Text

Forwards Response to NUREG-0737,Suppl 1, Requirements for Emergency Response Capability, Per Generic Ltr 82-33. Guidelines Will Be Incorporated Into Procedures to Ensure Overall Integration & Coordination
	ML18130A390
Person / Time
Site:	Surry, North Anna, 05000000
Issue date:	04/15/1983
From:	Stewart W; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Eisenhut D; Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0737, RTR-NUREG-737 006, 6, GL-82-33, NUDOCS 8304210250
	Download: ML18130A390 (107)
	v • d • e

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w. L. STBWABT VICE P:aESIDENT

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VIRGINIA ELECTRiC AND POWER COMPANY RICHMOND, VIRGINIA 23261

. NUCLEAR OPERATIONS April 15, 1983 Mr. Harold R. Denton Office of Nucle~r Reactor Regulation Attn:

Mr. D. G. Eisenhut, Director Serial No. 006 NO/JBL:ogw Docket Nos.

50-280 50-281 50-338 50-339 DPR-32 DPR-37 NPF-4 NPF-7 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:

License Nos~

RESPONSE TO SUPPLEMENT 1 TO NUREG-0737 REQUIREMENTS FOR EMERGENCY RESPONSE CAPABILITY (GENERIC LETTER NO. 82-33)

NORTH ANNA UNITS 1 AND 2 SURRY UNITS 1 AND 2 VEPCO has recently developed a comprehensive Five-Year Integrated Plan for all capital projects at its nuclear power stations to assist management in more effectively utilizing the

fiscal, engineering, construction and human resources of the company. is a description of the plan and presents the integrated schedule of all capital projects for VEPCO' s nuclear power stations during 1983-1987.

The integration of_ the ERC resource requirements with other station projects has, in part, provided the basis for the schedules provided in attachment 2.

Periodic updates of the plan will form the basis for schedule revisions of the Emergency Response Capability (ERG) projects.

The plan methodology incorporates requirements, priorities*

and restraints to optimize the scheduling of projects over a five year period.

The plan is dynamic and is updated over the life of each project.

The impact of new projects that appear during the five year period can be evaluated and appropriate *action taken if pl~n revision is necessitated.

Attacqment 2 furnishes the data* requested in Generic Letter No.

82-83 as clarified during the NRC. regional workshops conducted during January and February 1983.

Where the dates in this letter conflict with previous VEPCO commitments to NUREG-0737 items, this letter takes precedence.

provides viPcO's Integrated Implementation Plan Guidelines for Emergency Response Capability projects.

These guidelines will be incorporated into procedures to ensure overall integration and coordination of the ERC projects.

J\\~* os Attachments cc:

Mr. Leon B. Engle Mr. J. Don Neighbors Mr. J. P. O'Reilly Very truly.yours, f'1

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W. L. Stewart 8304210250 830415 PDR ADOCK 05000280 F

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COMMONWEALTH OF VIRGINIA )

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I The foregoing document was acknowledged before me, in and for the City and Conu:nonwealth aforesaid, today by W. L. Stewart, who is Vice President~Nuclear Operations, of the Virginia Electric and Power Company.

He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

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FIVE-YEAR PLAN 1*

NUCLEAR -OPERATIONS DEPARTMENT ATTACHMENT t

Vepco VIRGINIA ELECTRIC AND POWER. COMPANY. *

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ATTACHMENT 1 INTRODUCTION On February 1, 1983, a special Vepco Committee was established to develop an integrated plan for capital projects for North Anna and Surry.

The Integrated Five-Year Plan Committee was chartered to develop a plan which would help to reduce cost, establish logical completion dates for projects, provide a framework for budgeting and provide a framework for managing human resources.

The Committee included personnel representing the Nuclear Operations, Power Station Engineering and Power Station Construction Departments.

The initial activity of the Integrated Plan Committee involved basic research into the programs used by similar large corporations for developing long-range planning programs.

The Connnittee explored with several consultants the possibilities of support for development of the Vepco Five-Year Plan for Nuclear Operations.

Since the Connnittee had included in its planning and scheduling model the essential elements necessary for an integrated program, Vepco found that it was not necessary _to acquire outside assistance for the project.

As mandated in the charter, Vepco has been developing the methodology that is the foundation of a computer based long-range planning and scheduling model.

The model includes the requirements, priorities and restraints that must be addressed before either an active or anticipated project can be entered into the computer model.

The formats for presenting the output reports and schedules have been developed based on the the needs of management regarding content and substance of the reports.

PLAN METHODOLOGY The principles of the Five-Year Plan are symbolized in the diagram shown as Exhibit A.

At the core or center of the diagram is the computer based planning and scheduling model.

The Project Resource Evaluation and Management Information System (PREMIS) software package was chosen to manage the project data, since it can be used to generate an interacting network for all capital projects for the nuclear plants.

In addition, changes can be made to activities associated with each project and the total network is automatically updated to show impact on other projects.

The following input data and output information derived from the computer is based on specific elements that are unique to the Nuclear Operations Department:

Input --

0 Requirements 0

License or legal commitments Engineering and construction Surveillance and maintenance Training Future projects Priorities Safety Legal ALARA Operations/Maintenance Improvements

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Restraints.

Imposed deadlines Corporate goals NOD goals Budgets (current/forecast)

Outages (forced/routine)

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0 Reports 0

Management Line organizations Schedules Milestone Detailed PLANNING PROCESS ATTACHMENT 1 One of the key elements involved in the planning process was to establish a data base representing the current status of active and anticipated projects.

A Data Survey Questionnaire was devised to obtain information from those individuals most knowledgeable about specific project activities to generate information for the model.

The questionnaire requested data pertinent* to project budget, schedule, commitments, priorities and human resources.

A supplemental guideline titled, Project Priority Classification Matrix (Exhibit B), was* used to assign project priorities.

The classification matrix was utilized to compare the relative importance of one_project with another, such that a systematic meaningful prioritization system could be implemented.

The three primary variables employed in the matrix are the Requirement Classification Category, Priority Group and the Cost/Benefit factor.

The Requirement Classification Category has been designed to* categorize each project according to its importance to safe generation of electrical power at the plants.

Three Priority Groups were established to define whether a project could be deferred and what impact a deferral would have on the plant.

A limited cost/benefit analysis was performed on each project to ascertain the overall benefit, based on cost,. to* Vepco of implementing a given project.

Each three digit number in the matrix represents the relative priority for accomplishment of the project.

The project matrix code numbers were rank-ordered by the computer to establish the hierarchy for activity scheduling of each project.

An example of the* use of. the matrix in the assignment of a code number is the Technical Support Center (TSC) installation project.

The TSC is a non-deferrable, well defined NRC requirement to satisfy the Emergency Response Capabilities issue as specified in Supplement 1 to f

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ATTACHMENT 1 NUREG-0737.

In the opinion of Vepco, the installation of this facility represents a

low cost/benefit to the Company in terms of an overall improvement to the plants.

Therefore, the TSC project was given the matrix co~e number of 038 which means that the project is non-deferrable, a legal requirement weil defined and has minimal benefit to Vepco.

One of the most important aspects of the planning process is the appraisal of -the quality of the data base.

A confidence factor was assigned to each project based on the extent to which budget and schedule data. was available.

For example, project budget:s that were based on fixe~ price contracts or written contractor estimates were assigned high confidence ratings whereas project budgets based only on preliminary data or past experience at another site were assigned low confidence ratings.

Therefore, each project has identified with it a good, fair or poor confidence factor which is based on the guideline in Exhibit C.

A Masterfile Input Data Report (MIDR) was formulated to capture a total work list of all capital active and anticipated projects for North Anna and Surry, (Exhibit D).

The MIDR includes imposed dates, predece~sor activities and resource requirements that are necessary to implement a project.

The design and access to the MIDR provides for the maintenance of the data base as a *living schedule which can be updated whenever there is a significant change to a project.

A regular MIDR updating activity is necessary to provide sufficient schedule definition to gain management confidence to commit* to specific dates for projects.

At the nucleus of the computer model used for the Five-Year Plan is the PREMIS program which'utilizes a critical path method processor for scheduling projects.

Once the. known* durations and resources, logical relationships to other activities in the project and imposed commitment dates were placed in the main frame computer, PREMIS generated a network for each project.

All projects in the MIDR are shown with beginning and ending dates

for each project in the Time Scheduling Project Bar Chart (TSPBC) as shown as a sample in Exhibit E.

After the first network is generated in TSPBC, changes may be made to any activity parameters and the network would. be updated with the project dynamics.

ANALYSIS OF RESULTS The quality of the assumptions and estimates made in the data base are reflected in the confidence factors in the Five-Year Integrated Schedule, (Exhibit F).

Action has been taken to improve the quality of the data base and the procedures for upgrading the individual project estimations throughout the life of the project will further enhance the utility of the plan.

The Administrative and Procedures Manual will provide detailed guidance for the

submission of data required to support the plan.

Items from Supplement 1 to NUREG-0737 have been identified on Exhibit F by a single

asterisk (*) *.

Additionally, the Nuclear Operations Department's most significant projects have been identified by double asterisks(**).

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ATTACHMENT 1 PLAN OF ACTION The Integrated Plan Committee was chartered to develop a long-range computer-based planning program for the Nuclear Operations Department (NOD).

This has been accomplished and one of the major recommendations of the Cdnnliittee involved the establishment of a planning group within the department to coordinate planning and scheduling activities to support the maintenance of the dynamic Five-Year Plan.

The goal of the evolutionary planning process (Exhibit G) will be to attain at least a

+/-15%

margin in the estimates for engineering and construction by the completion of the Final Design stage of most of the projects.

In an effort to achieve the +/-15% estimate goal on present projects by January 1, 1984, interface procedures will be established between the stations, engineering/construction, and the NOD Planning Group.

The procedures will ensure that the Five-Year Plan computer model is updated

~uring and after each development phase of a

project.

The uncertainties surrounding the nuclear power industry dictate that the planning and scheduling model be designed such that changes to requirements, priorities and restraints can be reflected in the model at any time.

Therefore, the planning and scheduling model will be maintained and controlled in a dynamic (living) manner."

SPECPR/ogw/ds5 AttachmE>nt 1 Exhibi t A DYNAMIC R PLAN ING SCHED. I G ODEL PROGRAM Procedur s Data Collection For ms Comput r Model Entry Pion Revision Exhibit B PROJECT PRIORITY CLASSIFICATION MA TRIX REQUIREMENT CLASSIFICATION CATEGORY

1. MAINTAIN SAFE ELECTRICAL PRODUCTION

2. LICENSE COMMITMENT

3. LEGAL REQUIREMENTS (NRC, EPA, OSHA, NFPA) WELL DEFINED

4. ALARA

5. LEGAL REQUIREMENTS (NRC, EPA, SHA, NFPA) NOT WELL DEFINED

6. PLANT MAINTENANCE

7. REQUIRED SURVEILLANCE

8. PLANT EFFICIENCY IMPROVEMENT

9. PLANT SUPPORT IMPROVEMENT 032283

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ATTACHMENT I EXHIBIT C CONFIDENCE FACTOR GUIDELINES INPUT BASED ON ACTIVE PROJECT-MAJORITY OF ACTIVITIES ARE BEING PERFORMED ON A FIXED PRICE BASIS.

ACTIVE PROJECT-ACTIVITIES ARE BEING PERFORMED BY OUTSIDE ORGANIZATIONS UNDER TIME AND EXPENSE CONTRACTS OR BY VEPCO PERSONNEL BASED ON DEFINED SCOPES OF WORK.

ACTIVE OR PROJECTED PROJECT-ESTIMATES BASED ON SAME WORK PERFORMED ON OTHER UNIT AT SAME STATION.

ACTIVE OR PROJECTED PROJECT-ESTIMATES BASED ON SAME WORK PERFORMED AT OTHER STATION.

