitability 4/4 Refer to Appendix B 5.
Radiation Worker Exposure Data la/lc i
Base I
IV.
Practices and Procedures IV.A Strengthen Enforcement Process 1.
Seek Legislative Authority lb/lb l
2.
Revise Enforcement Policy lb/lb IV.B Issuance of Instructions and Information to Licensees IV.B.1 Revise Practices for Issuance of Ib/lb Instructions and Information to Licensees
TABLE 1 (Continued) 4 REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML IV.C Extend Lessons Learned to Licensed Activities Other than Power Reactors IV.C.1 Extend Lessons Learned from TMI to Ib/lb other NRC Programs IV.D NRC <taff Training IV.D.1 NRC Staff Training Ib/lb IV.E Safety Decision-Making
. tj 1.
Expand Research on Quantifica-Ib/lb tion of Safety Decision-Making 2.
Plan for Early Resolution of Ib/lb Safety Issues 3.
Plan for Resolving Issues at 1b/lb Construction Permit Stage 4.
Resolve Generic Issues by lb/lb Rulemaking t
5.
Assess Corrently Operating lb/lb Reactors l
IV.F Financial Disircentive to Safety 1.
Incres:ed IE Scrutiny of Power Ib/lb Ascension Test Program i
. _. _ _.. _ _. ~ _
l TABLE 1 (Continued)
REQUIREMENT CATEGORY l
ACTION ITEM ASSIGNMENT COMMENTS CP/ML 2.
Evaluate the Impact of Financial Ib/lb Disincentives to the Safety of Nuclear Power Plants IV.G Improve Safety Rulemaking Procedures i
1.
Develop a Public Agenda for Ib/lb Rulemaking 2.
Periodic and Systematic lb/lb Reevaluation of Existing Rules as 3.
Improve Rulemaking Procedures Ib/lb 4.
Study Alternative for Improved Ib/lb Rulemaking Process j
l IV.H NRC Participation in the Radiation Policy Council V.
NRC Folicy, organization and Management 1.
Develop NRC Policy Statement Ib/lb on Safety 2.
Study Elimination on Non-safety Ib/lb l
Responsibilities i
3.
Strengthen Role of ACRS lb/lb 4.
Study Need for Additional Ib/lb Advisory Committees i
i I
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML 5.
Improve Public and Intervenor Ib/lb Participation in Hearing Process 6.
Study Construction-During-Ib/lb Adjudication Rules 7.
Study Need for TMI-Related Ib/lb Legislation 8.
Study the Need to Establish an Ib/lb Independent Nuclear Safety Board in 9.
Study the Reform of the Licensing Ib/lb Process
- 10. Study NRC Top Management Ib/lb Structure and Process
- 11. Reexamine Organization and Ib/lb Functions of NRC Offices
- 12. Revise Delegations of Ib/lb Authority to Staff
- 13. Clarify and Strengthen the Ib/lb Respective Roles of Chairman, Commission, and EDO
- 14. Authority to Delegate Emergency Ib/lb Response Functions to a Single Comissioner
TABLE 1 (Continued)
REQUIREMENT CATEGORY ACTION ITEM ASSIGNMENT COMMENTS CP/ML
- 15. Achieve Single Location -
Ib/lb Long-term
- 16. Achieve Single Location - Interim Ib/lb
- 17. Reexamine Commission Role in Ib/lb Adjudication T
a
4 TABLE 1A 0FFICE OF INSPECTION AND ENFORCEMENT BULLETINS CP/ML Requirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 1.
Review TMI-2 PNS and detailed chronology 79-50&-5A (Item 1)
BWR and PWR Id/1d Refer to Action Plan of IMI-2 accident.
79-06&O6A (Item 1)
Items 1.A.2.2 and 79-06&O68 (Item 1)
I.A.3.1 2.
Review transients similar to TMI-2 that 79-05&05A (Item 2)
B&W ld/lc Refer to Action Plan have occurred st other facilities and Items I.A.2.2 and NRC evaluation of Davis-Besse transient.
I.A.3.1 3.
Review operating procedures for recog-79-05&05A (Item 3)
PWR Id/lc Refer to Action Plan hizing, preventing, and mitigating void 79-06&O6A (Item 2)
Item I.C.1 T
inrmation in transients and accidents 79-06&O6B (Item 2)
U 4.
Review operating procedures and training 79-05&05A (Item 4.a) PWR and BWR Id/lc 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 E5F 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 Id/lc Refer to Action Plan NUREG-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
l TABLE 1A (Continued)
CP/ML Requirement l
Source for Category Requirement Operating Reactors Applicability Assignment Comments i
4 i
c.
RCP operation NUREG-0623 PWR Id/lc Refer to Action Plan Items I.A.1.3 and I.C.1 L
- d. Operators are inicructed not to rely 79-05A (Item 4.d)
PWR and BWR Id/lc Refer to Action Plan on level indication alone in 79-06A (Item 7.d)
Items I.C.1, I.A.3.1, evaluating plant conditions.79-068 (Item 6.d) and II.F.2 79-08 (Item 5.b) i'5.
Safety related valve position 79-05&05A (Item 5)
PWR and BWR Id/ld Refer to Action Plan S$
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 Id/ld Refer to Action Plan design and procedures. Assure isolation 79-06A (Item 4)
Item II.E.4.2 of all lines that de not degrade safety 79-ObB (Item 3) features or cooling 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.
Yr--
l TABLE 1A (Continued)
CP/ML Requirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 8.
Immediately implement procedures that 79-05A (Item 8)
B&W Id/lc Refer to Action Plan assure two independent 100% AFW flow Item II.E.1.1 paths, or specify explicitly LCO with reduced AFW capacity.
9.
Review procedures 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 79-06B (Item 8)
(especially upon ESF reset). List all 79-08 (Item 7) applicable systems and interlocks.
2-4 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 79-06B (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 BWR Id/lc Refer to Action Plan personnel aware of the seriousness79-06A( Item 1.a)
Items I.A.3.a and and consequences of the erroneous79-06B (Item 1.a)
I.A.2.2 actions taken leading up to, and in early phases of, the TMI-2 accident.
