ML19207B455

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Responds to NRC 790814 & 22 Requests for Clarification Re 790731 & 0806 Ltrs.Forwards Replacement to Attachment 1 of 790731 Ltr Re Positions in App a of NUREG-0578.Forwards Response to HR Denton 790820 Memo Re NUREG-0578
ML19207B455
Person / Time
Site: 05000471
Issue date: 08/24/1979
From: Howard J
BOSTON EDISON CO.
To: Parr O
Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 7908290480
Download: ML19207B455 (42)
v • d • e


Text

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7 BOSTON EDISON COMPANY bod BovLaTON STREET BOSTON. MABBACHUBETTs 02199 J. EDWARD HOWARD wies case eam, seus6sas August 24, 1979 Director of Nuclear Reactor Regulation Attention: Mr. O. D. Parr, Chief Branch No. 3 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Additional information Related to Pilgrim Unit 2 (Docket No.50-47I)

Dear Mr. Parr:

The attachments are provided in response to August 14, 1979 and August 22, 1979 NRC Staff requests for further information and clarification related to our letters of July 31, 1979 and August 6, 1979.

Attachment #1 replaces Attachment #1 to our July 31, 1979 letter. It responds to the " positions" in. Appendix A of NUREG-0578 instead of the summarized versions of the recommendations found in Section 2 of the report.

Attachment #2 responds to the August 20, 1979 memorandum from Mr. H. R.

Denton to the NRC Commissioners which also addresses NUREG-0578.

Very truly yours,

&W fG4(/cly Commonwealth of Massachusetts)

County of Suffolk )

Then personally appeared before me J. E. Howard, who being duly sworn, did state that he is Vice President-Nuclear of Boston Edison Company, an Applicant herein, that he is duly authorized to execute and file the foregoing letter in the name and on behalf of Boston Edison Company and the other Applicants herein and that the statements in said letter are true to the best of his knowledge and belief.

! A L (k (

Notary PubIie/

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My~ Commission Expires: July 6, 1984 8 5 ;., 255 7908290FFd

BDSTON EDISDN CO M PANY Mr. O. D. Parr August 24, 1979 Page 2 Attachments:

1. 33 Sheets
2. 5 Sheets cc: Service List t

853 256

!sosirow Eo sqw co.wawy Distribution List: ,

George H. Lewald, Esq. 1.. G. Cummings Vice President Ropes & Gray Marsh & McLennan, Inc.

225 Franklin Street 1221 Avenue of the Anericas Bostop, Massachusetts 02110 New York, New York 10020 The Bo'ard of Selectmen W. R. Bisson, Vice President Town of Plymouth Montaup Electric Ccrnpany 11 Lincoln Street P. D. Box 391 Plymouth, Massachusetts 02360

  • Fall River, Massachusetts D2722 Robert H. Culp G. D. Gowdy, Pro.iect Engineer Lowenstein, Newman, Reis, Axelrad & Toll ' Stone & Webster Engineering Corp.

1025 Connecticut Ave., N.W. P. O. Box 2325 Suite 1214 Boston, Massachusetts 02107 Washington, D. C. 20036 Combustion Engineering, Inc.

Charles Brinkm"), Manager 1000 Prospect Hill Road Combustion Engineering, Inc. Windsor, Connecticut 06095 Nuclear Licensing Office At'tn: Mr. E. P. Mailman Triangli Towers '

Sui te A-1 kobert Waaczyk 4853 Cordell Ave. Yankee Atemic Electric Comoany '

Bethesda, Maryland 20014 Seabrook naclear Station 20 Turnpike Road Charles Bardes Westboro, Massachusetts 01581 NELIA The Exchange Thomas C. Stewart Farmington Avenue M&'t Protection Consultants Farmington, Connecticut 05032 200 Clarendon Street Boston, Massachusetts 02116 J. E. Booker Gulf State Utilities Co. Bruce W. McKinnon, Manager P. O. Box 2951 Conraunity Power Development Dept.

Beaumont, Texas 77704 Mass. Municipal Wholesale Electric Co.

Stony Brook Energy Center John J. Carney Post Office Box 426 Huclear Energy Property Ludlow, Massachusetts D1056 Insurance Association B5 Woodland Street John L. McLean Hartford, Connecticu' 06102 Teledyne Engineering Services 303 Bear Hill Road Dr. Charles Cole - Waltham, Massachusetts 02154 Holdsworth Hall - Natural Resources Center Loren K. Stanley Amherst, Massachusetts 01002 Nuclear Services Corp.

477 Division Street Les W. Cooley Cam'pbell, California 95008 EDS Nuclear -V 220 Montgomery Street _-

San Francisco, California 94104 -

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853 257

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bsinw Enisaw ca.wawv 2.

- - Paul Gorman C. T. flain, Inc.

William H. Dormer -

  • Mass. Dept. of Public Safet,y 8th Floor .

101 Huntington Avenue 1010 Comonwealth Avenue Boston, Massachusetts D2199 Boston, Massachusetts D2215 Directopate of Licensing John D. Fassett, President U.S. Nuclear Regulatory Comission The United Illuminating Co.

80 Temple Street o Phillips Building New Haven, Connecticut 06506 7920 Norfolk Avenue Bethesda, Maryland 2DD34 Attn: Emanuel Licitra Mr. Phillip C. Otness, General Mgr.

Mass.11unicipal Wholesale Electric Co.

Gerald S. Parker Stony Brook Energy Center Director, Radiation Control Programs Post Office Box 426 Mass. Dept. of Public Health  :- Ludlow, Massachusetts D1506 Room B35 B0 Boylston Street Boston, Massachusetts 02116 Dean P. Amidon Mass. Dept. of Public Works Bechtel Power Corporation - Division of Waterways 100 Nashua Street, Room 529 Post Office Box 3965 Boston, Massachusetts D2114 c/o Central Receiving -

San Francisco, Califormia 94119 Attn: Mr. B. N. Pusheck Mr. ' James B. Nuckerheide 755 Boylston Street Gerald E. Anderson, President Suite 306 New Bedford Gas A Edison Light Co. Boston, Massachusetts 02116 Post Office Box 190 Cambridge, Massachusetts 02139

. Ralph M. Wood, Esq.

Public Service Company of N.H.

1000 Elm Street Manchester, New Har.pshire R3105' Neil Todreas Nuclear Engineering Dept.

Room 24-109 Mass. Institute of Technology 77 Massachusetts Avenue Cambridge, Massachusetts 02139 Joel Watson '

Environmental Research & Technology 696 Virginia Road Concord, Massachusetts D1742 Anthony D. Cortese, Comissioner Dept. of Environmental Quality Engrg.

100 Cambridge Street Boston, Massachusetts 02108 853 258 9

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BOSTON EDIBDN COMPANY Ooo BovLsToN STACCT DosToN. M AssACHUSCTTs 02199 J. EDWARD MQWARD m u. -

August 24, 1979 Director of Nuclear Reactor Regulation Attention: Mr. O. D. Parr, Chief Branch No. 3 U. S. Nuclear Reguistory Commission '

Washington, D. C. 20555 Additional infonnation Related -

to Pilgrim Unit 2 ,

(Docket No.50-47I)

Dear Mr. Parr:

The attachments are provided in response to. August 14, 1979 and August-22, 1979 liRC 51aff requests for further information and clarification related to our letters of July 31, 1979 and August 6, 1979.