ACTIVE OR PROJECTED PROJECT-ESTIMATES BASED ON ITEMIZED MATERIAL AND MANPOWER PROJECTS FOR FIRST OF A KIND WORK.

ACTIVE OR PROJECTED PROJECT-ESTIMATES BASED ON SAME WORK PERFORMED AT ANOTHER UTILITY OR ON SIMILAR WORK IF VEPCO.

ACTIVE OR PROJECTED PROJECT-PRELIMINARY ESTIMATE.

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,,,, LIN! THERMAL St.'I 6165 IIEUTRO!f AASOM!NT Mt. RAC 70(,t SPAR! !t!C C!lff:UT'Oll REPLACE FUEi. JCFER ORlvt; "044 SYSTEM IN UNIT 2 REPLACE FUEi. XFER DRlVI "045 SYSTEM TURBIN& IILl'lC ROOP 6987 Ace & STNOPIP 4856 ltEIIUILD Ul LP TUU Kft 4901 MOD HAI!f lnnt otL PIT "904 CONT P!ftSOte1'!1, AIR IVPPLV 41J8S PltES! POl"1112 Ol'CIIADI 681'14 sa BtoUOOlnl 11!eot!IIT m 4CJCJ5 PRESSURE Ul'Mfflla 1.[GENO:

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FIVE-YEAR INTEGRATED SCHEDULE

Nuclear Operations* Department naw**._... -*

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A Power Station "°""' AlllfA u,if I *11d Z

. Pro;eet o.,cn,tioft 7011*

INSTALL M9l IIT1'A5S Ltlffl 702Z AUi IOU.ER COlfT ITS IMDD 70Z5 COM UPCIAD! t!OD 7029 tJU1I lll'!D COND POLISll1tlffl 41:zt Ml) 1'!"'11' VAST! SH AUX F'O PUHi" 4907 OIL S\\mP LVL At.AM 613]

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HORKAJ. TO EHERGEHCY BUS TIE UNIT Z N006 INSTALL BRAffat FLOW ORlF'lCE 489Z rutL 1t1ILD11'C II.oat VAU.

6186 CDC - 17 MODS REDUCI': SOLID T805 RADWASTE VOLUMI

41170 REC CUIDE 1.97 4871 NS DlSawtCI lt'U1'I 41172 NS DHCRAJICI IIJIIITOa ITUD' l,EGEND:

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S* Legol ~inment (rd wen deflnecl H - H,cjtl -Benefil l: ~01nt11111 Safe Electrical Production 6* Plant Mo,ntenon!~c_.,_, __

--=L:_*~L~o_:_w_B_e_n_e_i _______

f!.- License Cornmitnu11nt

ATTACHMENT t EXHIBIT F Power Statian-...

""-llTR ANN A __

FIVE-YEAR INTEGRATED SCHEDULE Paoe-!of.!.-.

u,if l and 2 Projec1 De1cription 4880 F\\I RECIRC Lift! IZAlt PIOD 7001 GUIDE TUl!J! DISPOSAL 4426 REDUN!JAHT STATION AIR COM 4826 l!JORIC ACID COHC!WT S'nJD'f 4885 PG WATER SToRAC! TAfflt 4832 ItlSTAJ..L F\\TEL/FER IILIND FLANGE UNIT INSTALL FUEL XFER BLIND 4835 fl.ANGE UNIT, 4914 TAAINING LAB !QUIPHElfl'

.. RESPIRATOR DECON REPAIR 7014 FACILITY 4886 FP HEAT/SHOk! DETECTORS CHARLESTON EARTffQUAU STD' 7012 SEAL IMPROV!)£1ff PACltAC!

4895 UPGRADE LAIJNDRT PACILITT FLASHDEHIN REGEN 4903 SYS UPGRADt:

4905 REFURBISH OLD IIACMIN! SRO INSTALL SANCAMO DIST 4910 ANALYZER 4911 INSTALL Dl!S!L C!" JIii DOOi L,EGENO:

!!!f!!!i!Y Graue.*

I* Not Oeferable 2-Deferoble (majar n,oocft 3-Deferablt ( minar inipactJ Nuclear Operations Department Dall 4-15-83 t*J -...

Project Cost Estimate~I.2!2°i2!!!Z Dflhl**...

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~qtirt1ment Classification Coftlgf!!J'*

5*Legol ~irernent(not well deflnlcl I* Maintain Safe Electrical Production 6-Plant Maintenance 2-Licens'! Commitment 7-Required Surveillance 3-Leoal Requirement lwall d.tlna.n 11-oi-* I:'";-:---***----**----*

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ATTACHM.F.NT t EXRtBIT F e

PovfW Sfafion, __

llOlffl AIIII_A __

FIVE-YEAR 1.NTEGRATED SCHEDULE Paie~of !._

u,it 1 Am> Z Project Dttcription 7031 ASDP IIID! M'NC! TEMP IND 4986 VCTcl,EVEL CONTROL CIR IT HOD 4916 SEC COM1'1JTR LOSS OP !P A008

!XPAJfSION JOINT LUK D!T!C 4842 R&LlAHi..r. uu ~nu...,DI> runr PWR UNIT 2 6114 R!Actol READ* '1!NT STST!H 6115 VENT ST!ITl:M U11IT Z 7004 APPENDIX l MODIFICATIONS 48271 NRC AUDIT PZR PORV & l,OJI.Jt' ~r.l\\a.

4875 DRAIN UNIT 2 6118 !~§t~" r.AN.Lt ~......,o.m.

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!M!RC Dl!S!L C!II CHANG!

A003 (CVCS)1'1 'ffl PT KTD CHANG!

A004 (RC) 1'1 TO ff l'l'D CIIARGII:

LEGEND:

f!f!!!i!t GrouP._-

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D 1983 19a4 198!9 1988 1981'

, ~qulrtJmenf Classilicofion Cof~eE!J'*. 5-Legal Re~irement(mt well defined).

I* Maintain Safe Electrical Production 6-Plant Maintenance 2-License Commitmel}t 7-Required Surveillance 3-Legal Requirement (well defined) 8-Plont Efficiency Improvement 4-AL"ARA o_ni--* "**----* *--

Cosl/Bent!fil-H - High--Benefit L* Low Benefit Confidence Focftr-

e ATTACHMENT I EXHIBIT F e

NffllTR ARPA Power Sfation ______ _

FIVE-YEAR INTEGRATED SCHEDULE Paoe..!..ot..?...-

Date ti-n-113 U,it

1 AND 2 IR Project De1cription A005 (Rt!R) NI TO PT RTD CRARCI I\\UU '",,. uIL rur,r A015 THROTTLE VALVE

-A022 AJJU P-IJ 1,;11:-1 LEADltAG CARD A024 NL K.,lR1111r,e,"... uu:.n.

STANDOFFS t..ttl\\Kl,t, t'Ul'II' l(f;C~K\\o A026 vt.V CIRCUIT "1 nu Nrl'R.E::;11 A027 REQ SENSE Lim'!

A030 ~a¥s~iP~fJARATOR ti882 NITROGEN SUPPLY TO PORV MO 4898 RTD BYPASS P'LOll INDICATION 7013 REPLACll! STATION IATl'!1\\1!S 7030 t~El'tNDlCATORL r<.l\\

N045 ~~r~c~Y;¥f.lt XFER 7016 INSERVICE INSPECTION PROC AQ.21 MODin FT-CC*11.5. A, 10 C PIP 7028 Rtt.Ut..Alt CORE COOLUQ M0"1'l'Oll N007 INS LALL PORT IODIN! PILT!II 7023 DIC>ll!X 101' CRltOM 7024 RWST C1l!N ADD TAR PIOl!S LEGEND:

et:iEJj!Y GrouP._-

1-Not O@ferable 2-0eferoble (major lmoactl 3-Deferable (minor Impact)

Nuclear Operations Department

--- ~-***

IN......

Project Cost Estimate < uoo.ooo Df1b1U1..,

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Cosf/B~fil-H - HTg"h-8enefjt L

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f e

A'ITACHMEN'l' t EXHIBIT F Power Stafion_'NOlt_TR_A1'1'_A ___ _

FIVE-YEAR INTEGRATED SCHEDULE

Page-!..of.!-

l)lit 1 ANO 2 IA

-Project Dt1eriptioft 7026 REPLAC! C!N 112 l'URITT MEn A029 COMT StlHP PUKP CHANG!

4993 RP I OUTAC! DtPROVEMEVITS 7010 COlfT !OLT Ttl'!I STRESS NIC 4284 P A CONTAlNH!lff M01'ITOR 4331 t.E. !ULL!TI1' 79-llt 4896 CONTROL RM A/C MOD 5600 NVl'O 1\\1'..un ITT DRAIN S!P T055 CONTAINMENT 1!0Rlt T!NP POWE Rt.lJlJNDANl :.uu£vn 4426 AIR COMPRESS 4830 ADD SECURITY TURNSTILE DWC UPDATE 4887 AN?ru!fClATOR MODS A002 EM!RC Dl!S!L C!1' CIWIG!

A006 RAD MOR PUMP 152 I 259 MO[

AOII TUttB D!Clt IPRnll LVL ALARt A013 PRC CA11Ilf!T lil!UP/IUID SP AOl4 3ft. 5 ff BHll llLT IIULff M LEGEND:

Priority Groul!_-

1-Not Oeferable 2-Deferoble (major,npoctl 3-0eferable ( miner impact)

Nuclear Operations Department Date 4-IS-sJ Project Cost Estimate< 9100*000 al:blu._

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s-Legol Requirement(not wen defined)

I-Maintain Safe Electrical Production 6-Plant Maintenance 2* License Commitment 7-Required Surveillance 3-Legal Requirement (well defined) 8-Plont Efficif!ncv lmnrovtHNtnt A-Al,.,",.

Cosf/B~fil-H - H,gh-Benefit L-Low Benefit Confidence Faetar-

ATTACHMENT I EXHIBIT F Power Station ___

lfOR._m_A_n_A __

FIVE-YEAR* INTEGRATED SCHEDULE Paae I of L-Date 4:.r.5-lr9 Urif.

1 AND 2 Project Olscription A016 ~~~s~1~wl'~nim, A009 ABV GRD OIL STO T.Aft~

FOAM lNJ A012 RM-P-1159 A lit II ELEC SEPAAATIOlf A017 LO TC f!INijN PROTEC 10 REQ'D A018 FIFTH PT DAAtN COOL VLV REP TURB TRIP TO LP HIK AOI9 DRAIN PUMP 7017 ROTATING tQUIP!4!11'r AMALYSI ADD CO ANALYZER A007 eom.ED AIR A025 INSTALL TI 6 VLV TO BS OF IIJR AOOI J:A ~Ny;9Pf0h9 A020 INSTALL RCS PRESS RECORDER N015 INSTALL PE11M EP AT TURBINE NOl7 INSTALL Tf.MP ON R

J MCCS our l N018 INSTALL TEHP ON R I J MCCS UIUT 2 4882 INSTALL LIOUID NITltOC!lf SY9T!N 4894 MULTI FREQ I !DDT CUltR!tff !QPT 1906 VElfT POR CHEN ADD TAJfl9 LEGEND:

f!/!!!i!Y GrauP..-

1-Not O,ferablt 2-Deferoblt (major lmooctl 3-Deferable ( miner Impact)

Nuclear* Operations Department

< $100 000 QllWlall..-

Pro1ect Cost Estimate UH._........,..

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19 8 3 1984 1985 1989 1987

, R~uir~menl Classification Colt1gf!!Y*

S*Legal Requirement(not wen deffnedt 1-Maintain Safe Electrical Production 6-Plant Maintenance 2* License Commitmeqt 7-Required Surveillance 3-Legal Requirement lwtll defined) 8-Plant Efficiency Improvement A-Al ADA Cosf/B~fil*

H - H1ghJ3enefit L

Low Benefit Confid_,~_tl Facttr-

\\

\\'I 1

'I'*

IIOft'ffl AMNA.

