ML20028B521

Requests Resources & Review Be Made Available for Licensing Issues to Be Considered in EPRI LWR Standardization Program. Response by 821119 Requested

ML20028B521
Person / Time
Issue date:	11/15/1982
From:	Harold Denton Office of Nuclear Reactor Regulation
To:	Hanauer S, Mattson R, Vollmer R Office of Nuclear Reactor Regulation
References
PROJECT-669A NUDOCS 8212020204
Download: ML20028B521 (24)

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NOV 151982 MEMORANDUM FOR: Roger J. Mattson, Director Division of Systems Integration Steven H. Hanauer, Director Division of Safety Technology Richard H. Vollmer, Director Division of Engineering Hugh L. Thompson, Director Division of Human Factors Safety FR0fi:

Harold R. Denton, Director Office of Nuclear Reactor Regulation

SUBJECT:

LICENSING ISSUES TO BE CONSIDERED IN EPRI's LWR STANDARDIZATION PROGRAM On July 20, 1982, I met with representatives of the Electric Power Research Institute's (EPRI's) Utility Steering Committee for the Industry Program for the Development of Standardized Light Water Reactor Designs for One-Step Licensing to discuss EPRI's standardization program. At that meeting, the EPRI representatives described their program and requested that NRC provide the necessary resources to complement the industry's commitment to the initial phase of the program.

n response to EPRI's request, I agreed to make available resources within NRR as necessary to work with EPRI on its program.

Specifically, I agreed to be the lead individual for matters of NRC policy and resource commitments.

In addition, I have designated Frank Miraglia, Assistant Director for Safety Assessment, and Cecil Thomas, Chief, Standardization & Special Projects Branch, as the individuals to help define the scope, outstanding issues and schedules for the reviet of the standard reactor designs.

The Standardization and Special Projects Branch, Division of Licensing has

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the lead responsibility for managing the staff's effort associated with EPRI's program. A Project Manager for that function will be designated d[

soon.

The other NRR Divisions are requested to make available resources as necessary to work on this program.

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2 NOV 151982 Division Directors On October 21, 1982, representatives of the Divisicn of Licensing met with EPRI representatives to discuss further its program. A sunmary of that meeting is provided herein as Enclosure 1.

At that meeting, the EPRI repre-sentatives gave us a copy of a September 29, 1982 letter from R. E. Engel of S. Levy, Inc. to Lou Martel of EPRI (see Enclosure 2).

In that letter, S. Levy, Inc., consultant to EPRI for its program, lists and prioritizes approximately 510 licensing issues which it believes need to be considered ire the program.

In response to ERPI's request, we agreed to review that list for completeness and appropriateness of prioritization.

Accordingly, I request that you review EPRI's list to ensure that it contains all of the licensing issues within your supe of responsibility that you believe should be considered by EPRI in its program.

In addition, I request that you review the appropriater.ess of EPRI's prioriti7ation of those issues within your scope of responsibility. Please let ne have the results of your review by November 19, 1982 as we plan to meet with EPRI representatives shortly thereafter to discuss the results of our review.

Gridwi%& try H. R. Der,tm Harold R. Denton, Director Office of Nuclear Reactor Regulation

Enclosures:

As stated cc:

D. Eisenhut DISTRIBUTION:

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SSPB Reading D. Eis'enhut F/. iiraglia Romas HEMORANDUM FOR: Roger J. Mattson, Director Division of Systens Integration Steven H. Hanauer, Director Division of Safety Technology Richard H. Vollmer, Director Division of Engineering tiugh L. Thonpson, Director / afety Division of Human Factors S

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FROM:

Darrell G. Eisenhut, Director Division of Licensing'

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SUBJECT:

LICENSING ISSUES TO BE CONSIDERED IN EPRI's LUR STANDARDIZATION PROGRAM

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On October 21, 1982, we met with representatives of the Electric Power Research Institute (EPRI) to discuss,its LWR standardization program.

A summary of that meeting is provided herein as Enclosure 1.