- 12. One hour notification requirement, and 79-05B (Item 6)
PWR and BWR Id/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)
I TABLE 1A (Continued)
Co/ML Requirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 13.
Propose Technical Specification changes79-05B (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 Id/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 2 15.
For facilities with non-automatic AFW 79-06A (Item 5)
W & CE Id/lc Refer to Action Plan n'
initiation, provide dedicated operator 79-068 (Item 4)
Item II.E.1.2 o
in continuous communication with 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)
Lo Level coincidence) will initiate safety injection. For test, reset to 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
.\\
TABLE 1A (Continued)
CP/ML Requiremen'.
Source for Category Requirement Operating Reactors Applicability Assignment Comments 19.
Describe design an procedure 79-05B (Item 3)
B&W Id/lc Refer to Action Plan modifications (based on analysis)
Item II.E.5 to reduce likelihood of automatic PZR PORV actuation in transients.
20.
Provide procedures and training to 79-05R (Item 4)
B&W 2/lc operators for prompt manual reactor trip for LOFW, IT, MSIV closure, LOOP, LOSG Level, & Lo PZR Level.
?g 21.
Provide automatic safety grade 79-058 (Item 5)
B&W id/lc Refer to Action anticipatory reactor trip for Plan Item II.K.2.10 LOFW, TT, or significant decrease in SG lesel.
22.
Describe automatic and manual actions 79-08 (Items 3)
BWR 4/lc Refer to Appendix B 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 Id/1d 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.1 time lapses between reactor trip 79-06C (short-and RCP trip, term Item 2)
TABLE 1A (Continued) i CP/ML Requirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 25.
Develcp operator action guidelines79-05C (short-PWR Id/1d Refer to Action Plan (based on analyses in Requirement 24 term Item 3)
Item I.C.1 above).79-06C (short-term Item 3) 26.
Revise emergency procedures and train 79-05C (short-PWR Id/lc Refer to Action Plan R0s-and SR0s based on guidelines term Item 4)
I.C.1, I.A.3.a, and developed in Requirement 25 above.79-06C (short-I.G.1 term Item 4)
T 27.
Provide analyses and develop guidelines79-05C (short-PWR Id/1d Refer to Action Plan St and procedures for inadequate core term Item 5)
-Items I.C.1 and II.F.2 cooling conditions. Also, define RCP 79-06C (short-restart criteria.
tere Item 5) 28.
Provide design that will assure NUREG-0623 PWR Id/ld Refer to Action Plan r
automatic RCP trip for all Item II.K.3.5 circumstances where required.
t i
r y,-
~
- ~.
~..
TABLE IB 9EQUIREMENTS FOR NEW B&W PLANTS DERIVED FROM COMMISSION ORDERS ON OPERATING B&W PLANTS CP/ML Requirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 1.
Upgrade timeliness and reliability Commission Order B&W Id/lc Refer to Action Plan Item II.E.1 c' AFW system.
2.
Procedures and training to initiate Commission Order B&W id/lt and control AFW independent of integrated contrnl systen.
3.
Hard-wired control grade anticipatory Commission Order B&W id/lc Refer to Action Plan Item II.K.2.10 reactor trips.
A" 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 1MI-2 simulator training for Commission Order B&W id/lc Refer to Action Plan
}
all operators.
Item I.A.2.6 6.
Reevaluate analysis for dual-level Commission Order Davis-Besse 1 1c/lc setpoint control.
7.
Reeveluate transient of September 24, Commission Ordar Davis-Besse 1 1c/lc 1977.
8 Continued upgtading of AFW system.
Commission Order B&W Id/lc Refer to Action Plan Item II.E.1 9
9
TABLE la (Continued)
CP/HL Requirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 9
Analysis and upgrading of integrated Commission Order B&W 4/lc Refer to Appendix B control system.
10.
Hard-wired safety-grade anticipatory Commission Order B&W 4/1:
Refer to Appendix B reactor trips.
11.
Operator training and drilling.
Commissien Order B&W Id/lc Refer to Action Plan Items I.A.3.1, I.A.2.2, I.A.2.5, and I.G.a P
g; 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 lb/?c of HPI on vessel integrity for small-B&W operating plants, break LOCA with no AFW.
8/121/79 14.
Demonstrate that predicted lift Letter, D. Ross to B&W Id/1d 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 le/le once-through steam generator tubes B&W operating-after primary system voiding.
plants, 8/21/79
TABLE IB (Continued)
CP/ML l
l Requirement Source for Category Requirement Operating Reactors Applicability Assignment Comments 16.
Impact of RCP seil damage following Letter, D. Ross to All 3/3 Refer to Appendix B small-break LOCA with loss of offsite B&W operating power.
Plants, 8/21/79 37.
Analysis of potential voiding in Letter, R. Reid All B&W Id/lc Refer to Action RCS during anticipated transients.
to all B&W operating Plan Item I.C.1 plants 1/9/80 18.
Analysis of loss of feedwater and Letter, D. Ross to All B&W Id/lc Refer to Action other anticipated transients.
B&W operating plants, Plan Item I.C.2 8/21/79 l
e 19.
Benchmark analysis of sequential Letter, D. Ross to All B&W Id/lc Refer to Action AfW flow to once-through steam B&W operating plants, Plan Item I.C.1 generator 8/21/79 20.
Analysis of system response to small-Letter, D. Ross to All B&W Id/lc Refer to Action break LOCA that iuses system pressure B&W operating plants Plan Item I.C.1 to exceed PORV setpoint.
8/?1/79 21.
LOTT 3-1 predictions.
Letter, D. Ross to All B&W 1e/lc l
B&W operating plants, l
8/21/79 l
A
~
TABLE IC FINAL REC 0t94ENDATIONS OF BULLETINS AND ORDERS TASK FORCE CP/ML Requirement Category I
Requireinent Source Applicability Assignment Comments 1.
Install automatic PORV isolation NUREG-0565 (2.1.2.1)
PWR Id/lc 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.t) 2.
Report on overall safety effect NUREG-0565 (2.1.2.d)
PWRs 3/3 Refer to Appendix B of PORV isolation system.