Attachment #1 replaces Attachment #1 to our July 31, 1979 letter, it respwds to the " positions" in Appendix A of NUREG-0578 instead of the semmarized versions of the recermendations found in Section 2 of the report.

Attachment #2 responds to the August 20, 1979 memorandum f rom Mr. H. R.

Denton to the NRC Commissioners which also addresses NUREG-0578.

Very truly yours, vWtt Vd' -

c Commonwealth of Massachusetts)

County of Su+ folk )

Then personally appeared before me J. E. Howard, who being duly sworn, did state that he is Vice President-Nuclear of Boston Edison Company, an Applicant herein, that he is duly authorized to execute and file the foregoing letter in the name and on behalf of Boston Edison Company and the other Appilcants herein and that the statements in said letter are true to the best of his knowledge and belief.

! n i/.i ml. bi. ,

Notary Public '

My Commission Expires: July 6, 1984 853 .259

BosT5N EDISDN COMPANY

\

Mr. O. D. Parr August 24, 1979 --

Page 2 Attachments:

1. 33 Sheets
2. 5 Sheets cc: Service List

'A 853 260

ostuu toisqu romwy Distribution Listi ,,-

1. S. Ctmings, Vice resident George H.1. ewald, Esq. Mrsh & McLennan, Inc.

Ropes & Gray 1221 Avenue of the kericas 225 Franklin Street New York, New York 10020 Bostop,, Hassachusetts 02110 W. R. Bisson, Vice President The Board of Selectmen Montaup Electric Company Tcwn of Plymouth P. D. Box 391 11 Lincoln Street Fall River, &ssachusetts D2722 Plymouth, &ssachusetts 02350 * .

G. D. Gowdy, Pro.iect Engineer Robert H. Culp 3 tone 1 WebsterIngineering Corp.

Lowenstein, Newran, Reis. Axelrad & Toll 1025 Connecticut Ave., N.W. P. D. Box 2325

- Boston, Massachusetts 02107 Suite 1214 Washington, D. E. 20035 Combustion Engineering, Inc.

Charles Brinkman, Manager 1000 Prospect Hill Road Combustion Engineering, Inc. Windsor. Ionnecticut 05095 Nuclear licensing Dffice Attn: Mr. I. P. Mailman Triangle Towers Robert Wanczyk ~.

5uite A-1 Yankee Atenic Electric Cormany 4853 Cordell Ave. Seabrook Nuclear Station Sethesda, bryland 20014 20 Turnpike Road Westboro, Massachusetts D1581 Iharles Bardes NILIA Thomas C. Stewart The Exchange Farmington Avenue M&'t Protection Consultants Farmington, Connecticut 06032 200 Clarendon Street Boston, Massachusetts 02115 J. I. Booker ' Bruce M. McKinnon, ttanager Gulf . State iftilities In. Cornunity Power Development Dept.

P. D. Box 2951 Mass. Municipal Wholesale Electric fo.

Beaumont, Texas 77704 Stony Brook Energy Center John J. Carney Post Office Box 426 Ludlow, kssachusetts DID55 Nuclear Energy Property Insurance Association John L. McLean 85 Woodland Street Teledyne Engineering Services Hartford, Connecticut 05102 303 Bear Hill Road

- Waltham, kssachusetts D2154 Dr. Charles tole Holdsworth Hall - 8(atural Loren K. Stanley Resources Center Nuclear Services Corp.

Amherst, Massachusetts 01002 477 Division Street Cashell, California 95008 les W. Cooley EDS Nuclear -

220 Montgomery Street -

San Francisco, California 94104

- 853 261

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Paul Gorman

- C. T. Main, Inc.

William H. Dormer a Mass. Dept. of Public Safety 8th Floor . .

101 Huntington Avenue 1010 Corronwealth Avenue Boston, mssachusetts D2199 Boston, Massachusetts D2215 John D. Fassett, President Directo, rate of Licensing The United Illuminating Co.

U.S. f:uclear Regulatory Ctmission o Phillips Building 80 Temple Street New Haven, Connecticut D6506 7920 Norfolk Avenue Bethesda, kryland .20D34 Mr. Phillip C. Otness, General' Mgr.

Attn: Emanuel Licitra Mass. Yunicipal %'holesale Electric Co.

Gerald S. Parker Stony Brook Energy Center Dimctor, Radiation Contml Progras Post Dffice. Box 426 -

Mass. Dept. of Ptblic Health -

Ludlow, &ssachusetts V15DS Room 835

  • 80 Boylston Street Boston, Massachusetts 02116 Dean P. Amidon 8.ns. Dept'. of Public Works ~

Bechtel Power Corporation - Division of Waterways 100 Nashua Street, Room.529 Post DTfice Box 3955 Boston, Massachusetts D2114 c/o Central Receiving --

San Francisco, Califomia 94119 .

Attn: Mr. B. N. Pusheck Mr.' James B. Nuckerheide 755 Boylston Street Gera'id E. Anderson, President Suite 306 New Bedford Gas & Edison Light Co. Boston, m ssachusetts D2116

. Post Office Box 190 Cambridge, Massachusetts 02139 Ralph M. Wood, Esq.

Public Service Company of H.H.

2000 Elm Street .

Manchester, New Haz:pshire .II31D5 Neil Todreas Nuclear Engineering . Dept.

Room 24-109 Mass. Institute of Technology 77 Massachusetts Avenue Cambridge, Massachusetts 02139 Joel Watson

  • Environmental Research & Technology 696 Virginia Road Concord, mssachusetts V1742 Anthony D. Cortese, Comissioner Dept. of Environmental Quality Engrg.

100 Cambridge Street Boston, Massachusetts 02108 853 262

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ATTACH'4ENT #1 Sh. 1 of 33 PILGRIM UNIT 2 BOSTON EDISON COMPANY DOCKET NO. 50-471 The following additional Information provides the NRC with the necessary assurance that Pi.lgrim Unit 2 will meet the intent of NUREG-0578, "TMI-2 LESSONS LEARNED TASK FORCE REPORT AND SHORT-TERM RECOMMENDATIONS," Ju ly 19, 1979.

This replaces Attachment #1 to our July 31, 1979 letter.

WhiIe the commitments to each appIicabIe position are addressed below, the implementation details will be described more fully in the FSAR.

853 263

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Sh. 2 of 33 NRR tessons tearned Task Force Short-Term Recommendations ,

TITLE: Emeroency Power Sucoly Requirements for the Pressurizer Heaters.

  • Power-Daerated Relief Valves and Block Valves and Pressurizer Level Indicators in PWRs (Section 2.1.1) --
3. P'051 TION Consistent with satisfying the requirements of General Design Criteria 'LD,14, 25,17. and 20 of Appendix A to 20 CFR Part 50 for the event of loss of offsite power, the following positions shall be implemented:

3.1 Pressurizer Heater Power Supply l

1. The pressurizer heater power supply design shall provide the capab111ty to supply, from either the offsite power source or the emergency pcwer source (when offsite power is not available), a predetermined number of, prMsmizer heaters and associated controls necessary to -

establish ano maintain natural circulation at hot standby conditions.

The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability. c  ;

2. Procedures and training shall be established to make the operat$r aware of when and how the required pressurizer heaters shall be connected to the emergency buses. If required, the procedures shall identify under what conditions selected emergency loads can be shed -

Trem the emergency power source to provide sufficient capacity for cne connection of the pressurizer heaters.