At that meeting, the EPRI repre-sentatives gave us a copy of a September 29, 1982 letter from R. E. Engel of j

S. Levy, Inc. to Lou Martel of EPRI (see Enclosure 2).

In that

+t ar, S. Levy,

Inc., consultant to EPR,I f or its program, lists and prioritiz

'nately 510 licensing issues wnich it believes need to be considered 6

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At the EPRI representstives' request, we agreed to review that corpleteness and app'ropriateness of prioritization.

Accordingly, I req /uest that you review EPRI's list to ensure that it contains all of the licensing is. sues within your scope of responsibility that you believe should be considered by EPRI in its program.

In addition, I request that you review the appropriateness of EPRI's prioritization of those issues within your scope of responsibility. Please let me have the results of your review by November 19, 1982 as we plan to meet with EPRI representatives shortly thereafter to discuss /the results of our review.

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Darrell G. Eisenhut, Director

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Division of Licensing l

I / concur in this effort J.

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ENCLOSURE 1

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D. C. 20555

(-e OCT 2 71992 MEMORANDUM FOR:

Darrell G. Eisenhut, Director Division of Licensing THRU:

, Frank J. Miraglia, Assistant Director for Safety Assessment Division of Licensing FROM:

Cecil 0. Thomas, Acting Chi.ef Standardization & Special Projects Branch Division of Licensing

SUMMARY

OF OCTOBER 21, 1982 MEETING WITH EPRI ON ITS LWR STANDARDIZATION PROGRAM V

On October 21, 1982, Gus Lainas, Frank Miraglia and I met in Bethesda with Vsarl Stahlkopf, Robert Nickell and Jack Berga of the Electric Power Research' Institute (EPRI) to discuss its LWR standardization program.

Among.the '

subjects discussed were the objectives and general features of EPRI's program, the NRC resources EPRI believes are needed to proceed with the first phase of igs program and EPRI's plans for future activities.

The objectives of EPRI's program are to develop up to three light water reactor plant " baseline designs" and the " approach" which could be used by vendors in obtaining NRC certification of plant-designs based on these baselinc designs.

Baseline designs would be developed for at least bne BUR plant and.one PWR plant.

The " approach" would include specification of the licensing issues which must be resolved and the level of design detail which must be provided to obtain NRC certification.

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EPRI's program consists of several phases.

Phase I, which is expected to take approximately one year to complete, involves the development of a list of " utility betterments," i.e., design features which are expected to improve plant reliability and availability; the determination of the licensing issues which must be resolved and the level of design detail which must be provided to obtain NRC certification; and the establishment of detailed procedures for resolving the licensing issues. The procedures for resolving the licensing issues would be established through interaction with NRC on three or four representative issues such as possible revision to Appendix K to 10 CFR 50, decay heat removal, inservice inspection and systems interactions.

Following the completion of. Phase I of the program, EPRI would issue Requests for Proposals from nuclear steam supply system vendors and architect-engineering firms for plant designs to be submitted by the vendors / firms for NRC certifi-cation.

These designs would be based on the specifications developed during Phase I of EPRI's program.

During NRC's review of these. designs, EPRI's role would be limited to that of a facilitator.

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OCT 2 71982 Darrell G. Eisenhut '

The EPRI representatives referred to their September 15, 1982 letter to Mr. Denton (see Enclosure 1) for a description of the NRC resources it believes are needed to proceed with the first phase of its program. As indicated in their letter, EPRI believes a Policy Comittee and a Program Group are necessary in order for NRC to complement the industry's commitment to the initial phase of its program. The EPRI representatives stated their expectation that the Policy Comittee would consist of one Comissioner, one ACRS Member, and Messrs. Dircks, Stello and Denton. They did not elaborate on their expectations for the Program Group.

We pointed out that we had responded to EPRI's letter by a let'er dated October 20, 1982 from Mr. Denton (see Enclosure 2).

In our letter, wa stated that Mr. Denton will be the lead individual for matters of NRC policy-and resource comitments, and Messrs. Miraglia and Thomas will be the individuals to help define the scope, outstanding issues and schedules for reviewing the standard plant design.

At the request of the EPRI represen-tatives, we agreed to make sure that Mr. Denton is aware of the discrepancy,

between the expected and designated compositions of the Policy Comittee.