NUREG-0611 (3.2.4.g)
(3.2.4.i)
NUREG-0635 (3.2.4.c)
T U 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)
NUREG-0626 (B.14)
NUREG-0635 (3.2.4.d) l 4.
Review and upgrade reliability NUREG-0565 (2.3.2.b)
All Ib/lb Refer to Action Plan and redundancy of nonsafety NUREG-0611 (3.2.2 b)
Items II.C.1, II.C.2, equipment for emall-break LOCA NUREG-0626 (B.12) and II.C.3 l
l mitigation NUREG-0635 (3.2.2.b) 5.
Continue to study need for NUREG-0565 (2.3.2.a)
PWR lb/lb 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 as appropriate.
TABLE IC (Continued)
CP/ML Requirement Category Requirement Source Applicability Assignment Comments 6.
Instrumentation to verify NUREG-0565 (2.6.2.b)
PWR Id/1d Refer to Action Plan natural circulation.
NUREG-0611 (3.2.3.b)
Item I.C.I II.F.2, NUREG-0635 (3.2.3.b)
II.F.3 1
7.
Evaluation of PORV opening NUREG-0565 (2.1.2.b)
B&W id/lc Refer to Action Plan probability during overpressure Item II.K.2.14 transient.
8.
Further staff consideration of NUREG-C565 (2.5.2.a)
PWR Td/1d Refer to Action Plan need for diverse decay heat NUREG-0635 (4.2.5),
Item II.C.1 and removal method independent App. VIII)
II.E.3.3 4,
of SGs NUREG-0611 (4.2.5, App. VIII) 9.
Proportinnal integral derivative NUREG-0611 (3.2.4.b)
W Ic/2 controller modification.
10.
Anticipatory trip modification 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 id/1d Deleted - this is by Control Components Inc., until covered by Action Plan item II.D.1 further review complete.
- 12. Confirm existence of anticipatory f.UREG-0611 (3.2.4.a)
W Ic/2
)
trip upon turbine trip.
j l
l
TABLE 1C (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 B system initiation levels.
Analysis and implementation.
14.
Isolation of isolation NUREG-0626 (A.2)
GE plants Ic/lc condensers on high radiation.
with isolation condenser
- 15. Modify break detection logic NUREG-0626 (A.3)
GE I/lc to prevent spurious isolation
]
m 16.
Reduction of challenges and NUREG-0626 (A.4)
GE 3/lc Refer to Appendix B failures of relief valves -
feasibility study and system 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 B feasibility study and modifica-tion for increased diversity for some event sequences.
' ' ~ " ' ' " - ' '
TABLE 1C (Continued)
CP/ML Requirement Category Requirement Source Applicability Assignment Comments I
19.
Interlock on recirculation NUREG-0626 (A.8)
GE Non-Jet Ic/lc l
pump loops.
Pump ops 20.
Loss of service water for NUREG-0626 (A.9)
Big Rock 1c/lc Big Rock Point.
Point 21.
Restart of core spray and LPCI NUREG-0626 (A.10)
GE 3/lc Refer to Appendix B systems on low 1cvel - design and modification.
w 22.
Automatic switchover of RCIC NUREG-0626 (B.1)
GE 1c/1c system suction - verity procedures and modify design.
23.
Central water level recording.
NUREG-0626 (B.2)
GE 4/lc Refer to Appendix B 24.
Confirm adequacy of space cool-NUREG-0626 (B.3)
GE 3/1c Refer to Appendix B ing for HPCI and RCIC systems.
25.
Effect of loss of AC power on NUREG-0626 (B.4)
GE 1d/1d Refer to Item II.K.2.16 pump seals.
TABLE IC (Continued)
CP/ML Requirement Category Requirement Source Applicability Assignment Comments 26.
Study effect on RHR reliability NUREG-0626 (B.5)
GE Id/lc Refer to Action Plan of its use for fuel pool cooling.
Item II.E.2.1 27.
Provide common reference level NUREG-0626 (B.6)
GE 2/lc for vessel level instrumentation.
1 28.
Study and verify qualification NUREG-0626 (B.7)
GE 3/lc Refer to Appendix B l
of accumulators on ADS valves.
2 29.
Study to demonstrate perform-NUREG-0626 (B.13)
GE Isolation ic/lc ch ance of isolation condensers Condenser ors
-with noncondensibles.
30.
Revised'small-break LOCA methods NUREG-0565 (2.2.2.a)-
All Ib/lc to show compliance with 10 CFR
'NUREG-0611-(3.2.1.a) 50, Appendix K.
NUREG-0626 (A.12) l NUREG-0635 (3.2.1.a)
(3.2.5.a) 31.
Plant-specific calculations to NUREG-0565 (2.2.2.b)
All-lb/lc show compliance with 10 CFR NUREG-0611 (3.2.1.b) 50.46.
NUREG-0626 (A.13, B.10)
NUREG-0635 (3.2.a.b)
=-
i
.s TABLE IC (Continued)
CP/ML Requirement Category Requirement Source Applicability Assignment Comments 32.
Provide experimental verifica-NUREG-0565 (2.6.2.a)
PWR lb/lb Refer to Action Plan tion of two phase natural NUREG-0611 (3'.2.3.a)
Item II.E.2.2 circulation models.
NUREG-0635 (3.2.3.a) 33.
Evaluate elimination of PORV NUREG-0565 (3.5)
PWR lb/lb Refer to Action Plan function.
NUREG-0611 (3.2.4.k)
Item II.C.1 NUREG-0635 (3.2.4.e) 34.
RELAP-4modeldevhlopment.
NUREG-0611 (3.2.5)
PWR lb/lb Refer to Action Plan NUREG-0635 (3.2.5)
Item II.E.2.2 2
J.
'" 35.
Evaluation of effects of core NUREG-0565 (2.2.2.c)
B&W id/lc Refer to Action Plan flood tank injection on small-Item I.C.1 break LOCAs.
- 36. Additional staff audit calcula-NUREG-0565 (2.4.2.a)
B&W lb/lc Refer to Action Plan tions of B&W small-break LOCA Item I.C.1 analyses.