Power Station ______ _

u,if l AND 2 ATTACHMENT 1 EXHIBIT F FIVE-YEAR INTEGRATED SCHEDULE e

PCIQl..!..of.!-.

Date 4-u-eJ Project Dt1cription ftilW!III'-

Nuclear Operations Department Project Cost Estimate< 9100 *000 Ill 7018 PREMS UPCRA1>!

M024 INSTALL TO COKPUTE1l GEN RTD OUTSIDE CONiAlrtMt.N.l 4825 IA COHP 4900 UPGrYE P*lJU vu<"lruu,"

UNIT UPCR.AD! P-250 COMPUTER.

4908 IINlT 2 GEN BRKR u,n **-* *- *-

6083 UNIT 2 CONTAIN """"'"" ~-,..

7015 FAC 7021 CONTROL "OOI( ll£HODELl1'G 0084 MISC. STAT. PROJECTS 0085 MISC. STAT. PROJ!CTS 0086 KlSC. STAT. PltoJECTS 0087 Ktsr.. STAT-PRO.J!CTS n9nt ***--.

LEGEND:

Prion*ty Graue*

I-Not Deferable 2-0eferable (mojar lmoact) 3-Deferable lminor lnipactJ PSEC 3

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- *:s 1984 1985

, ~q_lir~manl Classificalion CaltJ~y-

5-Legol Re(Jlirement(not wen defined 1-Maintairl Safe Electrical Production 6-Plant Main1enonce 2-License Commitme11t 7-Required Surveillance 3-Leoat Requirement,well defined)

&-Plant Efficiency imoroveffllltnt 4-ALARA

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D 1999 1911Y Cosl/8~fil*

H -H,gh -8enefit L'" Lowt Benefit Canfid11nr.lf F'~,,,,...

e**

ERC Items

- Vepco Most Significant Issues Power Statian-~SUR="'.:----

Uiit I and 2 Project Oltcriptioft "6097 Spent l'uel Btor lnatant PS!C CDC-17 Reanaly*la PS!

17002 tncore Replacellll!nt PS~

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p A'lTACHMENT t EXHIBIT P FIVE-YEAR INTEGRATED SCHEDULE PCIQIJ-of.!-

Nuclear Operations Department Date 4*15-e3 Project Cost Estimate:?9Joo,ooo I

f-1-L.

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I 1~!0201 Spent Puel Randll111 Mod*

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6124 Cond Retubln1 70]2 s[r-!D Trlp Td*ea

§.920_ 0t-1vlng UP!_1tee 4983 1/it~;~m~P l*pl 6072 I! Bulleti-n 79-14

608li TSC PacUhj 6109 React VH,Lnel Indication 613] cnc-U Mod*

6142 Poat Accident Sa1111l* S1*

6144 Cont RI R*n1* IIAD Notlltor

'614!1 Cont Acc. NCY. n 6155 l'!ffYlron. Qud. Vtllt I 6156 !ffYlron. Qual. Vnlt 2

" 6164 Control Roc,ii Ra*lt, LEGEND:

E!f!!!j!y Grou~ -

I* Not Oeferoble 2-0eferoble (major ~tl 3-Deferoble minor Impact}

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5-Legol Rt!1')irement(mt wen deflnecl I-Maintain Safe Electrical Production 6-Plant Maintenance 2-License Commitmer]t 7-Required Surveillance 3-Legal Requirement (well defined) 8-Plont Efficiency lmproYeanent 4-ALARA I

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Low Benef,t Cmfid,mc, Facftr-

AffAcmmff l DlllBl'I f

. I Power Stalion,-... 8-""...

tt...

FIVE-YEAR INTEGRATED SCHEDULE Ult I and 2 Pro;ect Detcriptm I *

4805 Leed !OP Pacll lty 692) 614) 49]% -

61511 6158' Ttl62 OtJ2 410]

7040 Tl79 T95J 1'8~

ISi Mcmlt. !pulp. teatall Meat Tracln1 lyatea Pc:lff I Safety **tn Ible.

I IC IIIJll'-0 0696 5'0! lnatlll.

111eore 'ftte.-co.a,te lepl.

"' ffl'lt l>ral11,-, Notn SF.I( I UtST A TR F.-f'dvat.-r Ht§ R.-plac""""t *6 Pte Ad'"1nutrat1on Buitdin* Additlo11 Llquld ladW91ta 111-oe.

Solid ladva1t* Toi.....

~ 4967 Sta, fl!r1onnf'I Fac1laty Rena**

4955 Pud lllclg. Hoel Vall 10]1 S.-c. Secuflt1 Acee** ftu ld. 1-..

909 Ha an St*~onotJall SuDoa.-t 428, Rad. Count IIOOII Coaputu Unrade I

l,[GENO:

Priority Gr~*

I* Not Oeferoble 2-0eferoble lmofcr =

3-Deferoblt min<< qNIC Nuclear Operations Department Profect Coif Estimate-= !*oo;ooo PS!C I,

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ATTAcmtENT I EXHIBIT P e

Fower Stafion_mm_n ____ _

FIVE-YEAR INTEGRATED SCHEDULE Paoe..!..ot!_

Date 1t-15-s3

.M Uiit J *ad Z Project Dncriptioft S. G, Prim, H*n T166 Han, D"'ice T164 chg, Pmp. cool SW Strainer T85l Feedvat*r Hll!acer Repl 1 ' 4 Ptl T860 Hl 1.t'VU lf,;

Screen Rep 0393 Fuel TranaFer S,a Node 4950 Plant c°l'Jr,:te~

(nterl111 11ra II Auto Control ror 0296 B*na**

4977 Core Upratlna Node

.-,1<'"7039 Sec, Plllnt Pert.

Ima, HSPH T863 Spare RCP' Ntr. Aueiably 0210 Regenereti*e Reat !*change 4288 Liquid Nitrog@l'I Storage 49l7 Contnlnated Laundry !qul1 4968 Tranefotwer Pir* Protect!,

0392 RCP Quick Dhccmnecte Tl57 Replace SST with Aato tTC T857 lnet. of 4th Paer1 Dleeel T858 lnet of 4th Ill THae.

LEGEND:

Prlon*ty Grouf!_-

1* Not Oeferable 2-Deferable (major fl'GOCft 3-Deferable ( minor impact)

Nuclear Operations Department Project Cost Estimate~ ' 100* 000 n.--*,,...

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II" 1993 19114 199!:S 1998 19117

, Rt1quirt1mtMI Classification Caft1{l!!!Y* * !s"- Legal Requirement(not wen deffnec9 I* Mointaio Safe Electrical Production 6-Plont Maintenance 2* License Commitmel')t 7-Re~ired Surveillance 3-Legal Requirement,well defined) 8-Plont Efflr.iP.nev lmnrnv,iww111,t A-Al ""*

~/B~fil-H - H,gti-Benefjt L* Low Benefit Cmfidtmt:tl Fadtr-

ATTACHMENT I EXHIBIT F e

Power Station:_.,...m...,..1t...,t __ _

FIVE-YEAR INTEGRATED SCHEDULE lnt 1 11nd 2 Ill Project Dncriptioft Ch11rle1ton Earthquake Stdy Conw,run ic at ion a 4819 Audibilih 4926 Hain Ste11111 Dl1c"9Tae Plu.

Imp. R11ni;:e Hain 4927 Stm Di8h Hon 4930 Hech Sa[e Rel Equ i Env. Oual.

49JJ Seiamic & Dyn11n11a Qual Elec

He 4935 RCS Support 4975 R* Ve11el Th@nllll ll'hoelt 7035 Si111iJlltor Conip. 1Jpp1de rrH,5 Spaye RCP Oil Col ectiori Sy1te1 4802 Containment Per,onne1 Hatch Hodific11tion 4816 Spare RffR Puiftp and Notcw 4'944 Replace 12/6" P1thon lffllbb 4855 Aux Bolter s.,, Noda

~ T181 Tr11ining Center Exp11naion, Ph111e 2

I 4818 Training Center E,q,aneloa r1'<MZl Warehou1e !xpanaloa 494] P-2,1~, COIIIP':lt@r Rep aceaent - Unit I LEGEND:

Priorify Gr~-

1-Not Oeferable 2-Deferoble (majcr hDactl 3-Deferablt l minar lnipacQ

Nuclear Operations* Department 1*A *-

Proiect Cost Estimate ~,100;000 ftab.1111...

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191111 19111' Co~/Q~fif-

~tiremtJnl Classificofion Caft1gf!!Y*. 5:.Legal ~irement(mt weft deflillcl

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4-ALARA

~

e ATTACHMENT l EXHIBIT F

.. e Power Station ___

SU'-_H __

FIVE-YEAR INTEGRATED SCHEDULE Paoe.J-ot !..-

Date 4-15-s3 Ulit

! and 2 Pro;lcf Dlseriptioft 4949 P-Z:,O C0111puter ReDlacei11ent untt.2 UIIU Oaf.dntlff.ed ProJeet1 Alto LEGENO:

f!i!!!j]Y Gr~ -

I* Not Deferob1e 2-Deferable (major lmnoctl 3-Deferablt ( minor hnpactJ

- - ~--- -

a.lbll11t...

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5-Legal Requirement(not well deflnedt I* Maintain Safe Electrical Production 6-Plant Maintenance 2-License Commitmef}t 7-Required Surveillance 3'" Legal Requirement (well defined) 8-Ptont Efficiency lmprovemen9

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Confidm,c1 F'acltr-I I*

e e

AffAtiiM!lft I 1111111, r e

,-.. Station,..... mll,IW,l;lu._ __ _

FIVE-YEAR INTEGRATED SCHEDULE

,..!..et..!-

\\)it.

I a11d 2

~---

PN,jacl Dltcriptiull 6116 IHCt IIO I PIE Ynt fl 6111 beet IIO I PH 9nt fl fllU SV *** Motlt..., IIDtM

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- 11 f:l*ctric,1 P*n*tr*taoft leetecllleltt

° ASST PruurJ C*ble R*plac...... t

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ln*~ct1011 5Ht-

942

"-*P Svlt\\h,:.'DI*

Lov Lf'Ye lnuk*

- ,940 Chewl*tr? l**t,.,....tetloa

'991 ApPf'nd i* ll rue ProtectlOII 4167 Auto lstr*c* lta * **

6108 llraclo( v,.. ~li Ln*

11d a ca t OIi fl

,on h1 C:.l'e l.'1 *

,6135 MET Moftlt t1,.r.. e 49]8 llf'pl. \\,n*ra*

Pu*p Dl*ch Yet. I 41107 lnt*r** 1.0n1er, 1100.

li-l*tor U,srede l.£GENQ:

Priority Gr!:!11!*

I* Not Oeteroble 2*Deteroble 1~jcr hDacfl

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... e

~,,,..,,,.,,, Clo~slficoliOfl Colt1f!!!Y* * !*legal ~(,mt WIii detrnld t

~ointoin Safe Electrical Productiaft 6*Plont.M(!l"Je~~*

License Commitment IA~* n

~l/llt!Mft'I*

r=~-1::m

ATTACHMENT I EXHIBIT F

~.