Accordingly, we suggest that you discuss this matter with Mr. Denton at your earliest convenience.

The (PRI representatives gave us a copy of a September 29, 1982 letter from R. E.Yngel of S. Levy, Inc. to Lou Martel of EPRI.

In that letter, S. Levy, Inc., consultant to EPRI for its program, lists and prioritizes approximately 510 licensing issues which it believes need to be considered in the program.

The EPRI representatives requested that we review that list for completeness and appropriateness of prioritization.

Accordingly' we have prepared for your signature and Mr. Denton's concurrence a memorandum to the other NRR Division Directors requesting their assistance in this effort.

w, Finally, the EPRI representatives requested that we meet again in 5-6 weeks to consider the corgleteness and appropriateness of prioritization of the licensing issues and to discuss with the cognizant task managers the three or four representative licensing actions to be considered in Phase I of the p rogram.

ce.eo. h Cecil 0. Thomas, Acting Chief Standardization & Special Projects Branch Division of Licensing

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Committee:

EPRI Utility Steering Address Writer at:

Committee - Standardized Light Water Reactor Design Wisconsin Electric Power Company 231 Wesc Michigan Street Milwaukee, Wisconsin 53201 Telephone: (414) 277-2121 September 15, 1982 Mr. Harold Denton, Director

' Office of Nuclear Reactor R*egulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C.

20555

Dear Mr. Denton:

On July 20, 1982, you met with members of the EPRI.

Utility Steering Committee for the Industry Program for Development of Standardized Light Water Reactor Design for One-Step Licensing.

At that time we outlined Phase I of :our proposed program,- and disassed possible interaction between NRC and the indu.stry program.

You expressed a willingness to assign personnel.to work with EPRI on the program.

Proceeding to Phase II of the program will depend, in some part, upon our ability to work constructively with NRC during Phase I.

The objectives of this interaction were to:

Help cefine scope, degree of detail, and starting point-for standardizing design, as well as site-enveloping characteristics.

Identify outstanding issues to be resolved and schedule for resolution.

C) I Develop detailed approach to resolve three typical outstanding issues,. including, mechanisms', plans, information and appropriate schedule.

In crder to achie've these objectives, it was requested that the NRC provide the necessary resources to complement the industry, commitment to the initial phase of'the program.

It is expected that two. levels of NRC participation are required initially:

a Policy Commj.ttee and

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a Program Group.

The Policy Committee would provide interpretation of NRC policy regarding. standardization and give direction and provide resources for the program.

They would also resolve any differing

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views that develop at the. program level and provide the focal point

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Mr. H.arold Danton September 15, 1982 for obtaining necessary decisions to enable the program to proceed.

As suggested by Chairman Palladino, legal and procedural, as well as techn.ical, input will be needed.

I have arranged to enlist the assistance of Barton 3. Cowan, Esq., Chairman of the Legal Committee of the Atomic Industrial Forum, to work with the Utility Steering Committee and interact with the NRC Policy Committee on such matters.

The Program Group is necessary to demonstrate the feasibility of the program.

This group would work with industry.

representatives and be responsible for identifying.all current licensing issues of a technical nature whose resolution is necessary for certification of standard light water reactor designs.

In addition, three such issues would be selected by the Utility Steering Committee and the NRC Policy Committee to demonstrate the feasibility of reasonably resolving issues independent of -

specific licensing activities.

The Program Group would work with' seJ ected industry representatives with the goal of developing a' plan and schedule for resolution of these sample issues.

Karl Stahlkopf plans to visit Daryl Eisenhut as soon as possible to establish the specifigs.of near-term activities of the Program Group and EPRI.

As.y.ou know, there is considerable divergence od opinion as to the approaches that should be taken both by the industry and by the NRC to further nuclear plant standardization.

Some of theseedifferences may affect the scope and schedule of this program.

It is believed, however, that the program proposed here by EPRI is consistent with the current views of safety goals, backfitting provisions, one-step licensing,.and other' programs intended to remove present regulatory uncertainties, stabilize. plant designs, and attain the benefits of standardization.