37.
Analysis cf B&W plant response NUREG-0565 (2.6.2.c)
B&W id/lc Refer to Action Plan
'o-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 I.C.1 a small-break'LOCA in the pressurizer spray line.
TABLE IC (Continued)
".IML dequirement Category
{
Requirement So'urce Applicability Assignment Commentr 1
39.
Evaluation of effects of water NUREG-0565 (2.6.2.e)
B&W id/lc Refer to Action Plan slugs in piping caused by MPI Itec; I.C.1 and CFT flows.
40.
Evaluation of RCP seal damage NUREG-0565 (2.6.2.f)
B&W Id/lc Refer to Action Plan and leakage during a small-Item II.K.2.16 break LOCA.
j 41.
Submit preaictiuns for LOFT Test NUREG-0565 (2.6.2.g)
BaW id/lc l'efer to Action Plan L3-6 with ACPs running.
Item I.C.1
?
g 42.
Submit requested information NUREG-0565 (2.6.2.h)
B&L' id/lc Refer to Action Plan on the effects of non-Item I.C.1 condensible gases.
43.
Evaluation of mechanical effects NUREG-0565 (2.6.2.1)
B&W ld/lc Refer to Action Plan of slug flow on steam generator II.K.2.15 tubes.
J l
44.
Evaluation of anticipated NUREG-0626 (A.14)
GE le/lc Deleted generic transients with single failure study was submitted to verify no significant fuel failure.
l 45.
Evaluate depressurization with NUREG-0626 (A.15)
GE 3/lc Refer to Appendix B other than full ADS.
4 4
5 r
TABLE IC (Continued)
CP/ML Require.ient Category Requirement Source Applicability Assignment Comments 46.
Response to list of concerns NUREG-0626 (A.17)
GE Id/lc from ACRS consultant.
47.
Test program for small-break NUREG-0626 (8.9)
GE Id/lc Refer to Action Plan LOCA model verification pretest Items I.C.1, and prediction, test prograa and model II.E.2.2 verification.
48.
Assess change in safety NUREG-0626 (B.15)
GE Id/lc Refer to Action Plan reliability as result of Items II.C.1 and T
implementation B&OTF II.C.2 ti recommendations.
49.
Review of Procedures (NRC).
NUREG-0611 (3.4.1)
W. CE lb/lb Refer to Action Plan NUREG-0635 (3.4.1)
I.C.8 and I.C.9 50.
Review of Procedares NUREG-0611 (3.4.2)
W, CE Id/lc Refer to Action Plan (NSSS vendors)
NUREG-0635 (3.4.2)
I.C.7 and I.C.9 51.
Sympton-based emergency NURGE-0611 (3.4.3)
W. CE Id/lc Refer to Action Plan procedures.
NUREG-0626 (B.8)
GE Item I.C.9 NUREG-0635 (3.4.3)
- 52. Operator awareness of revised NUREG-0626 (A.11)
GE 1d/lc Refer to Action Plan emergency procedures.
Items I.B.1, I.C.2, and I.C.5
1 j
TABLE IC (Continued)
CP/ML Requirement Category Requirement Source Applicability Assignment Comments 53.
Two operators in control room.
NUREG-0626 (A.16)
JE Id/lc Refer to Action Plan Item I.A.1.3 54.
Simulator upgrade for small-NUREG-0565 (2.3.2.c)
All Id/lc Refer to Action Plan J
break LOCAs.
NUREG-0611 (3.3.1.b)
Item I.A.4.1 NUREG-0626 (8.11)
NUREG-0635 (3.3.1.b)
- 55. Operator monitoring of control NUREG-0611 (3.5.1)
W. CE 1d/lc Refer to Action Plan board.
NUREG-0635 (3.5.1)
Items I.C.1, I.D.2 and I.D.3 i
co 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 57.
Identify water sources NUREG-0626 (A.5)
GE Id/lc Refer to Action Plan prior to manual I.C.1 activation of ADS.
APPENDIX B 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 car y 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.
I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING, DESIGN AND CONSTRUCTION EXPERIENCE 1
1 Applicants shall submit a description of their administrative procedures for evaluating operating, design, and construction experience and describe how they will assure that applicable important industry experiences originating from both within and outside the applicant's construction organization will be provided in a timely manner to those designing and constructing the plant.
Applicants shall suomit a general discussion of how the requirments will be met.
These procedures shall:
(1) Clearly identify organization responsibilities for review and identification of these important experiences and the feedback of pertinent information to those responsible for designing and constructing the plant; (2) Identify the administrative and technical review steps necessary in implementing applicable important experiences; (3) Identify the recipients of various categories of irformation from these experiences or otherwise provide means through which such information can be readily related to the job functions of the recipients; (4) Assure that applicant and contractor personnel do not routinely receive extraneous and unimportant experience-related information in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency; (5) Provide suitable checks to assure that conflicting or contradictory information is not. conveyed to applicant and contractor personnel for implementation until resolution is reached; and (6) Provide practical interim audits to assure that the feedback program functions effectively at all levels.
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.C.9 LGNG-TERM PROGRAM PLAN FOR UPGRADING OF PROCEDURES Applicants shall describe their program plan, which is to begin during construc-tion and follow into operation, for integrating and expanding current efforts to improve plant procedures.
The scope of the program shall include emergency B-1
procedures, reliability analysis, human factors engineering, crisis management and operator training.
Applicants shall also insure that their program will be coordinated, to the extent possible, with INPO and other industry group efforts.
Applicants will submit, prior to the issuance of construction permits, 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.
I.D.1 CONTROL ROOM DESIGN REVIEWS Applicants shall provide preliminary design information at a level consistent with that normally required at the construction permit stage of review.
Applicants shall provide a general discussion of their approach to control room designs that reflect human f actor principles by specifying the design concept selected and the supporting design bases and criteria.
Cosmetic revisions to conventional (1960 technology) designs are unacceptable.
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.
Applicants shall commit to control room designs reflecting human factors principles prior to issuance of a CP or ML and shall supply design information for review prior to committing to fabrication or revision of fabricated control room panels and layouts.