3. The time required to accomplish the connection of the preselected pressurizer heater to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.
4. Pressurizer heater motive and control power interfaces with the emeagency buses shall be accomplished through devices that have been qualified in acccrdance arith safety grade requirements. m 3.2 Power Suoply for Pressuriter 4telief and Block Valves and Pressurizer Level Indicators
2. Motive and control co=ponents af the power-sperated relief valves (PORVs) shall be capable of being supplied from either the offsite

. power source or the emergency power source when the offsite power is not available.

2. Motive and control components associated with the PORY block valves shall be capable of being supplied from either the offsita power source or the emergency power sourte when the offsite power is isot available. .
3. Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety grade requirements.
4. The pressurizer level indication instrument channels shall be powered from the vital instrument buses. Thesesbuses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not available.

853 264-

Sh. 3 Ed 33 5

RESPONSE

Pressurizer Heater Power Supply

1. The pressurizer heater power supply design will have the capability to supply, f rom either the offsite power source or the emergency power source (when of fsite power is not available), a predetermined number of pressurizer heaters and associated controls necessary to estab!!sh and reintain natural circulation at hot standby conditions.

. The required heaters and their controls will be connected to the emergency buses in a manner that will provide redundant power supply ca pab i l i ty.

2. Procedures and training will be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency buses. I f requi red, 'the procedures wi ll identi f y under what conditions selected emergency loads can be shed from the emergency power source to provide suf ficient capacity f or the connection of the pressurizer heaters.
3. The time required to accomplish the connection of the preselected pressurizer heater to the emergency buses will be consistent with the tirely initiation and maintenance of natural circulation conditions.
4. Pressurizer heater motive and control power interfaces with the emergency '

buses will be accomplished through devices that have been qualified in accordance with saf ety-grade requir'ements.

Power Supoly for Pressurizer Relief and Block Valves and Pressurizer Level Indicators

1. Motive and control components of the power-operated relief valves (PORV's) will be capable of being supplied from either the of fsite power source or the emergency power source when the of fsite power is not available.
2. Mctive and control components associated with the PORV block valves will be capable of being supplied f rom either the of fsite power source or the emergency power source when the of fsite power is not available.
3. Motive and control power connections to the emergency buses for the PORV's and their associated block valves will be through devices that

.. have been qualified in accordance with safety-grade requirements.

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4. The pressurizer level indication instrument channels will be powered f rom the vital instrument buses. These buses will have the capability of being supplied f rom either the offsite power source or the emergency power source when of f site power is not available.

853 .265

Sh 4 of 33

,,. RRR Lessons Learned Tas'k Force Short-Term Recommendations TITLE: Performance Testino for BWR and PWR Relief and Safety _...

Valves (Section 2.1.2) v

3. POSITION Pressurized water reactor and boiling water reactor licensees and applicants shall cinduct testing to qualify the reactor coolant system relief and safety valves under expected operating 1 conditions for design basis transients and accidents. The licensees .and applicants shall determine the expected valve operating conditions thrcugh the use of analyses of accidents and anticipated

- operational occurrencer referenced in Regulatory Guide 1.70, P* vision 2. The single failures applied to these analyses shall be chosen so t. . the dynamic forces on the safety and relief valves are maximized. Test pressures shall be the highest predicted by renventional safety analysis procedures. Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry piping and supports as well as the valves themselves.

t RESPONSE '

Pilgrim Unit 2 will support the industry efforts to conduct testing necessary to qualify the reactor coolant system relief and safety valves under expected

operating conditions for design basis transients and accidents. The expected valve operating conditions will be determined through the use of analyses of accidents Guide and 1.70, anticipated Rev. operational occurrences referenced in Regulatory

2. The single failures applied to these analyses will be chosen so that the dynamic forces on the safety and relief valves are maxi-

. mized.

analysis.Test pressures will be the highest predicted by conventional safety Reactor coolant system relief and safety valve qualification will include qualification of associated control circuitry, piping er.d supports necessary for proper valve performance, as well as the valves themselves.

853 266

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Sh. 5 of 33 s

NRR_ Lessons Learned Task Force Short-Term Recommendations

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TITLE: Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWRs and BWRs (Section 2.1.3.a)

3. POSITION Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detartion device or a reliable indication of flow in the discharge pipe.

RESPONSE

Reactor system reli$f and safety ~ valves will be provided with a positive 'in-dication in the control room derived f rom a reliable valve position detection device or a reliable Indication of flow in the discharge pipe.

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853 267 A

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. NRR Lessons Learned Task Force

'Short-Term Recommendations

  • TITLE: Instrumentation for Detection of Inadequate Core Coolino in PWRs and BWRs (Section 2.1.3.b)
3. POSITION L ifcensees'shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instru-mentation. The. licensee shall provide a description of the existing instrumentation for the operators to use to recognize these conditions.

A detailed description of the analyses needed to form the basis for operator training and procedure development shall be provided pursuant to anottjer aort-term requirement, " Analysis of Off-Normal Conditions, Including hastral Circulation" (see Section 2.1.9 of this appendix).

In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of coolant saturation condition.

Operator instruction as to use of thi,s meter shall include consid-eration that is not to be nsed axclusive of other related p1 Ant para:neters.

2. Licensees shall provide a description of any additional instrumenta-tion or controls (primary or backup) preposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling.

'A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed , equipment, the analysis used in developing these procedures, and a schedule for installing .he equipment shall be provided.

RESPONSE

1. Procedures to be used by the operator will be developed to recognize inadequate core cooling. A detailed description of the analyses which 7:m the basis for operator training and procedure developrent, and a description of the instrumentation for the. operators to use to recognize these conditions will be provided in the FSAR.

Instrumentation will be installed in the control room which will indicate the approach of the reactor coolant system to saturation conditions.

Operator instructions as to the use of this instrumentation will stress that it not be used exclusive of other related plant parameters.

2. Since the final design is not complete, any aspect of position 7 above will be appropriately addres;ed by our commitment in response to position 1 above.

-853 268

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Sh. 7 of 33 NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Containment Isolation Provisions for PWRs and BWRs (Section 71.4) 1

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3. POSITION 3.. All containment isolation system designs shall comply with the recommendations of SRP 5.2.4; 1.e., that, there be diversity in the parameters sensed for the initiation of containment isolation.
2. All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identify each system determined to i

. be non-essential, shall describe the basis for selection of each essential . system, shall modify their containment isolation designs accordi6 gly, and shall report the results of the re-evaluation to the NRC.

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3. All non-essential systems shall be automatically isolated by ihe containment isolation signal.

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4. The design of control systems for' automatic containment isolation valvos shall be such that resetting the isolation signal wii'. not '

result in the automatic reopening of containment isolation valves.

Reopening of containment isolation valves shall require deliberate operator action.

RESPONSE //

1. The Pilgrim Unit 2 containment isolation system design will comply with the recommendations of SRP 6.2.4; i .e. , that there be diversity in the parameters sensed for the initiation of containment isolation.
2. Careful consideration to the definition of essential and non-essential systems shall be given, and each system shall be identified as to whether it is essential or non-essential . The bases for the selection of each essential system shall be provided.
3. All non-essential systems will be automatically isolated by the con-tainrent isolation signal.
4. The design of control systens for automatic containmeni isolation valves will be such that resetting the isolation signals will not result in the autoratic reopening of cor.tainnent isolation valves. Reopening of con-tainrent isolation valves will require deliberate operator action.

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  • Sh. 8 of 33 .

NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems (Section 2.1.5.a)

3. 9055 TION flants using external recembiners or purge systems for post-accident combustible gas control of the containment atmosphere should/ provide contairment isolation systems for external recembiner or purge systems that are dedicated to that service only, that meet the redundancy and single failure requirements of General Design Criteria' 54 and 56 of Appendix A to 10 CFR Part 50 and that are sized is satisfy the . flow requirements af the recombiner or purge, system. ,

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RESPONSE

Post-accident combustible gas control of the containment atmosphere for Pilgrim Unit 2 will be performed by redundant Internal recombiners. Capability ~for post-accident containment ourging will also be provided. The containment iso-lation ,ystem for 1his coc-tainment purge will be dedicated only to combustible gas control, will meet the redundancy and single failure requirenents of General .

Design Cr iteria 54 and 56 of Appendix A to 10 CFR Part 50, and will be sized to satisfy the flow requirements of the purge s/ stem, i

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Sh. 9 of 33

. N_RR Lessons Learned Task Force Short-Term Recommendations -

TITLE: Inerting BWR Containments (Section 2.1.5.b) ~~..

3. POSITION It shall be required f. hat the Vermont Yankee and Hatch 2 Ma: k I BWR contain-sents be inerted in a manner similar to other operating BVR plants. Inerting shall also be required for near tetis DL licensing of Mark I and Mark II BWRs.

RESPONSE

This is not appli, gable to Pilgrim Unit 2.

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I 853 .271 .

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Sh. 10 of 33 s

NRR Lessons Learned Task Force

  • Short-Term Recommendations

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TITLE: Capability to Install Hydrogen Recombiner at Each Light hater Nuclear Power Plant (Section 2.1.5.c)

3. POSITION (Minority View)
1. All licensees of light water reactor plants shall have the capability to obtain and install,recombiners in their plants within a few days following an accident if containc.ent access is impaired and if such a system is needed for long-term post-accident combustible gas control.

. 2. The proce'dures and bases upon which the recombiners would be used on all plants should be the subject of a review by the licensees in considering sheilding requirements and personnel exposure limitations

, as demonstrated to be necessary in the case of THI-2.

RESPONSE ,

This is not applicable to Pilgrim Unit 2. Redundant internal recombiners are provided.

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853 272 1x

Sh. 11 of 33

,NRR Lessons Learned Task Force ,

Short-Term Recommendations ,

TITLES Integrity of Systems outside containment Likely to contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs (Section 2.1.6.a)

'l i 1

3. Pd5ITION Applicants and licensees shall immediately implement a v. uw m te Tetluce leakage from systems outside containment that would or could contain highly ,

radioactive fluids during a serious transient or accident to as-low-as practical levels. This program shall include the following:

.1. Immediate Leak Reduction ,-

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a. Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
b. Measure actual leakage rates wit,h system in operation and report them to the NRC.
2. Continuing Leak Reduction .

Establish and implement a program of preventive maintenance to reduce leakage to as-low-as practical levels.. This program shall includs periodic integrated leak tests at a frequency not to exceed refueling cycle intervals.

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RESPONSES

1. The design of systems outside conta'iment that would or could contain highly radioactive fluids during a serious accident or transient will be carefully reviewed. This review will include implementing all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment, to reduce leakage to as-low-as-practical levels. Actual leakage rates with the systems in operation will be measured and results reported to the NRC.
2. A program of preventive maintenance to reduce leakage to es-low-as-

. practical levels wt !! be established and implemnted. This program shall include periodic integrated leak tests at a f requency which will be established at the time Pilgrim Unit 2 receives an operating license.

853 273

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Sh. 12 of 33 4

NRR Lessons Learned Task Force ,

. Short-Term Recommendations t i

TITLE: Design Review of Plant Shielding of Spaces for Post-Accident -

Operatiens (section 2.1.6.b)

3. 90SITION With the ass'u mption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4, each licensee shall perform a 1 radiation and shielding design review of the spaces tround systems 1.The hat design way, as a result of an accidant, contain highly radioactive materials. ,,

review snould identify the location of vital areas and eculpment such as the control room, radwaste control stations, emergency power supplies, motor .

control centers, and instru=ent areas, in which personnel occupancy may be .

i unduly limited or safety equipment may be unduly degraded by the radiation ,

fields during post-accident cperations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall ~

determine which types of corrective actions are ne'eded for vital areas throughout the facility.

a t (1)

RESPONSE

With the assumption of a post-accident release of radioactivity equivalent to that described ,n Kegulatory Guide 1.4 (i .e. , the equvaient of 50% of the core radio-

, lodine and 100% of the core noble gas inventory are contained in the primary coolant) a radiation and shielding design review will be perf ormed of the spaces arount systems that may, as a result of an accident, contain highly radioactive materials.

The design review will identify the location of vital areas and equipment in which prsonnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operation of these systems.

Adequate access to vital areas and protection of safety eculoment by design changes, increased permanent or temporary shleiding, or post-accident procedural controls w i l l be p rovi ded. The design review will determine which types of corrective actions are needed for vital areas throughout the facility.

T1) This response corresponds to the revised position in attachment # 5to the memorandum f t om Mr. H. R. Denton to the NRC Commissioners, dated Aug. 20, 1979.

853 2L74 i

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. Sh, 13 of 33 NRR Lessons Learned Task Force Short-Term Recommendations

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TITLE: Automatic Initiation of the Auxiliary Feedwater System -

for PWRs (Section 2.1.7.a)

3. POSITION -

. Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system, the following requirements shall be implementad in the short term:

1. The design shall provide for the automatic initiation of the auxiliary

- feedwater system. *

2. The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function. ..,
3. Testability of the init. fating signals and circuits shall be a feature l cf the design. ,

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4. The initiating signals and circuits shall be powered from the emergency. 1 buses.
5. Manual capability to initiate the auxiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.
6. The a-c motor-driven pumps and valves in the auxiliary feedwater -

system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses. '

7. The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room. .

In the long term, the ' automatic initiation signals and circuits shall be upgraded in accordance with safety grade requirements. i

= *

~

f

. , n .

Sh. 14 of 33

RESPONSE

1. The design provides for the automatic initiation of the emergency f eedwater system. -,
2. The automatic initiation signals and circuits are designed so that a single failure will not result in the loss of emergency feedwater system function.
3. Testability of the initiating signals and circuits will be a feature of the design.
4. The initiating signals and circuits will be powered f rom the emergency buses.
5. bbnaal capability to initiate the emergency feedwater system f rom the control room is included in the design in such a manner that a single .

' failure in the manual circuits will not result in the loss of system function.

6. The a-c motor driven pumps and valves in the emergency f eedwater sysiem are included in the automatic actuation (simultaneous or sequential) of the loads to the emergency buses.
7. The automatic initiating signals and circuits are designed so that their ,

failure will not result in the loss of manual capability to initiate the emergency feedwater system f rom the control room.

8. The automatic initiation signals and circuits for the emergency feed-water system are in accordance with safety-grade requirements.

c3 J

276 893~~T 6

Sh. 15 of 33

'NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Auxiliary Feedwater Flow Indication to Steam Generators for PWRs (Section 2.1.7.b) s

3. POSITION Consistent with satisfying the requirements set 1 orth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements shall be implemented:
1. Safety grade indication of auxiliary feedwater flow to asch .stass _ _

generator shall be provided in the control room. .

2. The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of tha auxiliary feeawater system set '

forth in Auxiliary Systems Branch Technical Position 10-1 of the

. Standard Review 9isn. Section JD.4 9.