NRC participation in this proposed undertaking is necessary if this activity is to have any potential opportunity.

s s, Very truly yours, O

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Sol urstein, Chairman Utility Steering Committee f ' Copy to Mr. J. Tourtellotte c

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ENCLOSURE 2 PRC*

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DEisenhut/RPurple OCT 2 01982 FMiraglia/DNottingham t r

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EJordan, IE JTaylor, IE SECY (3)

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!!r. Sol Burstein, Chairman WDircks RMattson EPRI Utility Steering Committee JR e RVollmer Standardized Light Water Reactor Design TRehm HThompson (1232 Wisconsin Electric Power Company

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.!!ilwaukee, Wisconsin 53201 HDenton/ECase

Dear Mr. Durstein:

PPAS-This is in response to your letter dated September 15, 1982 concerning the Electric Power Research Institute's (EPRI's) standardized light water reactor design program.

In that letter you requested that we provide resources to complement the industry's commitment to the initial phase of your program.

I believe that early HRC involvement in a progran such as you have outlined will prove beneficial to both of our organizations and will enhance the role of' standardization in nuclear power plant design. Accordingly, I will be the lead.

indivgual for natters of HRC policy. and resource commitments.

In addition, I have de'signated fir. Frank J. liiraglia, Assistant"Jirector for Safety Assessment and !:r. Cecil 0. Thomas, Acting Chief, Standardizrtion & Special Projects l

Brcnch, as the individuals to help define the scope, outstanding issues, and schedules for a review of standard reactor designst l

Ue can meet with you and your representatives.as soon as practicable. The objective of that reeting should be,'to provide us with a better understanding of the details of your program.

That understanding will enable u.t to more l

' definitively plan to meet your needs.

To arrange that meeting, please contact Cecil Thomas on (301)492-7130.

Sincerely, Of$d i&W.t l

K R. MF.03 Harold R. Denton, Director Office of I'uclear Reactor Regulati g(,0 W y OV e0 h>,

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ENCLOSURE 2 e

S. LS/Y, INC.

Suite 725 1999 South Boscom Avenue Compbell, Californio 95008-2233 USA 408/377 4870 September 29, 1982 Mr. Lou Martel Electric Power Research Institute 3412 Hillview Ave.

P.O. Box 10412 Palo Alto, CA 94303

Subject:

flRC Generic Licensing Issues - Update Refe'rence:

Letter, R.

E.

Engel to Lou Martel, "NRC Generic Licensing Issues," September 7,1982 e

Dear Lou:

The purpose of this letter is to provide you with the current status of the' work performed to date to prioritize the generic NRC licensing j

issues.

This work issues.

expands upon the reference letter to include the TMI To_ date, a - total- -of approximately-- 510 licensing issues have been

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identified.

A descript. ion and currently available f;RC status on each 1

issue have been previously provided to you.

have reviewed each issue and provided a ranking for it.In

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,I the issues is provided in Tables 1 to 6.

The ranking of The guidelines that I used to establish the priority of individual issues are described below.

1.

Critical (Table 1) - A critical issue is one that could result in a major plant reaesign to resolve.

These issues are generally complex and not well defined nor are potential solutions apparent.

required on these issues.It is likely that substantial work will be There are currently 6 issues in this category.

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Mr. Lou Martel September 29, 1982 Page 2 2.

High (Table 2) - A high priority issue is one that will probably require resolution prior to' design certification.

Many of these issues could result in significant design changes; however, they are reasonably well defined.

Included in this category are:

Unresolved Safety Issues, NRR High Priority Issues and TMI Issues required by 10 CRF 50.34 (F).

There are currently 50 issues in this category.

3.

Medium (Table 3) - A medium priority issue is one that will probably have to be addressed prior to design certification.

Some of these issues could result in significant design changes; however, it is expected t' hat most can be accomodated by little or no change to current designs.

Included in this category are Active or On-Going Safety or Licensing Improvement Issues, f.'RR Medium Priority Issues and Active TMI Issues or Requirements.

There are currently 36 issues in this category.

4.