I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE Applicants shall describe how they intend to meet the staff criteria contained in NUREG-0696 for a plant safety parameter display console.
The console shall display to operators a minimum set of parameters defining the safety status of the plant, capable of displaying a full range of important plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded.
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 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 i
properly prior to the issuance of operating licenses.
B-2
I. F.1 EXPAND QA LIST Prior to issuance of the construction permits or manufacturing license, applicants shall revise their QA programs by expanding their QA lists to include all items and activities affecting safety as defined by Regulatory Guide 1.29 and Appendix A to 10 CFR Part 50, and shall provide a commitment to apply the revised QA program to all such items and activities.
I.F.2 DEVELOP MORE DETAILED QA CRITERIA Applicants shall describe the changes to their QA programs that have resulted from their review of the accident at TMI-2.
In addition, applicants shall address the appropriate matters discussed in this Action Plan item, including the establishment of a quality assurance (QA) program based on consideration of:
(a) ensuring independence of the organization performing checking functions from the organization responsible for performing the functions; (b) performing quality assurance / quality control functions at construction sites to the maximum feasible extent; (c) including QA personnel in the documented review of and concurrence in quality related procedures associated with design, construction and installation; (d) establishing criteria for determining QA programmatic r.quirements; (e) establishing qualification requirements for QA and QC personnel; (f) sizing the QA staff commensurate with its duties and responsibilities; (g) establishing procedures for maintenance of "as-built" documentation; and (h) providing a QA role in design and analysis activities.
Applicants shall submit, prior to the issuance of the construction permit:, or manufacturing license, a revised description of their QA program that includes consideration of these matters.
II.B.1 REACTOR COOLANT SYSTEM VENTS Applicants shall modify their plant designs as necessary to provide the capability of high point venting of noncondensible gases from the reactor coolant system, and other systems that may be required to maintain adequate core cooling.
Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity.
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, aplicants shall provide a general discussion of their approach to meeting these 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.2 PLANT 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, as a result of an accident, contain TID 14844* source
^ TID 14844, U.S. Atomic Energy Commission, 1962.
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term radioactive material and (2) implement plant d2 sign modifications necessary to permit adequate access to important areas and to protect safety equipment from the radiation environment.
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.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 provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID 14844* source term radioactive materials without radiation exposures to any individual exceeding 5 rem to the whole-body or 75 rem to the extremities.
Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentra-tions.
Applicants 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 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.8 RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS Applicants shall:
(1) commit to perfu..... J a site / plant-specific probabilistic risk assessment and incorporating the results of the aosessment into the design of the facility.
The commitment must include a program plan, acceptable to the staff, that demonstrates how the risk assessment program will be scheduled so as to influence system designs as they are being developed, lhe assessment shall be completed and submitted to NRC within two years of issuance of the construction permit.
The outcome of this study and the NRC review of it will be a determination of specific preventive and mitigative actions to be implemented to reduce these risks.
A prevention feature that must be considered is an additional decay heat removal system whose functional requirements and criteria would be derived f om the PRA study.
It is the aim of the Commission through these assessments to seek such improvements in the reliability of core and containment heat removal
^ TID 14844, U.S. Atomic Energy Commission, 1962.
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systems as are significant and practical and do not impact excessively on the plant.
Applicants are encouraged to take steps that are in harmony with this aim.
4 i
(2) include provisions in the containment design for one or more dedicated penetrations, equivalent in size to a single 3-foot diameter opening.
This shall be done in order not to preclude the installation of systeins to prevent containment failure, such as filtered vented containment systems.
(3) provide a system for hydrogen control capable of handling hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction.
j.
(4) provide preliminary design information at a level consistent with that normally required at the construction permit stage of review sufficient to demonstrate that:
t (a) Containment integrity will be maintained (i.e., for steel containments by meeting the requirements of the ASME Boiler and Pressure Vessel Code, Division 1, Subsubarticle NE-3220, Service Level C Limits, e).:ept that evaluation of instability is not required, considering pressure and dead load alone.
For concrete containments by meeting the requirements of the ASME Boiler and Pressure Vessel Code, Division 2, Subsubarticle CC-3720, Factored Load Category, considering pressure and dead load alone) during an accident that releases hydrogen generated from 100% fuel clad metal-water reaction accompanied by either hydrogen burning or the added pressure from post-accident inerting assuming carbon dioxide is the inerting agent, depending upon which option is chosen for control of hydrogen.
As a minimum, the specific code requirements set forth above appropriate for each i
type of containment will be met for a combination of dead load and an internal pressuie of.45 psig.
Modest deviations from these criteria will be considered by the staff, if good cause is shown by an applicant.
Systems necessary to ensure containment integrity shall also be demonstrated to perform their function under these conditions.
(b) The containment and associated systems will provide reasonable assurance that uniformly-distributed hydrogen concentrations do not exceed 10% during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100% fuel clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.
(c) The facility design will provide reasonable assurance that, based on a 100% fuel clad metal-water reaction, combustible concentrations of
~
4 hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
(d) If the option chosen for hydrogen control is post-accident inerting:
(a) Containment structure loadings produced by an inadvertent full inerting (assuming carbon dioxide), but not including seismic or B-5
design basis accident loadings will not produce stresses in steel containments in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Division 1, Subsubarticle NE-3220, Service Level A Limits, except that evaluation of instability is not required (for concrete containments the loadings specified above will not produce strains in che containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Division 2, Subsubarticle CC-3720, Service Load Category), (b) demonstrate that a pressure test, which is required, of the containments at 1.10 and 1.15 times for steel and concrete containments, respectively) the pressure calculated to result from carbon dioxide inerting can be safely conducted, (c) demonstrate that inadvertent full inerting of the containment can be safely accommodated during plant operation.
(e) If the option chosen far hydrogen control is a distributed ignition system, equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integrity shall be designed to perform-its function during and after being exposed to the environmental conditions created by activation of the distributed ignition system.