RESPONSE

1. Safety-grade Indication of emergency feedwater flow to each steam generator wiii be provided in the contrcl room.
2. The emergency feedwater flow instrument channels will be powered f rom the emergency buses consistent with satisfying the emergency power diversity requirements of the emergency feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.

gcg 277 h.

A

Sh. 16 of 33 NRR Lessons Learned Task Force

~S' hort-Term Recommendations TITH: 1eproved Post-Accident 3ampling tapability (Section 2.1.B.a)

3. -

' POSITION A design and operational review of the reactor coolant and containment atmosphere sampling systems shall be perfomed to detemine the capability of personnel to fromptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditic,ns without incurring a radiation exposure to any individual in excess of 3 and 18 3/4 Rems to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria. , ,

A design and operational review _of the radiological spectrum analysis facilities shall be perfomed to detemine the capability to promptly quantify (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) quantify certain radioisotopes that are indicatori of the degree of core damage. Such radionuclides are noble gases (which indicate cladding f ailure), iodines and cesiums (which indicate high fuel temperatures), and non-volatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum shou 1d correspond to a Regulatory fluide 1.3 nr 1.4 release.

The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct .

radiation from airborne effluents. If the review indicates that the analyses required cannot be perfomed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.

~

In addition to the radiological analyres, certain chemical analyses are necessary for monitbring reactor conditions. Procedures shall be provided to perfom boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source tem). Both analyses shall be capable of being completed promptly; i.e., the boron sample analysis within an hour and the chloride sample analysis arithin a shift.

o 3

Sh. 17 of 33 9

. (Section 2.1. 8.a)

RESPONSE

A design and operational review of the reactor coolant and containng.nt atmosphere sampling systems will be performed to determine the capability of pecsonnel to promptly obtain a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18 3/4 Rems to the whole body or extremities, respectively. Accident conditions will assume a Regulatory Guide 1.4 release of fission products, if the review Indicates that personnel could not promptly and safely obtain the samples, additional. design features or shielding will be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis f acilities will be performed to determine the capability to promptly quantify certain radioisotopes that are indicators of the degree of core damage.

Such radionuclides art noble gases (which indicate cladding f ailure). Iodines and cesluns (which indicate high fuel temperatures), and non-volatile isotopes (which Indicate fuel melting). _ Analysis of the initial reactor coolant spectrum will correspond to a Regulatory Guide 1.4 release. The review will also consider the effech- of direct rad'ation f ron, piping and components in The auxiliary building and possible contamination and direct radiation f rom airborne ef fluents.

If the review indicates that the analyses required cannot be performed in a prompt nenner with existing equipment, then design modifications or equipmrnt procurement will be undertaken to meet the criteria.

In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures will be provided to perform boron and chloride chemical analyses assuniing a highly radioactive initial sample (Regulatory Guide 1.4 source term). Both analyses will be capable of being com-pleted promptly.

O BSi$

tt. 18 of 33 9

. NRR Lessons Learned Task Force

$_ho{t-Term Recommendations TITLE: Increased Range of Radiation Monitors (section 2.1.B.b)

3. POSITION The requiremente associated with this recomrendation should be censidered as advanced,imple entation af certain requirements to be included in a revision to Regulatory Guide 1.97. " Instrumentation to Follow' the Course of an Accident."

which has cineady been initiated, and in etner Regulatory Guides, which will be prc=ulgaunt ta the near-term.

3. Nebie gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions; multiple monitors are considered,

., to be necessary to cover the ranges of interest. .

a. Noble gas effluent monitors with an upper range capacity of 105 pCi/cc (Xe-133) are considered to be practical and should be installed in r11 operating plants.
b. Noble gas effbent itenitoring shall be provided for the total range of concent ation extending from a minimum of 1D 7 pCi/cc (Xe-133) to a maxi.?um of 105 pCi/cc (Xe-133). Multiple monitors .

are considered to be recessary to cover the ranges of inter

  • t.

The range capacity of individual monitors shall overlap by .

Tactor of ten.

2. Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radiciodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.
3. In-containment radiation level monitors with a maximin. range of 10s rad /hr shall,be installed. A minimum of two sud 2JnitDTs that are physically reparated shall be provided. Monittes'shall be designed and gralified to function in an accident environment.

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  • 3 6

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Sh. 19 of 33 9

(Section 2.1.8.b)

(1)

RESPONSE

1. Noble gss of fluent monitors wiI L be installed with an extended range designed to f unction during accident conditions as well as Eiring normal operating conditions ; multiple monitors will be provided to cover the ranges of interest.

,,a. Noble gas ef fluent monitors with an upper r ange capacity of 10 uC1/cc (Xe-133) will be installed.

b. Noble gas ef fluent monitoring will be provided for the total range of concentration extending f rom nomal conditions (ALARA) concentrations to a maximum of 105 uCilce (Xe-133). Multiple monitors will be provided to cover the ranges of interest. The range capacity of individual monitors will overlap by a factor of ten.
2. Since iodine gaseous ef fluent inonitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radiolodines for the accident condition will be provided with campling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis,.
3. Monitors suitable for detection of in-containment radiation levels up to 10 6rad / hour w1II be provided. Such rnonitors wiII be redundant, '

physically separated, and will be quallfled to f unction in the accident environment to which they will be exposed.

(1) This response corresponds to the revised position in attachment p 5 to the memorandum f rom Mr. H. R. Denton to the NRC Ccmmissioners, dated Aug. 20. 1979.

(')

. Sh. 20 of 33 NRR tessons Learned Task Force Short-Term Recommendations TITL,E: Improved In-Plant Iodine Instrumentation (Section 2.1.8.c) ~--

3. POSITION Each licensee shall provide aquipment and as:.ociated training and procedures for accurately determining the airborne fodine concentration throughout the plant under accident conditions.

(1)

RESPONSE

Equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the f acility where plant personnel may be present during an accident will be provided using state-of-the-art techniques.

(1) This response corresponds to the revised position in attachment # 5 to the memorandum f rom Mr. H. R. Denton to the NRC Commissioners, dated Aug. 20, 1979 m

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Sh. 21 q ' 33 -

_NRR Lessons Learned Task Force >

_ Short-Term Recommendations

  • TITLE: Analysis of Design and Off-Normal Transients and Accidents ISection 2.1.9) ,
3. POSITION Analyses ' procedures, and trainin2 addressing the follo:.ing are required:
1. Small bmat less-of-coolant accidents;
2. Inadequate tore cooling; and
3. Transients and accidents.

Some analysis requiriments for small treaks have already been specified by the Bulletins and Orders Task Force. These should be completed. In addition',

pretest calculations of some of the Loss of Fluid Test (LOFT) small break tests (scheduled to start in September 1979) shall be performed as means to verify the analyses performed in support of the small break emergency procedures and in support of an eventaal long tem verification of compliance with-Appendix K af 3D CFR Part 50.

In the analysis of inadequate core cooling, the following conditions shall in -

analyzed using realistic (best-estimate) methods:

1 Low reactor coolant system inventory (two examples will be required -

LOCA with forced flow LOCA without forced flow).

2. Loss of natural circulation (due to loss of heat sink).

These calculations shall include the period of time during which inadequate core cooling is approached as well as the period of time during which inadequate core coolina exists. The calculations shall be carried out in real time far enough that all important phenomena and instrument indications are included. ,

Each case should then be repeated taking credit for corrnt operator action.

There additional cases will provide the basis for developing appropriate emergency procedures. These calculations should also provide the analytical basis for the design of any additional instrumentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy (see Section 2.1.3.b in this appendix).