Low (Table 4)

A low priority issue is one that will most likely have no impact on design certification.

This type of

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issue is one that will most likely have no impact on design certification.

This type of issue is generally. kept for historical purposes.

Included in this category are Inactive Safety or Licensing Improvement Is sues..

There are curr. fjRR Low Priority Issues and Inactve TMI Issues.

ently 66 issues in this category.

5.

Resolved (Table 5)

' Resolved issues t.ra issues conside' red technically resolved.

Included in this category are TMI issues that have been completed. There are currently 89. issues in -this-category.

6.

tiot Applicable (Table 6) - An issue in this category is one that is judged to be not applicable to design and certification.

Included in this category are Environmental, Superceded, Terminated or Subsumed Issues.

Also included are issues applicable only

• to previous designs, operator training or procedures, ftRC activities B8W

designs, and analysis There are curren,tly 263 issues in this category. -

requirements.

Because of the large number of issues di rected at specific potential problem areas, I have placed similar high, medium and resolved issues into possible significant groups.

Due to the number of issues in these groups, resolution of the genera 1 problem area may be considered critical to design certification.

-These ' groupings are provided in ' Table 7.

It should be noted that other groupings of issues are possible based strictly on the judgment of the person preparing the list.

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Mr. Lou Martel September 29, 1982 Page 3 Based on the NRC discussions I had on my meeting of September17, 1982, I expect that one more update will be required this year.

The basis for the update will be the issuance of a new Generic Issue Tracking System run and the new NRR Prioritization.

I should receive copies of this information and complete the update by November 1, 1982.

I understand that an additional 30 issues have been identified.

No further input on TMI issues is expected this year.

If you need any further information on this subject, please contact me.

Yours truly, bd R. E.'EngEi SLI A46achments REE/jt

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7 TABLE 1 9

CRITICAL ISSUES FOR DESIGN CERTIFICATION

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A-9 Anticipated Transients Without Scram A-17 Systems Interaction A-29 Nuclear Power Plant Design for the Reduction of Vulnerability to Industrial Sabotage 1

A-44 Station Blackout A-4 5-Shutdown Decay' Heat removal Requirements II.B.8 Safety Review Consideration - Rulemaking Proceeding on Degraded Core Accidents Dar -

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TABLE 2 HIGH PRIORITY IS'UES APPLICABLE TO DESIGN S

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1. 15 Radiation Effects on Reactor Vessel Supports

1. 18 Steam Line Break with Consequential Small LOCA

1. 23 Reactor Coolant Pump Seal Failures A-1 Water Hammer A-3 Westinghouse Steam Generator Tube Integrity A-4 CE Steam Generator Tube Integrity A-12 Fracture Tougnness of Steam Generator and Reactor Coolant Pump Suppcrts A-39 Determination of Safety Relief Valve Pool Dynamic

. Loads AC40 Seismic Design - Short-Term Program A-43 Containment Emergency Pump Performance A-47 Safety Implication of Contr 1 Systems A-48 Burns on Safety EquipmentHydrogen Control Measu i

A-49 Presurized Thermal Shock B-10 Behavior of Mark III Containments I

I I.D. 1(1)-(5)

Control Room Design Reviews I.D. 2 Control Room Design - Plant Safety Parameter Display Console I.F. 1 Quality Assurance - Expand QA List II.B.1 Safety Review Consideration - Reactor Coolant System Vents S

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TABLE 2 HIGH PRIORITY ISSUES APPLICABLE TO DESIGN Issue a

Title II.B.2 Safety Review Consideration - Plant Shielding to Provide Post-Accident Access to Vital Areas II.B.3 Safety Review Consideration - Post-Accident Sampling II.E.1.1 Auxiliary Feedwater System Evaluation II.E.1.2 Auxiliary Feedwater System Automatic, Initiation and Flow Indication I I. E.,3.1 Decay Heat Removal - Reliability of Power Supplies for Natural Circulation II.E.4.1 Containment Design - Dedicated Penetrations II.E.4.2 Containment Design - Isolation Dependability II.F(1)-(6)