II.D.1 TESTING REQUIREMENTS Applicants and their agents shall provide a test program and associated model development and conduct tests to qualify reactor coolant system relief and safety valves and, for PWR's, PORV block valves, for all fluid conditions expected under operating conditions, transients and accidents.
Consideration of anticipetcd transient without scram (ATWS) conditions shall be included in the test program.
Actual testing under ATWS conditions need not be carried out until subsequent phases of the test program are developed and not before l
issuance of an ATWS rule.
Applicants shall submit, prior to the issuance of the construction permits or manufacturing license, a general explanation or how the testing 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.
Applicants shall (1) demonstrate the applicability of the generic tests conducted
(
under II.D.1 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 these requirements and shall l
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 test results will be incorporated in the plant design.
Sufficient I
detail should be presented to provide reasonable assurance that the requirer:ents resulting from the test will be implemented properly prior to the issuance of operating licenses.
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.. Applicants shall, to the extent possible, provide preliminary design information at a B-6 w
i 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 th "equirements by specifying the j
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 issuance of operating licenses.
II.E.1.1 AUXILIARY FEEDWATER SYSTEM EVALUATION t
Applicants shall perform a reevaluation of their proposed auxiliary feedwater (AFW) system.
This reevaluation shall ir.clude the following:
(1) Performance of 31rplified auxiliary feedwater system reliability analyses using event-tree and fault-tree logic techniques to determine the potential for AFW system failure under various loss of main feedwater transient conditions, with particular emphasis being given to determining potential failures that could result from human 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.
i Applicants with plants with AFW systems with relatively low reliabilities l
shall submit proposed design changes and/or procedural actions which will improve the relative reliability of the AFW system to above average.
Applinants 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 i
relative to those AFW systems evaluated by the NRC and staff and reported in NUREG-0611 and NUREG-0635 respectively.
(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 flow 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 studiec are factored into the final designs.
II.E.1.2. AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION Applicants with PWR plants shall provide automatic and manual aexiliary feedwater (AFW) system initiation and auxiliary feedwater system flow indication in the control room.
These systems shall be safety grade and meet the requirements specified in NUREG-0737. 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 j
there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
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T II.E.3.1 RELIABILITY OF POWER SUPPLIES FOR NATURAL CIRCULATION Applicants shall (1) upgrade the power suppl,ies for the pressurizer heaters i
and associated motive and control power interfaces to meet the applicable requirements specified in NUREG-0737 and (2) establish procedures and training for maintaining the reactor coolant system at hot standby conditions with only onsite power available.
Applicants shall, to the extent possible, provide preliminary design information at a level consistent witn 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 spacifying the design concept selected and the supporting design bases and criteria.
Applicants shall also demoi trate that the design concept is technically feasible and within the state-of-the-art, and that there exists reasonable I
assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
j II.E.4.1 DEDICATED PENETRATION
~
Applicants for plant designs with e>.ternal hydrogen recombiners shall modify their applications as necessary to include redundant dedicated containment pentrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere.
Applicants shall submit, prior to the issuance of construction permits or the manufacturing license, a detailed explanation of how the requirements will be met in order to provide reasonable assurance that the requirements will be implemented properly.
i II.E.4.2 ISOLATION DEPENDABILITY 4
Containment isolation system drasigns shall comply with the recommendations of Standard Review Plan Section 6.2.4.
All plants shall give careful consideration to the definition of essential and 4
j non-essential systems, identify each system determined to be essential, identify each system determined to be non-essential, and describe the bai.!s for selection l
of each essential system.
All non-essential systems shall be automatically isolated by the containment isolation signal.
Revision 2 to Regulatory Guide 1.141 will contain guidance on the classification of essential versus non-essential systems and is due to be issued by June 1981.
For post-accident situations, each non-essential penetration (except instrument lines) is rq! ired to have two isolation barriers in series that meet the requirements of General Design Criteria 54, 55, 56, and 57, as clarified by Standard Review Plan, Section 6.2.4.
Isolation must be performed automatically (i.e., no credit can be given for operator action).
Manual valves must be sealed closed, as defined by Standard Eeview Plan, Section 6.2.4, to qualify as an isolation barrier.
Each automatic isolation valve in a non-essential penetration must recei/e diverse isolation signals.
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The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves.
Reopening of containment isolation valves shall require deliberate operator action.
Administrative provisions to close all isolation valves manually before resetting the isolation signals is 2
not an acceptable method of meeting this requirement.
Ganged reopening of containment isolation valves is not acceptable.
Reopening of isolation valves must be performed on a valve-by-valve basis, or on a line-by-line basis, provided that electrical independence and other single-failure criteria continue to be satisfied.
The containment setpoint pressure that initiates containment isolation for non essential penetration; must be reduced to the minimum compatible with normal operating conditions.
The containment pressure history during normal operation for similar operating plants should be used as a basis for arriving at an appropriate minimum pressure setpoint for initiating containment isolation.
The pressure setpoint selected should be far enough above the maximum observed (or expected) pressure inside containment during normal operation so that inadvertent containment isolation does not occur during normel operation from instrument drift or fluctuations due to the accuracy of the pressure sensor.
A margin of 1 psi above the maximum expected containment pressure should be adequate to account for instrument error.
Any proposed values greater than 1 psi will require detailed justification.
All systems that provide a path from the containment to the environs (e.g.,
containment purge and vent systems) must close on a safety grade high radiation i
signal.
Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979, must be sealed closed as defined in SRP 6.2.4, item II.3f during operational conditions 1, 2, 3, and 4.
Furthermore, these valves must be verified to be closed at least every 31 days.
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.
4 II.E.4.4 PURGING Applicants shall (1) provide a capability for containment purging / venting designed to minimize purging time, consistent with ALARA principles for occupa-tional exposure, (2) evaluate the performance of purging and venting isolation valves against accident pressure, (3) address the interim NRC guidance on valve operability, (4) adopt procedures and restrictions consistent with the revised requirements; and (5) provide and demonstrate high assurance that the i
purge system will be reliably isolated under accident conditions.
i B-9 we
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.5.1 DESIGN EVALUATION Applicants with B&W-designed reactors shall (1) identify the most severe overcooling events (considering both ar.ticipated 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 reactor 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 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 reasonalle assurance that the require-ments will be implemented properly prior to the issuance of operating licenses.