Power Station _ _.SUR;.;,;,;;,;R_T __

FIVE-YEAR INTEGRATED SCHEDULE Pclol..?...of...!-

Dat,

-n-113 lliit~~.;;.*...;a;..n_d_2;;.... __ _

Project Dt1cription 6119 Early 1'arning Sy1tea 0001 Boric Acid Stor, Tk Level Tran, DOOS RCP Stator RTD Rewire D010 Aux Sbll Tan\\ Vent M1MI.

fuel Tran11T

~947 Tube lllind Flg,

erv. Air D007 Contoreeeor Tie-la 0008 -SIG lllovdovn T3nk Rto,noval

-SIG Jrlowdovn D009 Sample Sy11t"' Hod, D004 Liquid Ws11t*

Punip Croa* Connect 4302 Reactor VH'tel ltud Tool*

D012 Et>G Au* Start Pine SU"""rt 0211 Rad Va1te Modification 4815 S/C LeveJ R~ference Le2 In11u 11tion 7036 Dieael Sequenct111 4111S Cont S~JF Relate !qulpaent 41169 Upfrade Low Level In ake Screen 4941 Major Roof Replecl!a!ftt 4960 Vain lurface Grlabr l,EGEND:

Prion*ty Gr~-

1* Not Oeferoble 2.-0eferoble \\major haoctl 3-Deferoble minor lnipai:Q Nuclear Operations Department Project Cost Estimate s1100 1600 COIIII 2

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1.

SAFETY PARAMETER DISPLAY SYSTEM (SPDS)

NRC Requirements "a. The SPDS should provide a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determining the safety status of the plant.

Although the SPDS will be operated during normal operations as well as during abnormal conditions, the principal purpose and function of the SPDS is to aid the control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to avoid a degraded core.

This can be particularly important during anticipated transients and the initial phase of an accident.

"b. Each operating reactor shall be provided with a Safety Parameter Display System that is located convenient to the control room operators.

This system will continuously display information from which the plant safety status can be readily and reliably assessed by control room personnel who are responsible for the avoidance of degraded and damaged core events.

"c. The control room instrumentation required (see General Design Criteria 13 and 19 of Appendix A to 10 CFR 50) provides the operators with the information necessary for safe reactor operation under normal, transient, and accident conditions.

The SPDS is used in addition to the basic components and serves to aid and augment these components.

Thus, requirements applicable to control room instrumentation are not needed for this augmentation (e.g., GDC 2, 3, 4 in Appendix A; 10 CFR Part 100; single-failure requirements).

The SPDS need not meet requirements of the singlefailure criteria and it need not be qualified to meet Class IE requirements.

The SPDS shall be suitably isolated from electrical or electronic interference with equipment and sensors that are in use for safety systems.

The SPDS need not be seismically qualified, and additional seismically qualified indication is not required for the sole purpose of* being a backup for SPDS.

Procedures which describe the timely and correct safety status assessment when the SPDS is and is not available, will be developed by the licensee in parallel with the SPDS.

Furthermore, operators should be trained to respond to accident conditions both with and without the SPDS available.

"d. There is a wide range of useful information that can be provided by various systems.

This information is reflected in such staff documents as NUREG-0696, NUREG-0835, and Regulatory Guide 1. 97.

Prompt implementation of an SPDS can provide an important contri-bution to plant safety.

The selection of specific information that should be provided for a particular plant shall be based on engineering judgement of individual plant licensees, taking into account the importance of prompt implementation.

1e..

Page 2 of 13 "e. The SPDS display shall be designed to incorporate accepted human factors principles so that the displayed information can be readily perceived and comprehended by SPDS users.

"f. The minimum information to be provided shall be sufficient to provide information to plant operators about:

(i)

Reactivity control (ii)

Reactor core cooling and heat removal from the primary system (iii) Reactor coolant system integrity (iv)

Radioactivity control (v)

Containment conditions The specific parameters to be displayed shall be determined by the licensee."

Vepco Response

a. Status of SPDS In January, 1981, Vepco initiated an engineering project to provide an SPDS.

In order to provide a reliable Data Acquisition System (DAS), it was decided that an Intelligent Remote Multiplex System (IRMS) and an Emergency Response Computer System (ERCS) would be used.

The Emergency Response Computer System is composed of the Data Communications Processor (DCP) and the Emergency Response Facility Input/Output Processor (ERFI/0).

The IRMS is composed of multiplexers which will receive those inputs required to support the development of the SPDS.

The multiplexers through buffers, submultiplexers and subbuffers, feed the Master receivers.

Two master receivers feed all inputs into the DCP.

Purchase of the IRMS is over 95% complete and engineering required for installation of the multiplexers and the wiring of their inputs is over 85% complete.

Installation of the multiplexers and wiring of the inputs is over 40% complete at Surry and 7% complete at North Anna.

Engineering to install the remainder of the IRMS is continuing.

The ERCS is composed of the DCP and ERFI/0 which completes the SPDS DAS.

The ERFI/0 computers were ordered in June 1982.

The purchase order includes a color graphic development software package and man machine interface (MMI) which will be used to develop the SPDS displays.

To date, all basic software and 47%,of the color graphic and MMI have been delivered.

Page 3 of 13 A request for proposal for the DCP was offered for bids on July 19, 1982 but award of the contract was delayed due to a review of the design of the TSC.

The contract bidding was reopened and on March 10, 1983 new bids were obtained.

The purchase orders for the DCPs was placed in late March 1983.

A review of industry documents related *to SPDS design has been performed and this information is being considered during

development of displays.

Vepco intends to visit several sites where SPDS is operational to review capabilities and displays.

Verification of the SPDS design will be performed using a program which is being developed by a consultant and being reviewed by Vepco.

The task analysis prepared by the Westinghouse Owners Group (WOG), which is basic to the generic emergency response guidelines and the generic technical guidelines, will also be used in the design.

Human factors design criteria will be used in the design of CRT displays and a human factors engineering review of the displays may be performed.

b. SPDS Safety Analysis Based on the scope of work as presently defined, the SPDS safety analysis will be prepared and submitted to the NRC by February 1, 1984.

c. SPDS Operation and Operator Training The installation of the computers which drive the SPDS in their area of the TSC is planned for Fall of 1984.

The computer system and all other portions of the DAS will be tested after installa-tion, and a six month availability analysis of the computers will be performed.

During this period operators will be provided _

training on the system prior to the system becoming operational in the control room.

Because the installation of the SPDS CRTs in the control room will require modifications to the present control room facilities and may require the replacement of the existing computer console, the installation of the SPDS in the control room can only be accomp-lished during a refueling outage.

Since the computers will not complete their availability test until the middle of 1985 and, placing untested displays in the control room could have negative long term impact on the use of SPDS and on operator confidence in the system, Vepco presently plans to have the SPDS installed in the control room at Surry during the first refueling outage of each unit after July 1, 1985, and at North Anna after the refueling in 1986.

.e Page 4 of 13 This date is predicated upon:

(1)

Favorable results from a study currently underway to determine if the vital power buses will require modification to support the additional load of the data acquisition system.

(2)

The results of the SPDS Safety Analysis do not require data for which sensors do not currently exist.

If either the vital bus requires modification or.additional data points are required to support the SPDS, then the above installation date is subject to modification.

d. Vepco does not desire a pre-implementation review of the SPDS.

e. Integrated Plan Guidelines for the SPDS with the other emergency response capability items is presented in attachment 3.

2.

DETAILED CONTROL ROOM DESIGN REVIEW NRC Requirements "a. The objective of the control room design review is to "improve the ability of of nuclear power plant control room operators to prevent accidents or cope with accidents if they occur by improving the information provided to them" (from NUREG-0660, Item I. D.

1).

As a complement to improvements of plant operating staff capabilities in response to transients and other abnormal conditions that will result from implementation of the SPDS and from upgraded emergency operating procedures, this design* review will identify any modifications of control room configurations that would contribute to a significant reduction of risk and enhancement in the safety of operation.

Decisions to modify the control room would include consideration of long-term risk reduction and any potential temporary decline in safety after modifications resulting from the need to relearn maintenance and operating procedures.

This should be carefully reviewed by persons competent in human factors engineering and risk analysis.

"b. Conduct a control room design review to identify human engineering discrepancies.

The review shall consist of:

(i)

The establishment of a qualified multidisciplinary review (ii) team and a review program incorporating accepted human engineering principles.

The use of function and task analysis (that had been used as the basis for developing emergency operating procedures Technical Guidelines and plant specific emergency operating procedures) to identify control room operator tasks and information and control requirements during emergency operations.

This analysis has multiple purposes and should also serve as the basis for developing training and staffing needs and verifying SPDS parameters.

Page 5 of 13 (iii)

A comparison of the display and control requirements with a control room inventory to identify missing displays and controls.

(iv)

A control room survey to identify deviations from accepted human factors principles.

This survey will include, among other things, an assessment of the control room layout, the usefulness of audible and visual alarm systems, the information recording and recall capability, and the control room environment.

"c. Assess which human engineering discrepancies are significant and should be corrected.

Select design improvements that will correct those discrepancies.

Improvements that can be accomplished with an enhancement program (paint~tape-label) should be done promptly.

"d. Verify that each selected design improvement will provide the necessary correction, and can be introduced in the control room without creating any unacceptable human engineering discrepancies because of significant contribution to increased risk, unreviewed safety questions, or situations in which a temporary reduction in safety could occur.

Improvements that are introduced should be coordinated with changes resulting from other improvement programs such as SPDS, operator training, new instrumentation (Reg. Guide 1.97, Rev. 2), and upgraded emergency operating procedures."

Vepco Response

a. Status of Detailed Control Room Design Review (DCRDR)

Vepco has reviewed available documents regarding the Detailed Control Room Design Review (DCRDR) and determined that, until the design of SPDS displays and location of SPDS equipment is specified, the implementation of new emergency operations procedures (EOPs) is complete, and the determination of which, if

any, Reg.

Guide 1.97 related instrumentation needs to be installed, the DCRDR cannot be completed.

Vepco is presently preparing a paint, tape, and label upgrade for Surry and has previously performed such an upgrade at North Anna.

The Surry upgrade program should be completed by March 31, 1984.

Vepco has been actively involved in the INPO NUTACs on CRDR and Emergency Response Capability (ERC) and intends to begin work on the CDCR plan in the fourth quarter of 1983.

b. Date for CRDR Program Plan Vepco plans to submit its CRDR Program Plans by March _1, 1984.

c. Date for Summary Report The scope of the control room design review and hence the schedule for completing the review are impacted by the CRDR Program Plan and the control room modifications required to meet Regulatory Page 6 of 13

.Guide 1.97.

Vepco will complete Regulatory Guide 1.97 studies in December of 1983 and know qt that time the extent of 1.97 modifi-cations required in the control room.

As stated above, the CRDR Program Plans will be submitted by March 1, 1984. The CRDR Program Plan submittal will define the date for completing the CRDR and submitting the Summary Report.

d. Integrated Plan Guidelines for the DCRDR with other ERC efforts is given in attachment 3.

3.

REGULATORY GUIDE 1.97 - APPLICATION TO EMERGENCY RESPONSE FACILITIES NRC Requirements "a. Functional Statement "b.

Regulatory Guide 1.97 provides data to assist control room operators in preventing and mitigating the consequences of reactor accidents.

Control Room Provide measurements and indication of Type A, B, C, D, E variables listed in Regulatory Guide 1. 97 (Rev. 2).

Individual licensees may take exceptions based on plant-specific design features.

BWR incore thermocouples and continuous off site dose monitors are not required pending their further development and consideration as requirements.

It is acceptable to rely on cur-rently installed equipment if it will measure over the range indicated in Regulatory Guide 1.97 (Rev. 2), even if the equipment is presently not environmentally qualified.

Eventually, all the equipment required to monitor the course of an accident would be environmentally qualified in accordance with the pending Commission rule on environmental qualification.

Provide reliable indication of the meteorological variables (wind direction, wind speed, and atmospheric stability) specified in Regulatory Guide 1.97 (Rev. 2) for site meteorology.

No changes in existing meteorological monitoring systems are necessary if they have historically provided reliable indication of these variables that are representative of meteorological conditions in the vicinity (up to about 10 miles) of the plant site.

Information on meteorological conditions for the region in which the site is located shall be available via communication with the National Weather Service.

These requirements supersede the clarification of NUREG-0737, Item III.A.2.2.

"c. Technical Support Center (TSC)

The Type A, B, C, D and E variables that are essential for perfor-mance of TSC functions shall be available in the TSC.