Additional Accident Monitoring Instrumentation w-II.F.3 Instrumentation for Monitoring Accident Co'nditions II.G.1 Power Supplies for Pressuriter Relief Valves, Blocl Valves ano Level Indications II.K.3(16)

Recuction of Challenges and Failures of Relief Valves 1

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Modification of ADS Logic II.K.3(25)

Effect of loss of AC Power on Pump Seals III.A.1.2 Upgrace Licensee Emergency Support Facilities 111.D.1.1(1)-(3)

Primary Coolant Sources Outside the Containment III.D.3.3(1)-(4)

Inplant Radiation Monitoring l

III.D.3.4(1)-(2)

Control Room Habitability l

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TABLE 3 MEDIUM. PRIORITY ISSUES APPLICABLE T3 DESIGN CERTI Issue

-Title

1. 16 BWR main Steam Isolation Valve Leakage Control Systems

1. 20 Effects of Electromagnetic Pulse on Nuclear Power Plants A-15 Primary Coolant System Decontamination and Steam Generator Chemical Cleaning A-21 Main Steam Break Inside Containment - Evaluation of Environmental conditions for Equipment Qualification A-32 Missle Effects B-19 Thermal Hydraulic Stability B-26 Structural Integrity of Containment Penetrations Best Assessment of Inelastic Analysis Techniques for Equipment and components B-58 Passive mechanical Failures,<..

B-59 Review of (n-1) Loop Operation in BWRs and PWRs B-64 Decommissionfhg of Reactors B-73 Monitoring for Excessive Vibration Inside the Reactor Vessel C-8 Main Steam Line Leakage Contr.ol System D-1 Advisability of a Seismic Scram I.D.3 Control Room Design - Safety System Status Monitoring I.D.5(1)-(5),

Control Room Design - Improved Instrumentation Research I.E.8 Human Error Rate Analysis II.C 3 Risk Assessment Systems Interaction II.C.4 Risk Assessment - Reliability Engineering

TABLC 3 MEDIUM PRIORITY ISSUES APPLICABLE TO D Issue Title II.E.3.5 Decay Heat Removal - Regulatory guide II.F.2 to Inadequate Core CoolingIdentification of and R II.F.5 Classification of Instrumentation, Control and Electrical Equipment II.K.3(1)

Operational TestInstall Automatic PORV Isolation system and II.K.3(5)

Continue to Study Need for Automatic Trip of RCPs II.K.3(10)

Anticipatory Trip modification to Confine Range of Use to High Power Levels II.K.3(19)

~ Interlock on Recirculation Pump Loops w.

II.K.3(22)

Automatic Switchover of RCIC System Suction II.K.3(27)

Provide Common Reference Level for Vessel Instrumentation

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II.K.3(40)

Small-Break L'OCA. Evaluation of RCP Seal Damage and Le 5

III.A.3.4 Nuclear Data Link III.D.2.4(1)-(2)

Offsite Dose Measurements

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O TABLd 4 LOW PRIORITY GENERIC ISSUES APPLICAB

1. 1 B-5 II.E.2.1

1. 2

.B-8 II.E.2.3

1. 3 B-22 II.E.4.3

1. 4 B-27 II.E.6

1. 7 B-29 II.F.4

1. 8 B-30 III.D.1.3(1)-(4)

1. 10 B-31 III.D.2.1(1)-(3)

1. 12 B-32 III.D.2.2(1)-(4)

1. 13 B-35 III.D.2.3(1)-(4)

1. 17 B-47

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1. 21 6-49 i

1. 22 B-50 4

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1. 26, B-54

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1. 27 B-61 A-14 B-68 A-16 B A-18 C-2 A-19 C-9 A-22 C-10 A-37 C-12 A-38 C-14 A-41 C-15 B-3 I.D.4 i,,,,,,-,.-----w,+r-.*m,..