II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION Applicants shall provide instrumentation to measure, record and readout in the control room:
(a) containment pressure, (b) containment water level, (c) contain-ment hydrogen concentration, (d) containment radiation intensity (high level),
and (e) noble gas effluents at all potential, accident release points.
The requirements for the specific monitors are listed in NUREG-0737.
Applicants shall also provide for continuous sampling of radioactive iodines and particu-lates in gaseous effluents from all potential, accident release points, and for onsite capability to analyze and measure these samples.
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 irvolved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the i
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.
1 II.F.2 IDENTIFICATION OF AND REC 0VERY FROM CONDITIONS LEADING TO INADEQUATE CORE COOLING Applicants shall describe their program for developing and implementing procedures to be used by the reactor operators to detect and recover from conditions leading to inadequate core cooling.
i B-10 1
i 1
Applicants shall provide instruments that provide in the control room an unambiguous indication of inadequate core cooling, such as primary coolant saturation meters in PWR's, and a suitable combination of signals from indicators of coolant level in the reactor vessel and incore thermocouples in i
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 2
the design concept selected and the supporting design bases and crite~ia.
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.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 defined design bases and Regulatory Guide 1.97, Rev. 2, December 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 demnstrate that the design concept is technically i
feasible and within the state-of-the-art, and that there exists reasonable asturance 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 provide power supplies for the pressurizer relief valves, block valves, and pressurizer level indicators to meet the applicable requirements specified in NUREG-0737.
Level indicators shall be powered from vital buses, motive and control power connections to emergency power sources shall be through devices qualified in accordance with requirements applicable to systems important to safety, and electric power shall be provided from emergency sources.
Applicants with PWR plants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the coastruction 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 support 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.
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i II.J.3.1 ORGANIZATION AND STAFFING TO OVERSEE DESIGN AND CONSTRUCTION Applicants shall describe their prog am for the management oversight of design and construction activities.
Specific items to be addressed include:
(1) the organizational and management structure which is singularly responsible for the direction of the design and construction of the proposed plant, (2) technical resources which are directed by the utility organization, (3) details of the interaction of design and construction within the utility organization and the manner by which the utility will assure close integration of the architect engineer and nuclear steam supply vendor, (4) proposed procedures for handling the transition to operation, and (5) the degree of top level management over-sight and technical control to be exercised by the utility during design and construction, including the preparation and implementation of procedures necessary to guide the effort.
Draft NUREG-0731, " Guidelines for Utility Management Structure and Technical Resources" is the keystone for similar development of guidelines for this task.
Therefore, the principal applicable elements of NUREG-0731 shall be used by CP and ML applicants in addressing this task.
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.I.22 DESCRIBE AUTOMATIC AND MANUAL ACTIONS FOR PROPER FUNCTIONING 0F AUXILIARY HEAT REMOVAL SYSTEMS WHEN FW SYSTEM NOT OPERABLE Applicants with BWR plants shall design auxiliary heat removal systems such that necessary automatic and manual actions can be taken to ensare proper functioning when the main feedwater system is not operable.
A general explana-tion of how this requirement will be met is required prior to issuance of the construction permits.
Sufficient detail shall be presented to provide reason -
able assurance that the requirements wili be implemented properly.
l 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 issued to operating B&W plants in May 1979 regarding the analysis and upgrading of the integrated control system (ICS), and perform a failure modes and effects analysis of the ICS to include considerations of failures and effects of input and output signals to the ICS.
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 discussic of their approach to meeting the requirements by specifying the design cc apt selected and the supporting design bases and criteria.
Applicants l
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 provide, as part of the reactor protection system, an anticipatory reactor trip that would be actuated on loss B-12
[
of main feedwater and on turbine trip.
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 s' ate of the art, a d that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
II.K.2.16 IMPACT OF RCP SEAL DAMAGE FOLLOWING SMALL-BREAK LOCA WITH LOSS OF 0FFSITE POWER Applicants shall perform an evaluation of the potential for and impact of reactor coolant pump seal damage following small-break LOCA with loss of offsite power.
If damage cannot be precluded, provide an analysis of the limiting small-break loss-of-coolant accident with subsequent reactor coolant pump seal damage.
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.3.2 REPORT ON OVERALL SAFETY EFFECT OF PORV ISOLATION SYSTEM Applicants with PWR plants shall address the requirements set forth in Item 3.2.4.e and 3.2.4.f of NUREG-0611 and perform an analysis of the prob-ability of a small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve (PORV).
If this probability is a significant contributor to the probability of small-break LOCA's from all causes, provide a descripti;n and evaluation of the effect on small break LOCA probability of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls after the PORV has opened.
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.3.13 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 as they apply to HPCI and RCIC systems, and perform an evaluation of the safety effectiveness of providing for separation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) system initiation levels so that the RCIC system initiates at a higher water level than the HPCI system, and of providing that both systems restart on low water level.
Applicants shall provide sufficient information to describe the nature of the studies, ho,. 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.
- For plants with high pressure core 4 ay systems in lieu of high pressure coolant injection systems, substi.ute tne words, "high pressure core spray" for "high pressure coolant irNtion" and "HPCS" for "HPCI".
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l II.K.3.16 REDUCTION OF CHALLENGES AND FAILURES OF RELIEF VALVES - FEASIBILITY STUDY AND SYSTEM MODIFICATION Applicants with BWR plants shall address the requirements set forth in Item A.4 of NIJREG-0626, and perform a study to identify practicable system modifications that would reduce challenges and failures of relief valves, without compromising the performance of the valves or other systems.
Applicants shall provide suf-ficient 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.3.18 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 and perform a feasibility and risk assessment study to determine the optimum automatic depressurization system (ADS) design modifications that would eliminate the need for manual activation to ensure adequate core cooling.
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 MODIFICATION Applicants with BWR plants shall address the requirements set forth in Item A.10 of NUREG-0626 and perform a study of the effect on all core-cooling modes under accident conditions of designing the core spray and low pressure coolant injection systems to ensure that the systems will automatically restart on loss of water level, after having been manually stopped, if an initiation signal is still present.