(i)

BWR incore thermocouples and continuous offsite dose monitors are not required pending their further development and consideration as requirements.

Page 7 of 13 (ii)

The indicators and associated circuitry shall be of reliable design but need not meet Class lE, single-failure or seismic qualification requirements.

"d. Emergency Operations Facility (EOF)

(i)

Those primary indicators needed to monitor containment conditions and releases of radioactivity from the plant shall be available in the EOF.

(ii)

The EOF data indications and associated circuitry shall be of reliable design but need not meet Class lE, single-failure or seismic qualification requirements."

Vepco Response

a. Status of Regulatory Guide 1.97 Vepco, through a contractor, is in the process of comparing the instrumentation specified by Reg. Guide 1. 97 to that currently installed at Surry and North Anna.

b. Regulatory Guide 1.97 Report Vepco will provide a report to the NRC by January 31, 1984 with the results of a comparison and schedule for upgrades and modi-fications necessitated as a result of the comparison.

c. Vepco plans to satisfy the requirement to provide necessary Reg.

Guide 1.97 variables in the TSC and EOF using the SPDS data base and, if necessary, by adding points to a separate data base within the DAS system to provide needed information.

d. Integrated Plan Guidelines for Regulatory Guide 1. 97 with other ERC efforts is included in attachment 3.

4.

UPGRADE EMERGENCY OPERATING PROCEDURES (EOPs)

NRC Requirements "a. The use of human factored, function oriented, emergency operating procedures will improve human reliability and the ability to mitigate the consequences of a broad range of initiating events and subsequent multiple failures or operator errors, without the need to diagnose specific events.

"b. In accordance with NUREG-0737, Item I.C.l, reanalyze transients and accidents and prepare Technical Guidelines.

These analyses will identify operator tasks, and information and control needs.

The analyses also serve as the basis for integrating upgraded emergency operating procedures and the control room design review and verifying the SPDS design.

"c. Upgrade EOPs to be consistent with Technical Guidelines and an appropriate procedure Writer's Guide.

e Page 8 of 13 "d. Provide appropriate training of operating personnel on the use of upgraded EOPs prior to implementation of the EOPs.

"e. Implement upgraded EOPs".

Vepco Response

a. Emergency Operating Procedures Upgrade Status The emergency operating procedure upgrade at Surry is 20% complete using the WOG Technical Guidelines and Emergency Operating Proce-dures Implementation Assistance (EOPIA) documentation.

Although Surry used Rev. 0 of the WOG Generic Technical Guidelines, the procedures will be compared with Rev. 1 of the WOG Generic Tech-nical Guidelines for possible additional upgrade of the emergency operating procedures.

North Anna will use Rev.

0 of the WOG Guidelines for initial procedure upgrade.

b. Rev. 0 of the WOG Generic Technical Guidelines have b*een submitted to the NRC for review and approval.

c. The Procedures Generation Package will be submitted to the NRC by July 1, 1983.

d. The EOPs will be validated using the validation program included in the July 1, 1983 submittal.

The training program description will also be included in the July 1, 1983 submittal.

Since the development and preparation for implementation is a high priority, Vepco does not plan to delay training for three months at Surry after its July 1, 1983 submittal.

Vepco plans to train its operators on the validated procedures and implement the procedures at Surry by October 1, 1983 and at North Anna by March 1, 1984.

e. Integrated Plan Guidelines for EOPs with other ERC efforts is given in attachment 3.

5.

INTEGRATED TRAINING REQUIREMENT NRC Requirement "The design of the Safety Parameter Display System (SPDS), design of instrument displays based on Regulatory Guide 1.97 guidance, control room design review, development of function oriented emergency operating procedures, and operating staff training should be integrated with respect to the overall enhancement of operator ability to comprehend plant conditions and cope with emergencies."

Vepco Response

/"'"

The integrated training plan will be completed by the first quarter of 1984.

Training will be completed by the first quarter of 1986 except for items, presently undefined, which may extend beyond this date e.g., Control Room design deficiencies remaining to be corrected, Reg. Guide 1.97 modifications not completed, etc.

6.

EMERGENCY RESPONSE FACILITIES Technical Support Center (TSC)

NRC Requirements Page 9 of 13 "a. The TSC is the onsite technical support center for emergency response.

When activated, the TSG is staffed by predesignated technical~

engineering, senior management, and other licensee personnel, and five pre-designated NRG personnel.

During periods of activation, the TSG will operate uninterrupted to provide plant management and technical support to plant operations personnel, and to relieve the reactor operators of peripheral duties and communications not directly related to reactor system manipulations.

The TSG will perform EOF functions for the Alert Emergency class and for the Site Area Emergency class and General Emergency class until the EOF is functional.

"The TSC will be:

"b. Located within the site protected area so as to facilitate necessary interaction with control room, OSG, EOF and other personnel involved with the emergency.

II C

Sufficient to accommodate and support NRC and licensee predesignated personnel, equipment and documentation in the center.

"d. Structurally built in accordance with the Uniform Building Code.

"e. Environmentally controlled to provide room air temperature, humidity and cleanliness appropriate for personnel and equipment.

"f. Provided with radiological protection and monitoring equipment necessary to assure that radiation exposure to any person working in the TSG would not exceed 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

"g. Provided with reliable voice and data communications with the control room and EOF and reliable voice communications with the

OSG, NRG Operations Centers and state and local operations centers.

"h. Capable of reliable data collection, storage, analysis, display and communication sufficient to determine site and regional status, determine changes in status, forecast status and take appropriate actions.

The following variables shall be available in the TSG:

(i) the variables in the appropriate Table 1 or 2 of Regulatory Guide 1.97 (Rev. 2) that are essential for performance of TSG functions; and

j**

i9*

e (ii)

Page 10 of 13 the meteorological variables in Regulatory Guide 1.97 (Rev.

2) for site vicinity and National Weather Service data available by voice communication for the region in which the plant is located.

Principally those data must be available that would enable evaluating incident sequence, determining mitigating actions, evaluating damages and determining plant status during recovery operations.

"i. Provided with accurate, complete and current plant records (drawings, schematic diagrams, etc.) essential for evaluation of the plant under accident conditions.

"j. Staffed by sufficient technical, engineering, and senior designated licensee officials to provide needed support, and be fully operational within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after activation.

"k. Designed taking into account good human factors engineering principles."

Operational Support Center (OSC)

NRC Requirements "a. When activated, the OSC will be the onsite area separate from the control room where predesignated operations support personnel will assemble.

A predesignated licensee official shall be responsible for coordinating and assigning the personnel to tasks designated by control room, TSC and EOF personnel.

"The OSC will be:

"b. Located onsite to serve as an assembly point for support personnel and to facilitate performance of support functions and tasks.

"c. Capable of reliable voice communications with the control room, TSC and EOF."

Emergency Operations Facility (EOF)

NRC Requirements II a. The EOF is a licensee controlled and operated facility.

The EOF provides for management of overall licensee emergency response, coordination of radiological and environmental assessment, development of recommendations for public protective actions, and coordination of emergency response activities with Federal, State and local agencies.

When the EOF is activated, it will be staffed by predesignated emergency personnel identified in the emergency plan.

A designated senior licensee official will manage licensee activities in the EOF.

e Page 11 of 13 Facilities shall be provided in the EOF for the acquisition, display and evaluation of radiological and meteorological data and containment conditions necessary to determine protective measures.

These facilities will be used to evaluate the magnitude and effects of actual or potential radio-active releases from the plant and to determine dose projections.

"The EOF will be:

"b. Located and provided with radiation protection features as described in Table 1

(previous guidance approved by the Connnission) and with appropriate radiological monitoring systems.

"c. Sufficient to accommodate and support Federal, State, local and licensee predesignated personnel, equipment and documentation in the EOF.

"d. Structurally built in accordance with the Uniform Building Code.

"e. Environmentally controlled to provide room air temperature, humidity and cleanliness appropriate for personnel and equipment.

"f. Provided with the TSC and facilities to centers.

reliable voice and data communications facilities to control

room, and reliable voice connnunication OSC and to NRC, State and local emergency operations "g. Capable of reliable collection, storage, analysis, display and communication of information on containment conditions, radiological releases and meteorology sufficient to determine site and regional status, determine changes in status, forecast status and take appropriate actions.

Variables from the following categories that are essential to EOF functions shall be available in the EOF:

{i) variables from the appropriate Table 1 or 2 of Regulatory Guide 1.97 (Rev. 2), and (ii) the meteorological variables in Regulatory Guide 1.97 (Rev.

2) for site vicinity and regional data available via communication from the National Weather Service.

"h. Provided with up to date plant records (drawings, schematic diagrams, etc.), procedures, emergency plans and environmental information (such as geophysical data) needed to perform EOF functions.

"i. Staffed using Table 2 (previous guidance approved by the Commission) as a goal.

Reasonable exceptions to goals for the number of additional staff personnel and response times for their arrival should be justified and will be considered by NRC staff.

"j. Provided with industrial security when it is activated to exclude unauthorized personnel and when it is idle to maintain its readiness.

Page 12 of 13 "k. Designed taking into account good human factors engineering principles *."

Vepco Response Technical Support Centers (TSC)

a. TSC Status Engineering of the new Technical Support Center at North Anna and Surry has been in progress since early 1980.

The construction for the facility at Surry was in progress when in April 1982 the project was placed on hold to allow a review of the layout of the operational area.

Due to the generic deadline for completion imposed by NUREG 0737, design and construction had proceeded in parallel until it became apparent that the original design might not support the functional requirements of the facility.

The review, which resulted in some structural changes and in a total redesign of the layout of the operational area, was completed in December, 1982 when a human factors engineering review approved the revised layout.

These modifications will require changes to the design and some reengineering.

The TSC facility at North Anna and Surry have been designed in accordance with the BOCA Code, not the Uniform Building Code.

VEPCO's plans to design the TSCs in accordance with the BOCA Code were outlined in VEPCO's June 1981 submittal, Serial No. 312.

The requirement to design the TSCs in accordance with the Uniform Building Code was not included in NUREG-0696.

VEPCO intends to complete construction of the TSCs, based on current design, as paragraph 3. 7 of Supplement 1 to NUREG-0737 allows for previous work done in good faith.

The Reg. Guide 1.97 variables needed to perform the functions of the TSC will be provided by displays generated using the DAS and the ERGS installed to support the SPDS.

b. The TSCs will be habitable and available for use without full data communication capability during the fall of 1984.

The TSC will become fully operational July 1, 1985 or the first refueling thereafter based on present plans for Surry and after the refueling in 1986 for North Anna.

This can be accomplished provided the results of the review of the TSC function does not require data for which sensors do not exist.

Operational Support Centers The Operational Support Centers presently in use at North Anna and Surry will not be modified with the exception of communications facilities that may be modified when other ERFs are upgraded.

e e

Page 13 of 13 Emergency Operations Facilities The Emergency Operations Facilities presently in use at Vepco proved adequate during previous exercises and Vepco has previously submitted its plans with regards to EOFs.

The plan.,submittals are being revised to provide a hardened local Emergency Operations Facility with protection factors which exceed the NRC proposed requirements and a central Emergency Operations Facility located at the corporate headquarters.

This revision will be.provided for the Commissioners approval.

Within sixty days of receipt of NRC appro-val, VEPCO will provide a projected completion date for the EOFs.

,)

e e

jdm/056$/1 INTEGRATED IMPLEMENTATION PLAN GUIDELINES FOR VEPCO ERC PROJECTS

e Generic Letter 82-33 dated December 22, 1982, requested a description of plans for phased implementation and integration of emergency response activit-ies.

Vepco personnel have participated in the INPO NUTAC for Emergency Response Capability.

The NUTAC has developed a draft "Guideline for an Integrated Implementation Plan."

The draft "Guidelines for an Integrated Implementation Plan" is based on a classical approach with

all activities

. initiated at the same time.