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TABLL 5 RESOLVED GENERIC ISSUES APPLICABLE TO DE' S

1. 6 B-21 II.K.3(9)

1. 14 B-25 II.K.3(11)

1. 25

'B-33 II.K.3(12)

A-2 B-36 II.K.3(14)

A-10 B-48 II.K.3(20 )

A-11 B-53 II.K.3(21)

A-13' B-56 II.K.3(29)

A-23 B-60 II.K.3(43)

A-24 B-62

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A-25 B-63 w-A-25 B-66 A-27 B-70

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A-23 B-71 A-30 C-1 N-A-31 C-5 A-34 C-6 A-35 C-7 A-36 C-17 A-42 0-3 B-6 II.B.7 B-9 11.0.3 B-11 II.E.1.3(1)-(2)

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TABLE 6 GENERIC ISSUES NOT APPLICABLE TO DESIGN CERTI

~

1. 5 B-23 C-13 I.B.2.1

1. 9 B-24 C-16 I.B.2.2

1. 11 B-28 D-2 1.3.2.3

1. 19 B-34 I.A.1.1 1.B.2.4

1. 24 B-37 1.A.1.2 I.C.1(1 }-( 4)

1. 28 B-38 I.A.1.3 I.C.2 A-5 B-39 I.A.l.4 I.C.3 A-6 B-40 I.A.2.1(1)-(3)

I.C.4 A-7

,y, B-41 I.A.2.2 I.C.5 A-8 B-42 I.A.2.3 I.C.6 A-20 B-43

1. A.274 -

I.C.7 A-33 B-44 1.A.2.5 I.C.8 A-46 B-45 I.A.2.6(1)-(6)

I.C.9 B-1 B-46 1.A.3.1 1.D.6 B-2 B '52 1.A.3.2 I.E.1 B-4 B-55 1.A.3.3 1.E.2 B-7 B-57 I.A.3.4 I.E.3 B-14 B-65 1.A.3.5 I.E.4 B-15 B-67 I.A.4.1(1)-(2)

I.E.5 B-16 B-69 I.A.4.2(1)-(4)

I.E.6 m u mm mma um ai m um

d TABi.E 6 (C0tiTItlUED)

B-17 C-3 I.B.I.1(1)-(7)

I.F.2 B-18 C-4 I.B.1.2(1)-(2)

I.G.1 B-20 C-11 I.B.1.3 1.G.2 II.A.1 II.J.2.1 II.K.3(33)

III.A.1.3 II.A.2 II.J.2.2 II.K.3(34)

III.A.2.1(1)-(4) 11.B.4 II.J.2.3 II.K.3(35)

III.A.2.2 II.B.5(1)-(3)

II.J.3.1 II.K.3(36)

III.A.3.1 II.B.6 II.J.3.2 II.K.3(37)

III.A.3.2 II.C.1 II.J.4.1(1)-(4)

II.K.3(38)

III.A.3.3(1)-(6)

II.C.2 II.K.2(1)-(21)

II.K.3(39)

III.'A.3.5 11.~D.1 II.K.3(2)

II.K.3(41)

I I I.'A. 3. 6 II.D.2 II.K.3(3)

II.K.3(42)

III.B.1 II.E.2.2(1)-(6)

II.K.3(4)

II.K.3(44)

III.B.2

-II.E.3.2 II.K.3(6)

II.K.3(45)

III.C.1 II.E.3.3 II.K.3(7) ' -

II.K.3(47)

III.C.2 II.E.3.4 II.K.3(8)

II.K.3(48)

III.D.' 2. 5 II.E.5.1 II.K.3(13)

II.K.3(49) 111.0.2.6 I1.E.5.2 II.K.3(15)

II.K.3(50)

III.D.3.1 1

III.D.2(1)-(4)

II.H.1 II.K.3(17)

II.K.3(51)

III.D.3.5(1)-(3)

II.H.2 I1.K.3(23)

II.K.3(52)

IV.A.1 II.H.3 II.K.3(24)

II.K.3(53)

IV.A.2

• • we,.

M w-w

_ - ~,,, -... - _ __

i TABLE 5 (C0tlTINUED) t 4-I II.H.4(1)-(2)

II.K.3(26)

II.K.3(54)

IV.a.1 t

l II.J 1.1 II.K.3(28)

II.K 3(55)

Iv.c.1 II.J.1.2 II.x.3(30)

II.x.3(ss) ty,o,1(1)_(3)

II.J.1.3 II.K.3(31)

II.D.3(57)

IV.E.1 II.J.1.4 II.K.3(32)

III. A. I.1(1)-(2)

IV.E.2

. IV.E.2 IV.E.4 i

IV.F.1 IV.F.2 IV.G.1 ss>.