The analyses of transients and accidents shall include the design basis events specified in Section 15 of each FSAR. The analyses shall include a single active failure for each system called upon to function for a particular event.

Consequential failures shall also be considered. Failums of the operators to

~

perform required control manipulations shall be given consideration for persuta-tions of the analyses. Operator actions that could cause the complete loss of function of a safety system shall also be considered. At present, these analyses need not address passive failures or multiple system failurt:s in the  %

short term. In the recent analysis of small break LOCAs, complete loss of CD

auxiliary feedwater was considered. The complete loss of auxiliary feedwater  %

may be added to the failures being considered in the analysis of transients and accidents if it is concluded that more is needed in operato: training  %

beyond the short-term actions to upgrade auxiliary feedwater system reliability. (t)

Similarly, in the long term, multiple failures and passive failures may be CD considered depending in part on staff review of the results . f the short-term analyses. s

, Sh. 22 of 33 The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system ,

response to operator actions is unclear or these calculations could be used to provide important quantitative information not available from an event tree. i For example, failure to initiate high pressure injection could lead to core uncovery for some transients, and a computer calculation could provide informa-tion on the amount of time available for cor'rective action. Reactor simulators i may provide some information in defining the event trees and would be useful l in studying the information available to the operators. The transient and accident analyses are to be performed for the purpose of identifying appropriate and inappropriate operator actions relating to important safety considerations such as natural circulation, prever. tion of core uncovery, and prevention of more serious accidents.

The information derived from the preceding analyses shall be included in the -

plant. emergency procedures and operator training. It is expected that analyses performed by the HSSS vendors will be put in the form of emergency procedure guidelines and that the changes in the procedures will be implemented by each licensee or applicant.

In addition to the analyses performed by the reactor vendors, analyses o'f selected transients should be performed by the NRC Office of Research, using the best available computer codes, to provide the basis for comparisons with the analytical methods being used by the reactor venders. These comparisons .

together with comparisons to data, including LOFT small break test data, will constitute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emergency procedures.

e

~

(Section 2.1.9) Sh. 23 of 33

RESPONSE

Analyses, procedures, and training will address the following: --

" Small break loss-of-coolant accidents; 1.

2. Inadequate core cooling, and i5. Transients and accidents.

In the analysis of inadequate core cooling, the following conditions will be analyzed using realistic (best-estimate) methods: ,

1. Low reactor coolant system inventory (two examples will be provided - LOCA with f orced f low, LOCA with forced flow).
2. Loss of natural d ?culation (due to loss of heat sink).

These calculations will include the period of time during which inadequate core cooling is approached as well as the period of time during which in-adequate core cooling exists. The calculations will be carried out far enough that all important phenomena and instrument indications are included.

Each case will then be repeated taking credit for correct operator action.

These additional cases will provide the basis for developing appropriate emergency procedures. These calculations will also provide the analytical -

basis for;the design of any additional instru.aentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy.'

The analyses of transients and accidents will include the design basis events speciti.ed in Section 15 of the FSAR including of f-normal conditions and events.

The analyses will include the most limiting single active failure for a particular event. Consequential f ailures will also be considered. Failures of the operators to perform required control manipulations will be con-sidered. Operator actions that could cause the complete loss of function of a safety system will also be considered. These analyses will not address passive f ai lures or multiple system f ai lures in the short term.

The transient and accident event tree methodology used by Boston Edison is the safety sequence diagram method. This technique will be continued to support operator training. Best-estimate transient and accident analyses will be performed to the extent required for the purpose of identifying appropriate ani inappropriate operator actions relating to important safety considerations such as natural circulation, prevention of core uncovery, and prevention of more serious accidents.

The Information derived from the preceding analyses will be included in the plant emergency procedures and operating tralhing. .

853 285

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

Sh. 24 of 33 NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Shif t supervisor's Responsib111 ties (Section 2.2.1.a)

3. POSITION ,
1. The highest level of corpor:te management of each licensee shall issua and periodically reissue a management directive that emphasizes the primary management responsibility of the shift supervisor for safe operation of the plant under all conditicas so bis shift Jind that clearly establishes his command duties. .

~ '

2. Plant procedures shall be reviewed to assure that the duties, responsi-bilities, and authority of the shift supervisor and control room

. operators are properly Afefined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel. Particular emphasis shall be placed on the following:

i

a. The responsibility and authority of the shift supervisor shall be to maintain the broadest perspective of operational conditions affecting the safety nf the plant as a matter of highest priority at all times when on duty in the control room. The idea shall '

be reinforced that the shift supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the control room.

t. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators. Persons autnorized to relieve the shift supervisor shall be specified.
c. If the shift supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the contral room cornand function.

These temporary duties, responsibilities, and authority shall be clear 1y speciffed.

I

3. Training programs for shift supervisors shall emphasize and reinforce the responsibility for safe operation and the management function

,. the shift supervisor is to provide for assuring safety.

4. The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility responsible for plant operations.

Administrative functions that detract from or are subordinate to the management responsibility for assuring the safe operation of the plant shall be delegated to other operations personnel not on duty in the control room.

853 286 O f

Sh. 25 of 33

  • ~

(Section 2.2.1.a)

RESPONSE

1. The Vice President - Nuclear of Post;n 7.dison will issue and par.iodically reissue a management directive inat emphasizes the primary management responsibility of the Nuclear hatch Engineer for safe operation of the plant under all conditions on his shif t and that clearly establishes his command duties.
2. PIant procedures will be developed and reviewed to assure that the duties, responsibilities, and authority of the Nuclear Watch Engineer, Operations Supervisor and control room operators are properly defined to af fect the establish-ment of a definite line of command and clear definition of the command decision authority of the Nuclear Watch Engineer in the control room relative to other plant management personnel. Particular emphasis will be placed on the following:
a. The responsibility and authority of the Nuclear Watch Engin~eer will be to maintain the broadest perspective of operational conditions af fecting the safety of the plant as a retter of highest priority at all times when on duty in the control room. The idea will be reinforced that the Nuclear Watch Engineer should not become totally involved in any single operation in times of , emergency when multiple operations are required in the control room.
b. The Nuclear Watch Engineer, until properly relieved, will remain in the control room at all times during accident situations to direct the activities of control room operators. Persons authorized to relieve the Nuclear Watch Engineer wi ll be 'specified.
c. li.f tha f{uclear Watch Engineer is temporari ly absent f rom the control room during operations, an Operations Supervisor will be designated to assume the control room command f unction. These temporary duties, responsibilities, and authority will be clearly specified.
3. Training programs for Nuclear Watch Engineers and Operations Supervisors will emphasize and reinforce the responsibility for safe operation and that the management function of the Nuclear Watch Engineer is to provide for assuring safety.
4. The administrative duties of the Nuclear Watch Engineer will be reviewed by the VP-Nuclear to assure that administrative functions that detract from or are subordinate to the Nuclear Watch Engineer's ranagement responsibility for assuring the safe and reliable operation of the plant are delegated to
  • personnel who are not directly responsible for control room operations.

853 287

Sh. 26 of 33 i

, . NRR Lessons Learned Task Force Short-Term Recom.endations

~

TITLE: Shift Technical Advisor (Section 2.2.1.b) .

3. ~ POSITION - I Each visor. licehsee shall provide an on-shift tec nical advisor to the shift super- i The shift technical advisor may serve more than one unit at a multi-unit site if'qua11 tied to perfore the advisar function for the marious nnits.