Since Vepco has already initiated emergency response upgrades, the INPO document was used as guidance for developing an l'Integrated Implementation Plan Guidelines for VEPCO ERC Projects" which is included as part

of this attachment.

A summary of Vepco' s plan for inte-gration follows.

SPDS As indicated by the current status given in attachment 2, data acquisi-tion and processing equipment have been procured and basic software develop-ment for the SPDS has begun.

Actual definition of the SPDS CRT displays and selecting a location of the SPDS displays in the control room are currently under way.

The Technical Guidelines and work performed to date on Emergency Operating Procedures will be utilized to define data to be displayed on the SPDS.

A human factors consultant has been actively involved in the SPDS development and has already reported the acceptability of hardware and soft-ware thus far developed for the SPDS.

Input from the human factors consul-tant, the SPDS users and the critical Safety Functions Safety Analysis will be used to complete the design of the SPDS.

The hardware, CRT, installed in the control. room will continuously monitor the five critical safety functions defined in Supplement 1 to NUREG 0737, and at

the same time. have the

.capability to display other plant data for day-to-day use by the control room personnel.

DCRDR The program plan to be developed for the DCRDR will fully develop the integration with other ERC efforts.

The timing of the actual DCRDR will be such that; SPDS display design.is complete and can be verified by the review, modifications to the control room as a result of Reg. Guide 1. 97 can be included in the review, and the generation of EOP' s that will be used with:

these modifications are utilized in the control _room survey.

The intent of the DCRDR review is to verify that all modifications made in the-control room satisfy the objective of improving the operators ability to cope with an accident should one occur.

The schedule for implementing modifications required to address HED' s, if any, will be negotiated with the NRC after submittal of the Summary Report at completion of the DCRDR.

Regulatory Guide 1. 97 Application to Emergency Response Facilities The review conducted in order to prepare the Regulatory Guide 1.97 Report

. will compare instrumentation installed in the plant with that specified by Regulatory Guide 1.97.

When discrepancies are identified, the Technical Guidelines, Emergency Operating Procedures and Plant design features will be reviewed to determine if the missing variables is. required to enhance the operators ability to monitor the course of an accident.

Any additional data that should be displayed in the control room will be compared with the SPDS i

design to determine if the SPDS could satisfy requirements.

Data displayed on the SPDS -

CRT in the control room (as part of the SPDS or a supplemental display utilizing SPDS hardware), or on continuous indicators added to the control board, will be reviewed as part of the DCRDR prior to finalizing the SPDS and display instrumentation location.

Those 1.97 variables necessary to perform TSC and EOF functions will be displayed in these facilities.

EOP's As stated above the EOP's or technical guidelines will provide input to the SPDS design, Reg. Guide 1. 97 review and the DCRDR.

The EOP' s will be developed based on,the Westinghouse Owners Group's Generic Technical Guide~

lines, the task analysis that was inherent in the development of these guide-lines and a writers guide.

EOP training and update programs will be defined in the Procedures Generation Package submittal.

The package will also define the methods used to develop plant specific EOP's from the Generic Technical Guidelines and the verifications and validation process for the procedures.

Procedure validation may be performed as part of the DCRDR or provide input to the DCRDR.

Emergency Response Facilities Upgraded Emergency Response Facilities will be. provided to meet Regu-lations and meet the functional requirements provided in Supplement 1 to NUREG 0737.

Appropriate voice connnunication links between the TSC, OSC, EOF, State and local Governments, the NRC and the National Weather Service will be provided in the upgraded facilities.

The data acquisition and processing system provided for the.SPDS will also collect and process those variables necessary for performance of the TSC and EOF functions.

Human factors princi-ples, review of Emergency Plans and Emergency Plan Implementing Procedures will be utilized to define the data needed in the facilities.

Those Regula-tory Guide 1.97 variables necessary to perform TSC and EOF functions will be displayed in these facilities.

Any new or existing data displayed in the control room that is necessary to pe*rform TSC and EOF functions will be provided in these facilities

ii

e FOREWORD This Integrated Implementation Plan outline for ERC Projects has been developed using the guidance document developed by the INPO RJTAC on ERC.

The basic concepts will be applied to develop a complete integrated plan

-for Vepco projects. It should be noted that some diagrams developed here may require revision to better satisfy integration requirements.

jdm/05658/2

e e

SH9'1ARY The _p~rpose of this document is to outline an integrated implementation plan for Vepco's ERC projects.

Detailed development of the plan will be performed as a part of each ERC project.

The guidance provided in this document addresses the major elements involved in the implementation process and their interactions.

The guidance developed by the INPO NUTAC on ERC has been utilized to develop this document.

The NUTAC Guidance has been applied as it appropriately relates to the current status of each Vepco project.

For simplicity of presentation, the integrated implementation plan has been divided into the following sections:

0 Emergency Operating Procedures (EOPs) 0 Control Room Design Review (CRDR) 0 Regulatory Guide 1.97 (R.G. 1.97) 0 Safety Parameter Display System (SPDS)

0 Emergency Response Facilities *(ERF) 0 Implementation Each section describes the elements involved and how these elements interface.

More. importantly each section addresses the interactions between the EOPs, CRDR, R.G. 1.97, SPDS, and ERF that should be considered to ensure proper integration of all ERC elements.

jdm/0 56 SB/3

{.

e 1.1 SCOPE SECTION 1 INTROOO CT ION This guidance identifies and describes the major elements and interfaces that will be considered when preparing a plan and schedule for the implementation of the integrated Emergency Response Capabilities provisions (NUREG-0737, Supplement 1).

1.2 BACKGROUND

Realizing the impact of Supplement 1 to NtJREG-0737, an industry review group recognized the need for an ad hoc committee to provide guidance to the industry in implementing the provisions of Supplement 1 to NtJREG-0737.

Therefore, an industry meeting was held September 26, 1982 to determine whether there was sufficient utility interest and a recognized need to pursue an industry program to develop such guidance.

The INPO NUTAC on ERC was formed and guidance documents prepared by

the NU TAC are utilized here to develop guidelines for a Vepco intergrated plan.

jdm/056 5B/4

1.3 OVERVIEW The NRC document (Supplement 1 to tlJREG-0737) defines various provisions that are directed toward enhancing the control room operator's ability to deal with emergency conditions.

These provisions include installing a safety parameter display sys*tem, conducting a

control room design

review, upgrading plant-specific emergency operating procedures, implementing R.G.

1.97 instrumentation provisions, and providing emergency response facilities.

As specified in Supplement

1 to WREG-0737, coordination* and integration must occur during the development and implementation of these provisions.

This document provides guidance to assist in the development of an integrated implementation program.

The method developed by the M.JTAC for the integration of the provisions of Supplement 1 to NUREG-0737 is shown in Figure 1-1.

Figure 1-1 was drawn to illustrate the five major provisions (EOPs, CRDR, R.G. 1.97, SPDS, ERF) and the basic elements and interfaces for each provision that should be considered in the development of an integrated implementation plan.

Each element and its relationship to other elements is discussed in the following sections.

jdm/056SB/5

Step 0 Plan tntttattan Step I Onelop Input Crtterta Step 2 lntttal Eva hiat1an Step J.

Deterwt111tt1111 Step 4 lntegrat1c,n Step I Yertftca1 tall Stt1t 7 Revtew Step 8 111111 lerientatton Step 9 Yaltdatfc,n Step 10 Cal!lp 1 et ton THII Anal,sh e

Nureg-0737 Supplement 1

.-@~~1:f:~

I RG l,97 Pla

RG 1,97 Dest gal Crlterta SPDS Plan Critical Safety Functt11111 SPOS Dest911 Bases (lntttal)

Completed lqilementatton Figure 1-1 Integration method for supplement 1 to NUREG 0737 provisions

EOP Element The Emergency Operating Procedure (EOP)

Plan consists of tasks that will prq_vi_de a

documented method for developing, utilizing,

revising, and controlling EOPs.

The product will provide a framework that enhances an operator's ability to avoid'degraded conditions.

These tasks should include defining source documents, determining manpower requirements, establishing a schedule, and specifying a method of document control.

The plan should also define the interfaces with other elements of Supplement 1 to NUREG-0737 to ensure complete integration.

Initial plant-specific EOPs are developed for the purpose of mitigating the consequences of a broad range of initiating events and subsequent multiple e

failures, without the need to diagnose a specific event.

These procedures are function-oriented and include human factors considerations that enhance human reliability.

These initial EOPs are developed based upon a writer's guide and NSSS vendor generic technical guidances.

NSSS generic technical guidance was prepared based on re-analysis of transients and accidents and included generic task analysis.

EOPs should be checked for completeness, understandability, technical accuracy, usability, and compatibility with the control room design.

In order for operators to have confidence in the EOPs, all these criteria must be met.

A walk-through of the initial EOPs provides a method for evaluating these criteria.

jdm/056~/6 r*.. \\

The EDP walk-through may be performed in the control room, in a *simulator, using a mock-up or thf! control room, or any combination or the three.

Prior

*to making* this

decision, resource availability and advantages versus disadvantages or each available evaluation method should be determined.

Although Figure 1-1 indicates only one EDP evaluation (walk-through), this process should be conducted following any major modifications to the EDPs.

The EDP walk-through will provide input to the DCRDR.

EDPs may* be refined or revised based on the impact of control room human engineering discrepancies (HEDs),

specific application of R.G.

1.97 recommendations and SPDS design bases.

1.3.2 Control Room Design Review (CRDR) Element The CRDR Plan is the first step toward performing a Control Room Design Review and provides methodology for performing the entire review.

The objective of

the Control Room Design Review is to provide an environment that supports and enhances the operator's ability to operate within the EDP framework.

The Control Room Survey is used for identification and documentation of existing equipment.

This task can be done as part of the EOP walk-through.

The Operating Experience Review is performed to identify any operational problems resulting from design inefficiencies or identify any modifications to the control room which would enhance the ability or an operator to respond to an *emergency condition, with con~idera tion given to the importance or safety, impact to plant and personnel, and economics.

jdm/0 56 SB/7

e e

e In performing the CRDR, accepted human ractors principles should be used.

Good human engineering practices should be incorporated in any control room design since the operator must interface with this equipment under abnormal, as -we-11 as normal, conditions.

The Control Room Survey should include EOP walk-through or utilize results from the EOP walk-through, and* opera ting experience data, as well

as human engineering criteria, to uncover any control room design problems.

This survey should include, among other things, an assessment of control room layout, the coritrol room environment, the usefulness of audible and visual alarms, the readability of displays, the adequacy of instrumentation,* and the information recording and recall capabilities.

This survey is. essential. to ensure that all human engineering discrepancies (HEDs) are found.

The operator's tasks and information requirements are validated by the EOP walk-through and provide input criteria to the control room survey process.

The walk-through may be performed during the CRDR Survey.

Control room additions associated with the SPDS and incorporation of selected R. G. 1. 97 modifications will be considered in the CRDR, along with changes resulting from other programs.

Supplement 1 to NU REG 0737 also states that the CRDR should be used to verify SPDS parameters.

jdm/0 56 SB/8

.e e

1.3.3 R.G. 1.97 Element The R.G. 1.97 plan provides the administrative guidance needed to assess and docW11ent all aspects of R.G. 1.97 consideration.

The objective of the R.G.

1.97 instrument provisions is.to provide the data and information required by the opera tor.

The data required are mandated by the opera tor's needs for effectively executing the EOP and ERF functions.

A complete set of criteria is developed in the R.G. 1.97 plan to form a basis for instrument selection.

Utilizing the criteria, a plant specific iist of accident monitoring instrumentation, qualification criteria and locations is developed.

The plant list also provides feedback to the control room survey and SPDS design basis.

ERF design criteria may provide additional input to the plant list.

1.3.4 SPDS Element The SPDS plan describes the tasks which will provide a method for developing, revising, assessing and implementing the safety parameter display system design bases, a method for documenting these efforts, and a method for implementing an SPDS.