IV.G.2 IV.G.3

.IV.G.4 IV.H N-i i

4 w

e 4

4 I

'% e.

A A

+-.,,~erme,

,,--+r~,

-,e-,+yy-,,_3,,.-,%-+,

,wy,,

-,.m-w-

---r7 e-'v-v+N"Fw * = - t e m t~r

---+-ve**

rw

+ - - - -

-w-w'*e--&-

e aw-**

• - - - - -=-&*

• '~~

-w'*m*

~

TABLC 7 POTENTIAL PROBLEM AREAS STEAM GENERATORS

1. 18 Steam Line Break with Consequential Small LOCA A-3 Westinghouse Steam Generator Tube Integrity A-4 CE Steam Generator Tube Integrity

. Primary Coolant system Decontamination and Steam Generator A-15 Chemical Cleaning a

MATERIALS

1. 14.

PWR Pipe Cracks

1. 15 Radiation Effects on Reactor Vessel Supports A-10 BWR Feeowater Nozzle Cracking A-11 Reactor Vessel Materials Toughnes Fracture Toughness of Steam Generator and Reactor A-12~

Coolant Pump Supports A-42' Pipe Cracks in Boiling Water Reactors

~

C-7 PWR System Piping

_ CONTROL ROOM DESIGN-

~

I.D.1(1)-(5) Control Room Design Reviews I.D.2 Control Room Design - Plant Safety Parameter Display Console I.D.3 Control Room Design - Safety System Status Monitoring 1.D.5(1)-(5)

Control Room Design - Improved Instrumentation Research I.E.8 Human Error Rate Analysis III.A.3.4 Nuclear Data Link III.D.3.4 (I)-(2)

Control Room Habitability

TABli 7 POTENTIAL PROBLEl AREAS (CON'T)

CONTAINMENT A-23 Containment Leak Testing A-39 Determination of Safety Relief Valve Pool Dynamic Loads A-43 Containment Emergency Pump Performance l

B-9 Electrical Cable Penetrations of Containment B-10 Behavior of Mark III Containments

~

4 B-12 Containment Cooling Requirements (Non LOCA)

II.E.4.1 Containment Design - Dedicated Penetrations II.E.4.2 Containment Design - Isolation Dependability II.E.4.4

({}-(5)

Containment Design - Purging 111.0.1.1 (1)-(3)

Primary Coolant Sources Outside Containment

~~.

ENVIRONMENTAL QUALIFICATION A-21 Main Steam Line Brelk Inside Containment - Evaluation Environmental Conoitions for Equipment Qualification A-24 Qualification of Class IE Safety Related Equipment

' A-40 Seismic Design - Short Term Program A-48 Safety EquipmentHydrogen Control Measures and Effects of Hydrog C-1 Assurance of Continuous Long Term Capability of Hermetic Seals on Instrumentation and Electrical Equipment II.F.5 Classification of Instrumentation, Control and Electrical Equipment A-

r TABLE 7 POTENTIAL PROBLEM AREAS (CON'T)

INSTRUMENTATION AND CONTROL

1. 20 Effects of Electromagnetic Pulse on Nuclear Power Plants A-25 Non-Safety Loads on Class IE Power Sources A-34 During AccidentsInstruments for Monitoring Radiation and Proce A-47 Safety Implication of Control Systems 11.0.3 Coolant System Valvss - Valve Position Indication II.F.1(1)-(6) Additional Accident Monitoring instrumentation II.F.3 Instrumentation for Monitoring Accident Conditions II.G.1 Power Supplies for Pressurizer Relief Valves, Block Valves and Level Indications

- ea -

II.K.1 TMI - IE Sulletins II.K.3 Final Recommendations 880 Task Force 111.0.2.4 (1)-(2)

Offsite Dose Measu,rements

~ ~..

e 6

==m M

e