-l The shift tectmical advisor shall tave a bachelor's WegTwe wr equivalent in a scientific or engineering afiscipline and have received specific training in -

the response and analysis of the plant for transients and accidents. The shift technical advisor shall also receive training in plant design and layout, including 1 room.

the capabilities of instrumentation and controls in the control The licensee shall assign actmal stuties to the shift tectmical advisnes -  :

that pertain to the engineering. aspects of assuring safe operations of the  !

plant, including the review and evaluation of operating experience.

RESPONSE .

~

Pilgrim Unit 2 will comply with the necessary staffing requirements at the time t operating license issuance.

The information submitted in the FSAR will include details of how the following will be achieved.

1. Personnel capable of accident assessment based on comprehensive education in engineering and science subjects related to nuclear power plant design and on training and experience in the dynamic response of Pilgrim Unit 2 will be rapidly available in the control room.
2. An engineering group with diverse technical knowledge, experience and perspective in relevant areas such as electrical, mechanical, fluid systems and human factors wi11 be evallable to maintain and upgrade safe plant operations through the cognizance and evaluation of applicable operating experience.

'e, 853 288

-,m . . . . , . , . . _,,,..,......__..._.._._.-...._.._____-.-..c....,.u...._.,mn . . . - . . . . . -

Sh. 27 of 33 }

NRR Lessons Learned Task Force Short-Term Recommendations - '

TITLE: Shift and Relief Turnover Pro..Jures (Section 2.2.1.c) ..

3. POSITION The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:
1. A' checklist shall be provided for the encoming and offgoing control room operators and the oncoming shift supervisor to complete and ,

sign. The following items, as a minimum, shall be included in the checklist:

a. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the' checklist).
b. Assurance of the availability and proper alignment of all '

systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable status shall be

. included on the checklist);

c. Identification,ef systems and components that are in a degraded mode of operation permitted by the Technical Specifications.

For such systems and components, the length of time in the degraded ende shall be compared with the Technical Specifications action stateme.;t (this shall be recorded as a separate entry on the checklist).

2. Checklists or logs shall be provided for completion by the offgoing -

and oncoming auxiliary operators and technicians. Such checklists er logs shall include any equipment under saintenance of test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accider,ts or initiate an operational transients (what to check and criteria for acreptable status shall be included on the caccklist); and

3. A system shall be established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent

.. verification of system alignments).

853 289

Sh. 28 of 33 (Section 2.2.1.c)

RESPONSE

Plant procedures for shift and relief turnover will include the foll$ wing:

1. A checklist will be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to

- complete end sign. The following 1 tens will be included in the checklist:

a. Assurance that critical plant parameters are within allowable limits (paraneters and allowable limits will be listed on the checklist);
b. Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable ' status will be included on the checklist);
c. Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications. For such systems and components, the length of tine in the degraded mode will be compared with the Technical Specification action state-ment (this will be recorded as a separate entry of the checklist).
2. Checklists or logs will be provided for completion by the offgoing and oncoming auxiliary operators and technicians. Such checklists or logs will include any equipment under maintenance or test that by them-selves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate operational transients (what to check and criteria for acceptable status will be included on the checklist); and
3. A system will be established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent verification of system alignments).

853 290

e ,

. Sh. 29 of 33 NRR Ler;ons Learned Task Force Short-Term Recommendations TITLE: Control Room Access (Stetion 2.2.2.a) _,

~

3. POSITION The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor, shift supervisor, and control room operators),

to technical advisors who may be requested or ren.uired to support the operation, and to predesignated NRC personnel. Provisions shall include the following:

J.

Sevelop and in:plement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access.

2. Develop and ieplement. procedures that establish a clear line of authority and responsibility in the control room in the event of an emergency. The line of succession for the persen in charge of the control room shall be established and limited to persons possessing a current senior reactor operator's. license. The plan shall clearly

' define the lines of communicatien and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room. '

RESPONSE

Provisions will be made for limiting access to the control room to those individuals respcnsible and necessary for the direct operation of Pilgrim Unit 2, including the Chief Operating Engineer, iluclear Watch Engineer, Operations Supervisor, control room opurators, and technical advisors or other specialists who may be requested or required to support the operation, and to predesignated NRC personnel, Provisions will include the following:

1.

An administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access; and 2.

Procedures that establish a clear line of authority and responsi-bility in the control room in the event of an energency. The line of succession for tre person in charge of the control room will be

  • established and limited to persons possessing a current senior reacter operator's license. The plan will clearly define the lines of communice-tion and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the conirol room.

~,

853 291 s s

.s .

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. Sh. 30 of 33 NRR Lessons Learned Task Force Short-Term Recommendations TITLE: Onsite Technical Support Center (Section 2.2.2.b) _ _ . ,

w

3. POSITION -

Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The tenter shall be habitable t.o the same degree as the control room for postulated accident conditions.

The licensee shall revise his emergency plans as necessary to incorporate the role and location of the technical support center.

A complete set of as-built drawings and other records, as described in -

ANSI M45.2.9-1974, shall be properly stored and filed at the site and accessible to the technical support center under emergency conditions. These documents shall include, but not be limited to, general arrangement: drawings, P& ids, piping system isometrics, electrical schematics, and photographs of components installed without layout specifications (e.g., field run piping and instrument tubing).

RESPONSE

An onsite technical support center for Pilgrim Unit 2 wil'l be provided, separate from and in close proximity to the control room that has the capability to cisplay and' transmit plant status to those individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center will be habitable to the same degree as the controi room for postulated accident conditions to' enable the on-site technical support function to be accomplished. Emergency plans will describe the role and location of the technical support center.

Records that pertain to the as-built conditions and layout of structures, systems and components will be stored and filed at the site and accessible to the technical support center under emergency conditions. Examples of such records include system descriptions, general arrangement drawings, piping and instrument diagrams, piping system isometrics, electrical schematics, wire and cable lists, and single line electrical diagrams, it is not the Intent that all records described in ANSI N45.2.9-1974 be stored and filed at the site and accessible to the technical support center under emergency conditions; however, storage systems will provide for accurate retrieval of all pertinent information without undue delay."

(1) This response corresponds to the revised position in attachment #5 to the-memorandum from Mr. H. R. Denton to the NRC Comnissioners, dated Aug. 20, 1979 853. 292-

- Sh. 31 of 33

~ 'NRR Lessons Learned Task Force Short-Term Recomendations

-~

TITLE: Onsite Operational Support center (section 2.2.2.c) -

3. POSITION An area' to be designated as the onsite operational support center shall be established. It shall be separate from the control room and shall be the place to which the operations support personnel will report in an emergency situation. Communications with the control room shall be provided. The emergency plan shall be revised to reflect the existence of the center and to establish the methods and lines of comunication and management.

e R_esconse:

An area to be designated as the onsite operational support crater wi11 be established. It will be separate from the control room and will be the place to which the operations support personnel will report in an emergency situation. Communications with the control robm will be provided. The emergency plan will reflect the existance of the center and will establish the methods and lines of communication and management. .

m 9

.. 1853 293

. . . . . . . . = . . . . - . _ ..

. Sh. 32 of 33 I

NRR Lessons Learned Task Force '

Short-Term Recommendations TITLE: Revised Limiting Conditions for Operation of Nuclear Power Plants  !