The objective of the SPDS is to provide a concise set or information* to aid operating personnel in assessing plant safety status.

jdm/05653/9

The plan will provide a description of

each task involved and give the

~*

administrative guidance needed to perform the tasks, including defining source documents, determining manpower requirements, specifying vendor involvement,

es~~~J.ishing a schedule, and specifying a method of configuration control.

Interfaces with other Supplement 1 to R.JREG-0737 elements should be clearly defined to ensure complete integration.

A list of human factors criteria pertaining to the SPDS should be developed as a basis for developing and assessing plant-specific SPDS designs.

This list of criteria may be developed in conjunction, with the human factors criteria required as input for the performance of a control room survey.

The

EOPs, as a

result of the efforts of the NSSS Owners Groups. and plant-specific considerations, specify the critical safety functions for a

, e plant.

The SPDS design bases should incorpor.ate this information so that the operator can use the SPDS, if available, in conjunction with the EOPs.

To ensure an effect! ve SPDS, the design bases must specify hardware, inputs, and

software, identify SPDS user(s),

specify

location, and define availability.

The SPDS design is mandated by operator usability and compatibility with plant-specific EOPs.

Once the SPDS design bases have been determined, the adequacy of the design should be verified.

jdm/0 56 SB/10

i._.

ERF Element The ERF pl~n describes a method for designing, implementing, and utilizing the emergency response facilities.

The purpose of an ERF is to support the operating personnel during degraded conditions.

The criteria that provide a basis for the design or upgrade of the Emergency Operating Facility {ECF), and Operational Support Center {OSC) need to be determined.

The bases for these criteria should include consideration of 10 CRF 9).47, 10 CFR 9::>, Appendix E, WREG-0696, Emergency Plans, and guidance provided by nuclear industry organizations.

1.4 THE ITERATION PROCESS e

The EOP,

CRDR, R.G. 1.97, SPDS and ERF elements should be involved in a iterative process that includes control room enhancements, plant-specific EOPs, specific R.G. 1.97 application, SPDS design, and ERF criteria.

This I

iterative process should continue until all of the ERC criteria have been

considered and satisfied.

e jdm/056$/11

e*

OVERVIEW SECTION 2 CONSIDERATIONS IN THE DEVELOPMENT CF AN INTEGRATED IMPLEMENTATION PROGRAM This section identifies and discus.ses the factors.that should be considered in the development of an integrated plant-specific implementation program.

Each of these factors is discussed in detail in the following sections of this document.

2.1 DEFINITION OF ELEMENTS AND INTERFACES

(-

An effective integrated implementation program requires a clear definition of each element involved and how these elements interface.

The primary elements involved in this program include emergency operating procedures, control room design review, R.G. 1.97, safety parameter disp~ay system, and emergency response facilities.

None of these elements* are totally isolated from one another, as depicted in Figure 2-1.

Because of their interrelationships, it is essential to clearly define the interfaces between the elements.

Section 3 provides guidance in_ defining these elements and interfaces.

jdm/0 56 5B/12

1.e e

~

May Include 1.97 Variabl~

./

/

Provide Data Required By EOPS J:~)

./

Resolution of HEDS May Pro vi de Information to ERF Fig. 2-1 SPDS Interfaces i

e SECTION 3

. DEFINITION CF ELEMENTS AND INTERFACES OVERVIEW An effective integrated implementation program requires a clear definition of each element involved and how these elements interface.

This section defines the elements and interfaces contained in Figure 1-1. In order to provide this information in a usable and logical format, this section has been divided into the following subsections:

o Emergency Operating Procedures *(EOPs) o Control Room Design Review (CRDR) o Reg. Guide 1.97 Provisions (R.G. 1.97) o Safety Parameter Display System (SPDS) o Emergency Response Facilities CERF) o Implementation Figure 3 illustrates which elements and interfaces will be included in each subsection.

jdm/0565B/13

e Ste,0 Plan Jn1t1at1on r Step l DeveloP Input Cr1ter11 I

I Ste, 2 lnit11l Ev1lu1tt011 Ste, 3 Oetel'lllinatton

\\ *-

Step 4

/

Jntegratton

/

/'

neric idellne I

I

\\

I I

I

\\

.,,.,, H1111111n Factors

/

l111Jrovements And U rades

/

CROR Program Plan _J HU11111n Factors Engineering I,_

SPDS Plan* I I I'--&--...

1, Critical Safety 1: Functions I,,

i I

'I I

I

'""s~Po"'s ___ ___,,

Design Bases (Initial)

_1

\\

I I

I,.

I

_J I

Step I f

¥erif1c1tton \\

Step 7 Review Step B l1111lementatton Step 9 Valldatton Step 10 Completion Figure 3-l111J 1 ementat ion ystem Va 11 d~...:.1=:on-'-__,

~eted j~mentatlon I

(

I I

-* I

,re te e

3.1 EMERGENCY OPERATING PROCEDURES (EOPs)

. The, elements and interfaces described in this section are detailed on

. __ _figure 3-1 Line 1

,1, WRITERS GUIDE Block B Line 4 jdm/05658/14 EOP PLAN Block A

,c.

PLANT SPECIFIC EOPS (I~IPAL)

Blol" Line 7

d.

Line 2 I,

TECHNICAL TASK Iino '.:l.

ANALYSIS GUIDELINES I'

Block c Rlnrlc D Line 5

rm,--~

Line 6 WALKTHROUGH J Block F

~,

FIGURE 3;.1 EOP ELEMENTS AND.

INTERFACES

3.1.1 EOP Plan - Block A tlJTAC - Recommendations A* detailed EOP Plan is necessary for the development and implementation of Emergency Operating* Procedures.

The EOP Plan consists or those tasks which will provide a documented method for developing, utilizing, revising, and controlling Emergency Operating Procedures.

The plan will be plant-specific, measurable in terms of deliverables, and consistent with current NRC guidance.

The objective of the EOPs is to provide a framework to enhance the operator's ability to avoid degraded conditions.

References for Emergency Operating Procedures (Section 3.1)

e "Emergency Operating Procedure Implementation Guideline" (INPO 82-016), EOPIA Review Group, June 1982 "Emergency Operating Procedures Generation Package Guideline" (INPO 83-007), EOPIA Review Group, February 1983 "Emergency Operating Procedures Writing Guideline" (INPO 82017), EOPIA Review Group, August 1982 "Emergency Operating Procedures Verification Guideline" (INPO 83-0XX), EOPIA Review Group, 1983 tlJREG-0899, "Guidelines for the Preparation of Emergency Operating e

Procedures", USNRC, August 1982 jdm/056 SB/15

e NJ REG-0660, "NRC Action Plan Developed as a Result of the TMI II Accident",

NS.NRC, May 1980 Plant-Specific Writer's Guide Existing Plant-Specific Emergency Operating Procedures NSSS Technical Guidelines Generic Task Analysis Supplement 1 to NJ REG-0737, "Requirements for Emergency Response Capability",

USNRC, December 1982 Vepco's EOP Plan will be described in the procedures generation package that will be submitted to the NRC.

3.1.2 Writer's Guide - Block B The Writer's Guide should be developed to define specific methods for preparing plant-specific EOPs from technical guidelines.

Vepco will use guidance documents developed by the INPO EOPIA NJTAC in combination with the WOG technical guidelines to develop EOPs.

jdm/0565B/16

e I

3.1.3 Technical Guidelines - Block C The purpose of these guidelines is to provide a technical foundation for the pla!)t:-specific EOPs.

These guidelines identify operator tasks, information

needs, and control functions required for emergency operating conditions.

Vepco will use Technical Guidelines prepared by the WOG.

3.1.4 Task Analysis - Block D The Task Analysis consists of an analysis. of t*ransients and accidents to identify operator tasks and information and control needs to avoid degraded conditions.

This Task Analysis was performed and has been used by the Westinghouse O,mers Group as a basis for the emergency operating procedure Technical Guidelines.

Technical Guidelines can be used as input criteria for the Control Room Design Review and the R.G. 1. 97 1esign.

3.1. 5 Plant Specific EOPs (Initial) - Block E Plant-specific EOPs are developed for the purpose.of mitigating a broad range of initiating events and subsequent multiple failures or operator errors without the need to diagnose a

specific event.

These procedures are function-oriented and written with human factors considerations (provided by the Writer's Guide) to enhance human reliability.

jdm/0 56 58/17

"{,**.

3.1.6 EOP Walk-Through - Block F EOPs should be checked for completeness, understandability, technical co!r~~tness, usability, and compatibility with the control room.

EOP walk-through ls a method used to provide assurance that the procedures are adequate.

The EOP walk-through may be performed as part of the EOP development. process alone or with the CRDR.

Due to the timing of the CRDR it may be necessary to perform a walk-through with both the CRDR and the EOP development process.

The EOP walk-through serves a~ validation of the EOP.

Initial validation. has been completed by the W.O.G.

This is documented in WCAP 10204, Summary Report for Emergency Response Guideline Validation Program *.

3.1.7 EOP Plan/Writer's Guide ~nterface - Line 1 The EOP Plan should describe the development and use of a writer's guide in.

the preparation of the plant-specific EOPs.

This is defined in the Procedures Generation Package.

3.1.8 EOP Plan/Technical Guidelines Interface - Line 2 The EOP Plan should describe the use of the WOG Technical Guidance in the preparation of the plant-specific EOPs.

jdm/0 56 58/18

~-~-~-:---:--------:---------,,.----:-------:-~------~

e I.

I 19

. --* < -~***..,

3.1.9 Task Analysis/Technical Gui9elines Interface - Line 3 The results of the Task Analysis performed by Westinghouse was used by the WOG as-.a- _basis for the Technical Guidelines.

It should be noted that preparation of Technical Guidelines in this manner included a generic Task analysis and a plant specific task analysis is not required.

3.1.10 Writer's Guide/Plant-Specific EOPs (Initial) Interface - Line 4 The Writer's Guide is used in the development of the plant-specific EOPs.

3.1.11 Technical Guidelines/Plant-Specific EOPs (Initial) Interface - Line 5*

The WOG Technical Guidelines are used plant-specific EOPs.

as the basis for preparing* the 3.1.12 Plant-Specific.EOPs (Initial) and EOP Walk-Through Interface - Line 6 Plant-specific EOPs are used to* perform an EOP walk-through.

The results of the EOP walk-through may identify potential modifications as input to the CRDR *

. 3.1.13 EDP/Iteration Interface - Line 7 The initial. generation of* EOPs will be revised as ERC projects provide modifications to the control room i.e., SPDS, 1.97 and potential DCROR modifications.

jdm/0 56 5B/l 9

I 3.2.Control Room Design Review (CRDR)

The elements jnd interfaces described in this section ar~ detailed on

__ F_igure 3-2.

TECH GUIDELINE See Section 3.1 EOP WALK THROUGH I

jdm/05658/20 Line 6 Figure 3-2 CRDR PROGRAM PLAN

)

CRDR PROGRAM PLAN Bl Line 1 E TING EXPERIENCE Block Line 9 HEDS Block E Line 10 iLine 2 MAN FACTORS GINEERING ock C Line 5 I r-::s=p E=c,.........I=-F I--c......,R,,,_.-G-. -

Line i 1. 97 APP. {INT)

See Section 3. 3 I

Line B* SPDS DESIGN BASIS( INITIAL)

I

e CRDR Program Plan - Block A The timing_ of the control room design review is critical to integration with other ERC projects (i.e., SPDS design, EOPs, and Reg.

Guide 1.97).

The overall control room design review should start only after it can take into account all* other ERC design changes that are occurring in the control room.

In addition the impact of these changes should be assessed against the initi~l Emergency Opera ting Procedures such that the generation of EOPs that will be used in the CRDR walk-through includes/account for planned modifications.