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 CENTRAL WAFER LEVEL RECORDING Applicants with BWR plants shall provide the capability to record reactor vessel water level in one location on recorders that meet normal post-accident recording requirements.
Applicants shall implement design modifications as necessary to meet the requirements.
Applicants shall submit, prior to issuance of constructi<n 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.
II.K.3.24 CONFIRM ADEQUACY OF SPACE COOLING FOR HPCI* AND RCIC SYSTEMS Applicants with BWR plants shall address the HPCI and RCIC systems requirements set forth in Item B.3 of NUREG-0626, and perform a study to determine the need
- For plants with high pressure core spray systems in lieu of high pressure coolant injection systems, substitute the words, "high pressure core spray" for "high pressure coolant injection" and "HPCS" for "HPCI".
B-14
for additional space cooling to ensure reliable long-term operation of these systems following a complete loss of offsite power to the plant for at least two hours.
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 VERIFY QUALIFICATION OF ACCUMULATORS ON ADS VALVES Applicants with BWR plants shall provide information to ensure that the ADS valves, accumulators, and associated equipment and instrumentation will be capable of performing their intended functions during and following an accident situation while taking no credit for non-safety related equipment or instru-mentation.
Air (or nitrogen) leakage through valves must be accounted for to ensure that enough inventory of compressed air (or nitrogen) will be available to cycle the ADS valves.
Applicants shall commit that these requirements will be met in the final design at the OL stage.
In addressing this item prior to CP issuance, applicants should note that sa"ety analysis reports claim that air (or nitrogen) accumulators for the ADS va res provide sufficient capacity (inventory) to cycle these valves open five i
times at design pressures.
Also, General Electric has stated that the emergency core cooling systems are designed to withstand a hostile environment and still perform their functions for 100 days following an accident.
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, and provide an evaluation of depressurization methods, other than by full actuation of the automatic depressurization sytem, that would reduce the possibility of exceeding vessel integrity limits during rapid cooldown.
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.
III.A.1.2 UPGRADE LICENSEE EMERGENCY SUPPORT FACILITIES Applicants shall address the requirements for a lechnical Support Center, Operational Support Center and the Emergency Operations Facility.
Applicants shall provide preliminary design information in accordance with the functional criteria in NUREG-0696 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 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.1.1 PRIMARY COOLANT SOURCES OUTSIDE THE CONTAIN"ENT STRUCTURE Applicants shall review the designs of such systems outside containment, and their provisions for leakage control and detection, overpressurization design, B-15
discharge points for waste gas venting systems, etc., with the goal of minimizing potential exposures to workers and public following an accident, and providing reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency.
Applicants shall provide for leakage control and detection in the design of systems outside containment that contain (or might-containing) TID 14844* source term radioactive materials following an accident, and submit a leakage control program, including an initial test program and a schedule for retesting these systems, and the actions to be taken for minimizing leakage from such systems.
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 satisfactorily prior to issuance of operating licenses.
III.D.3.3 IN-PLANT RADIATION MONITORING Applicants shall review their designs to ensure that provisions for monitoring inplant radiation and airborne radioactivity are appropriate for a broad range of routine and accident conditions.
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.
III.D.3.4 CONTROL ROOM HABITABILITY Applicants shall review the design of their facilities for conformance to requirements stated in the Action Plan.
Applicants shall evaluate potential pathways for radioactivity and radiation that may lead to control room habita-bility problems under accident conditions resulting in a TID 14844* source term release and make necessary design provisions to preclude such problems.
Applicants shall address prior to the issuance of the construction permits or manufacturing license, how they will implement the existing requirements set forth in this Action Plan item.
Applu nts shall also address the extent to which improvements have been made to present 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 yequired 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.oncept 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.
- TID 14844, U.S. Atomic Energy Commission, 1962.
B-16
O.S. CUCLE AD RE!UL ATORY COMMISSION BIBLIOGRAPHIC DATA SHEET p
4 TITLE AND SUBTITLE (Aad Veeme Na, #1earersamt
- 2. (Leme 6/m Al Licensing Requirements for Pending Applications for J. RECIPIENT S ACCESSION NO.
Construction Permits and Manufacturing License
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLETED M ON TH l YEAR June 1981
- 9. PERFORMING ORGANIZATION N AMt AND MAILING ADDRESS tlaciude Isa Codri DATE REPORT ISSUED MONTH lVEAR U.S. Nuclear Regulatory Ccmmission June 1981 Office of Nuclear Reactor Regulation
- s. (t,,v, y,,,
Washington, DC 20555 B. (Leave Nmkl
- 2. SPONSORING ORGANIZATION NAME AND MAILING ADORESS (tac /udr Isa Codr/
n.CONrRACrNO.
Same as 9 above 13 TYPE OF REPORT PE RIOD COVE RED (/nclusive dams)
- 15. SUPPLEMENTARY NOTES 14 (Le Jve c/m Al
- 16. ABSTR ACT 000 words or less)
The TMI-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 units with light water reactors and one manufacturing license application for eight floating nuclear plants.
Staff review of these applications had been suspended si.1ce the TMI-2 accident pending the formulation of a policy to appropriately reflect the lessons learned from the accident.
The Commission is considering a new rule which will state the TMI-related requirements to be applied to these applications. NUREG-0718 was issued, and has now been revised, to provide guidance that the NRC staff believes should be followed to account for the lessons learned from the TMI-2 accident. NUREG-0718 is not a substitute for the regulations, and compliance is not a requirement. However, an approach or method different from the guidance contained herein will be accepted only if the substitute approach or method provides an equivalent basis for meeting tha requirements.
17 KEY WORDS AND DOCUMENT ANALYhlS 17a DESCRIPTORS 17b. IDE NTIFIE RS/OPEN EN DE D TE RMS 18 AV AILABILITY STATEMENT 19 SECURITY CLASS (This reportl 21 NO OF P AGE S Unclassified 22 PRICE Unlimited 20gECUR YgASgThes pagel N00C FORM 33517 77)
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