_ , Based Upon Safety System Availability (Section 2.2.3)

3. POSITION All NRC. nuclear power plant licensees sha11' provide information to define a limiting operational condition based on a threshold of complete loss of safety function. Identification of a human or operational error that prevents of could prevent the accomplishment of a safety function required by NRC regula-tiens and analyzed in the license application shall require placement of the plant in a hot shutdown condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in a cold shutdown condi-tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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Tise loss of operability of a safety function shall include consideration of -

the necessary instrumentation, controls, emergency electrical power sources, cooling or seal water, lubrication, operating procedures, maintenance procedures, test procedures and operator interface with the system, which must also be capable of perfoming their auxiliary or supporting functions. The limiting conditions for operation shall define the minispum safety functions for modes 1,

.2. 3. 4. and 5 of nperation.

The limiting conditions of operation shall require the following: . .

1. If the plant is critical, restore the safety function (if possible) and place the plant in a hot shutdown condition within .B hours.
2. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, bring the plant to cold shutdown.
3. Determine the cause of the loss of operability of the safety function.

Organizational accountability for i..e k.ss of oparability of the safety system shall be established.

4. Determine corrective actions and measures to prevent recurrence of the specific loss of operability for the ;particular safety function and generally for any safety function.
5. Report the event within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirm by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designee.

5 Prepare and deliver a Special Report to the NRC's Director of Nuclear Reactor Regulation and to the Director of the appropriate regional office of the Dffice of Inspection and Enforcement. The report shall contain the msults of steps 3 and 4, above, along with a basis for allowing the plant to nturn to power operation. The senior corporate executive of the licensee responsible and accountable for safe plant operation shall deliver and discuss the contents of the report in a public meeting with the Office of Nuclear Reactor Regulation and the Office of Inspection and Enforcement at a location to be chosen by the Director of Nuclear Reactor Regulation.

7. A finding of adequacy of the licensee's Special Report by the Director of Nuclear Reactor Regulation will be required before the licensee returns the plant to power.

'- Sh. 33 of 33 (Section 2.2.3)

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RESPONSE

Boston Edison will participate in the rulemaking procedures and will comply with the results of that process as applicable to Pilgrim Unit 2.

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853 295 1

ATTACHMENT #2 Sh. i of 5 PILGRIM UNIT 2 BOSTON EDISON COMPAfJY DDCKET NO. 50-471 The following additional information provides theNRC with the -

necessary assurance that Pilgrim Station Unit 2 will meet the intent of the positions in the memorandum f rom Mr. H. Denton to the NRC Commissioners, " Resumption of Licensing Reviews for Nuclear Power Plants", August 20, 1979. While the commitments to each position are addressed below, the implementation details will be described more f ully in the FSAR.

A 853 29H6

'N

e Sh. 7 of 5

3. Position Consistent with satisfying the requirements set forth in General Design Criterion 13 to provide the capability in the control room to~ ascertain containment conditions during the course of an accident, the -following requirements shall be Irnplemented:

.,1. A continuous Indication of containment pressure shall be

  • provided an the control room. Measurement and indication capability shall include three ticaes the design pressure of the containment for concrete, four times the design pressure for steel, and eninus five psig for all containments. .
2. A continuous indication of hydrogen concentration in the cofr tainment atmcsphere shall be provided in the control room.

Measurement capability shall be provided over the range of 0 to 10% hydrogen concentration under both positive and negative .

, ambient pressure. .

3. A continuous Indication of containment watersievel shall be provided in the control room for all plants. 'A narrow range instrument shall be provided for PWRs and cover the range f rom the bottom to the -top of the containment sump. Also tcr PWRs, a wide range instrunent shall be provided and cover the range '

f rom the bottom of the containment to the elevation equivalent to a 500,000 gallen capacity. For BWRs, a wide range instrument shall be provided and cover the range f rom the bottom to 5 feet above the normal water level of the suppression pool.

- The containment pressure, hydrogen concentration and wide range containment water level measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy, and testability.

The narrow range containment water level measurement Instrumentation shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodically tested.

RESPONSE

Consistent with satisfying the requirements set forth in General Design

. , Criterion 13 to provide the capability in the control room to ascertain

  • containment conditions during the course of an accident, the following requi rements will be imp lemented:

er d

- 853 2?7

' Sh. 3 of 5 s.

1. A continuous indication of containment pressure will be provided in the control room. Measurement and indication capability will include three times the design pressure of the containment and minus five psig, m 2. A continuous indication St hydrogen concentration in T5e con-tainment atmosphere wi l l be provided in the control room.

Measurement capability wi ll be provided over the range of 0 to 10% hydrogen concentration onder both positive and nega-tive ambient pressure.

3. A continuous indication of containment water level will be provided in the control room. A narrow range instrument will be provi.ded and cover the range f rom the bottom to the ,

top of the containment liquid waste collection sumps. A wide range instrument will be provided and cover the range f rom the bottom of the containment to the elevation equivalent to at least a 500,000 gallon capacity.

The containment pressure, hydrogen concentration and wide range containment water level measurements will meet the design and qualification provisions of Regulatory Guide 1.97 (Rev. 1 ), including qualification, redundancy, and feasibility. The narrow range containment water level measurement instrumentation will be qualified to meef the requirements of Regulatory Guide 1.B9 (Rev. 0) and w111 be capable of being periodically tested.

853 298

Sh. 4 of 5 6:

3. Position Each applicant and .llcensee shall install reactor coolant system and reactor vessel head high point vents remotely operated from the control room.

Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A

- to 10 CFR Part 50 General Design Criteria, in particu lar, these vents shall be safety grade, and shall satisfy the single f ailure criterion and the requirements of IEEE-279 in order to ensure a s aw probability of inadvertent actuation.

Each applicant and licensee shall provide the following information con-cerning the design and operation of these high point vents:

1. A description of the construction, location, size, and power supply for the vents along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe.

The results of the analyses should be demonstrated to be acceptable in accordance with the accetpance criteria of 10 CFR 50. 46.

2. Analyses demonstrating that the direct venting of noncondensable gases with perhaps high h/drogen concentrations does not result in violation of combustible gas. concentration limits in contain-ment as described in 10 CFR Part 50.44, Regulatory Guide 1.7 (Rev. 1) and Standard Review Plan Section 6.2.5.
3. Procedural guidelines for the operators' use of the vents. The information aval'able to the operator for initiating or terminating vent usage shall be discussed.

RESPONSE

The Pilgrim Unit 2 design will provide reactor coolant system (pressurizer and reactor vessel head)high point vents remotely operated f rom the control room. The design of the vents will conform to the requirements of Appendix A to 10 CFR Part 50 General Design Criteria. in particu lar, these vents will be safety grade, and will satisfy the single failure criterion and the requirements of IEEE Std. 279-1971 in order to enstre a low probability of inadvertent actuation.

The following information concerning the design and operation of the high point vents wi ll be provided in the FSAR.

1. A description of the construction, location, size, and power supply for the vents along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe.

The results of the analyses will be demonstrated to be acceptable in accordance with the acceptance criteria of 10 CFR 50.46.

853 299

^ Sh. 5 of 5 a

2. Analyses demonstrating that the direct venting of noncondensable gases with perhaps high hydrogen concentrations does not resu lt in violation of combustible gas concentration limits in containment as described in 10 CFR Part 50.44, Regulatory Guide 1.7 (Rev. 1) and Standard Review Plan Section 6.2.5 (Rev. 1). -.
3. Procedural guidelines for the operators' use of the vents. The inf ormation ava . lable to the operator f or initiating or terminating vent usage will be discussed.

300

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