Thus, in order for the CRDR to fully support integration with other efforts

the actual review will include existing control room conditions along with planned modifications that have reached final design.

The total assessment will be by design drawings (possible simulator mock-ups) and pre ERC-implementation control room conditions.

A CRDR Program Plan is required for NRC submittal and will define administrative guidance necessary to develop, perform, assess and document all aspects of the CRDR review.

This includes establishing the review program and

schedule, providing criteria for assessment of

HEDs, and methods for correction.

The detailed plan when developed will clearly identify all interfaces with other ERC efforts.

References for Control Room Design Review (Section 3.2)

Supplement 1 to NJREG-0737, "Requirements for Emergency Response Capabilities, US ll!C, December, 1982 jdm/0 56 53/21

"Control Room Design Review :Implementation Guideline", draft

{INPO 83.;.oxx KJTAC), CRDR IIJTAC "C9n~rol Room Design Review Survey Development Guideline", draft h NPO 83.:.oxx 111 TAC) v CRDR IIJTAC "Human*Factors Principles for Control Room Design Review", draft (INPO 83-*oxx.NJTAC), CRDR KJ1'AC "Control Room* Design Review Task *Analysis Guideline", draft

{INPO 83-0XX HJ TAC), CRDR HJTAC "IlPO/TVA Pilot Systems Review" (INPO 82-014), INPO and TVA, June, 1982

e. NJREG-0700., "Guidelines for Control Room Design Review", US me, September, 1981 HJREG-0801, "Evaluation Criteria f'or Detailed Control Room Design Reviews",

us.me (draft for comment)

.: jdm/0565B/22

-* -~*.... -. **-;--..... ~.

3.2.2 Operating Experience - Block B Operating experience assists in identifying any operational problems resulting fr91D_.design discrepancies.

Available operating experience should be used to further ensure that the review is complete.

Various resources are available for the accumulation of operating experience data.

The most important source or data is the operating personnel.

This information may be obtained through operator interviews and reviews of industry experience, including LERs.

3.2.3

Human Factors Engineering - Block C Human factors engineering principles should be used to determine if the design of the control room incorporates good human engineering practices.

Various human engineering principles have been provided by the military, me, and other nuclear industry related organizations.

As part of the detailed program plan, a composite list of those principles pertaining to nuclear power plant control rooms should be developed as a basis for the determination of HEDs.

3.2.4 Control Room Survey - Block D The Control Room Survey is a static verification of the control room performed by comparing the existing instrumentation layout, lighting, and noise levels with accepted human engineering principles.

This survey will utilize EOPs/task

analysis, opera ting experience

data, and human engineering principles to uncover any control room design problems.

This survey should jdm/0 56 $/23

e include, among other things, an assessment of control room layout, the control room environment, and usefulness of audible and visual alarms, the readability of displays, the adequacy of instrumentation, and the information recording and recall capabilities.

3.2.s Human Engineering Discrepancies (HEDs) - Block E HEDs are characteristics of the existing control room that do not comply with accepted human engineering principles.

HEDs must clearly be defined and enough information must be provided to allow some type of prioritization.

This establishes a good basis for assessing the significance of a HED and determining whether some modification is required.

The criteria for prioritizing may include, among other things, impact on plant

safety, impact on plant availability, involvement in previous operating events, economics, and impact of modification on plant operation.

3.2.6 CRDR Program Plan/Operating Experience Interface - Line 1 The CRDR program plan should require the accumulation and evaluation of operating experience data.

CRDR Program Plan/Human Factors Engineering Interface - Line 2 The CRDR program plan should describe the use of existing human factors engineering principles as a basis for determining HEDs.

This will be accomplished in the Program Plan development.

jdm/0565B/24

_ _J

e 3.2.8 Task Analysis (Technical Guidelines)/Control Room Survey - Line 3 The information and control needs identified in the Task Analysis (Technical au;de_lines) (See Section 3.1.4), provide input criteria for the control room survey.

The results of the Task Analysis are compared with the control room instrumenta-tion during the survey process to determine if any enhancement of displays or controls is needed.

3.2.9 Operating Experience/Control Room Survey Interface - Line 4 The results of the operating experience review provide additional information for review in the control room survey.

3.2.10 Human Factors Engineering/Control Room Survey - Line 5 The. human factors engineering principles developed for. nuclear power plant control rooms and criteria developed in the plan provide the basis for identifying HEDs in the control room survey.

3.2.11 EOP Walk-Through/Control Room Survey Interface - Line 6 The qperator's tasks and informational requirements validated by the EOP walk-through (see Section 3.1.6)

provide input criteria to the control room survey process.

This interface is classified as one-way.

The results of an EOP walk-through may provide valuable information to the control room survey, but the survey has no effect on the walk-through.

jdm/056~/25

- -- ~.---~-.. *.... **- *-**

If EDP production has been completed or is performed in parallel with the CRDR e.

program, it may be possible to utilize the EOPs and information resulting from the walk-through instead of task analysis as input to the control room survey.

e 3.2.12 Specific R.G. 1.97 Application/Control Room Survey Interface - Line 7 Enhancements required by R.G. 1.97 provisions (see Section 3. 3. 3) should be considered during the control room survey.

The instrumentation list prepared as part of the Reg. Guide 1.97 study may be used to supplement the control room survey.

The timing of the control room survey and EOP wal k-throughs should be such that 1.97 modifications planned in the control room are a part of the total CRDR process.

3.2.13 SPDS Design/Control Room Survey Interface - Line 8 Enchancements associated with the SPDS (see Section 3.4.4) may be included in the control room survey.

This could minimize control board modifications.

jdm/0 56 $/26

~

. 3.2.14 Control Room Survey/BED Interface - Line 9 The evaluation of the results of the control room survey should provide a list of-.HEDs.

It is important to incorporate final Reg. Guide 1.97 design, SPDS/PDS design and know modifications to EOPs associated with SPDS and Reg.

Guide 1.97 in the CRDR to minimize the BEDs.

3.2.15 BED/Iteration Interface - Line 10 The control room enhancements should be coordinated with changes resulting from other programs, such as EOP, R.G. 1.97, SPDS, and ERF.

The iteration with the CRDR program is an ongoing process as long as design changes are made to any of the other basic elements.

jdm/056SB/27

e 3.3 Regulatory Guide 1.97 The elements and interfaces described in this section are detailed on Figure 3-3 i TECH GUIDELINES EOPS) Section 3.1 Line 2 R.G. 1. 97 PLAN Block A ine 1 R.G. 1L97.

DESIGN CR.ITER IA.

Block CONTROL ROOM I Line 4 Line 5

spos DESIGN IE-------,BASES ( INITIAL)

SURVEY See Section j dm/0565,B(.28

,Line 6 I

Figure 3-3 Reg. Guide.1.97 ELEMENT AND INTERFACES See Sectio

t :*

Reg. Guide 1.97 Plan - Block A Studies which provide a comparison of plant instrumentation with Reg. Guide l.9J _requirements are in progress.

These studies will be completed.

The results of these studies should be compared with the ongoing Equipment Qualification and Appendix R Work, the SPDS design, and the Task Analysis performed by Westinghouse as reflected in the Technical Guidelines.

Appropriate Integration with the SPDS design may minimize the number of modifications required on the control board.

When all design bases are established and modifications required for 1. 97 are known, schedules can be defined.

References for Regulatory Guide 1.97 (Section 3.3) e "Accident Monitoring Instrumentation Implementation Guideline", draft (I NPO 83~oxx RJTAc), ERc RJTAC "Comments on Regulatory Guide 1.97", AIF Committee on Power Plant Design, Construction and Operation, February 1982 "Report on Emergency Response Facility Accident Monitoring",

AIF Safety Parameter Integration Subcommittee to the Policy Committee on N.lclear Regulation, tt>vember 1980 "A PRA-Based Approach to Establishing Priorities for Equipment Qualification Needs",

D.E.

Leaver, W.A.

Brinsfield, R. N.

Kubik, presented at the International Meeting on Thermal N.tclear Reactor Safety, August 1982 jdm/0 56 58/29

AN:; 4.5-1980, "American N!tional Standard Criteria for Accident Monitoring Functions in Light-Water-Cooled-Reactors," December 1980.

IEEE Std 323-1974, "IEEE Standard for Qualifying Class lE Equipment for Rlclear Power Generating Stations",

IEEE, 1974 IEEE Std 497-1981, "IEEE Standard Criteria for Accident Monitoring Instrumentation for N..iclear Power Generation Stations", IEEE, 1981 IEEE Std 344-197 5, "IEEE Recommended Practice for Seismic Qualification of Class lE Equipment for.N.iclear Power Generating Stations", IEEE, 1975 EPRI Report NP-2110, "On-Line Power Plant Signal Validation Technique Utilizing Parity-Space Representation and Analytic Redundancy", EPRI, tbvember 1981 (an alternate to hardwired redundancy)*

EGG-EE-6043, "Preliminary Recommendations for Changes to Regulatory Guide 1.97, Revision 2", September 1982 Atf;I Nl.3.1, "Guide to Sampling Airborne Radioactive Materials in.N.iclear Facilities", 1969 Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled

.N.iclear Pow~r Plants to Assess Plant and Environ~ Conditions During and Following an Accident", us~c, December 1980 "Code of Federal Regulations", Title 10, Part 50, Appendix A, General Design Criteria 13, 19, and 64, published by the Office of the Federal Register, N!tional Archives and Records Service, General Services Administration jdm/0 56 5B/30

Generic Letter 82-09, "Environmental Qualification of Safety-Related

<l e

Electrical Equipment", USmC, April 1980 Memorandum and Order CLI-80-21, usmc, May 1980 KJREG-0588, Revision 1, "Interim Staff Position on Environmental Qualification or Safety-Related Electrical Equipment", usmc, July 1981 Division of Operating Reactors, "Guidelines for Evaluating Environmental Qualification of Class lE Electrical Equipment in Operating Reactors" (DOR Guidelines),

us me, N:>vember 1979 IE Bulletin 79-0lB, "Environmental Qualification of Class lE Equipment", USMlC, October 1980 NJREG/CR-1440, "Light Water Reactor Status Monitoring During Accident Conditions", usmc, June 1980 Supplement 1

to llJ REG-073 7,

"Requirements for Emergency

Response

Capabilities", usmc, December 1982 Owners Groups positions R.G. 1.97 Design Criteria - Block B An accurate assessment of plant instrumentation is needed to develop a set of explicit design and qualification criteria for necessary accident instrumentation.

jdm/0 56 53/31

3.3.3 Specific R.G. 1.97 Application (Initial) - Block C Completion of the studies above will define those plant parameters and loops that require modification or new loops that may be needed.

R.G~ 1.97 Plan/R.G. 1.97 Design Criteria Interface - Line 1 The R.G.

1.97 Plan will describe the development of criteria for the determination of required accident-monitoring instrumentation.

3.3.5 Task Analysis(Technical Guidelines (EOPs) /Specific R.G.

1.97 Application (Initial) Interface - Line 2 The information needed to mitigate degraded. conditions is identified in the

e Task Analysis/Technical Guidelines (EOPs) (see Section 3.1.4) and should be in the specific R.G. 1.97 application.

3.3.6 R.G. 1.97 Design Criteria/Specific R.G. 1.97 Application (Initial)

Interface - Line 3 The R.G.

1.97 provisions should provide a

basis for developing a

plant-specific instrument list.

jdm/0 56 58/32

e

-........ -~-. ------ *-* -----.. "" -:

Control* Room Survey/Specific R.G.

1.97 Application (Initial)

Interface - Line 4 An!- modifications initiated as a result of specific R.G. 1.97 application should be input to the control room survey (see Section 3.2.4).

The instrumentation identified as part of the Reg. Guide 1.97 listings can be used for Control Room survey.

SPDS Design Basis/Specific R.G. 1.97. Application (Initial) Interface

- Line 5 This interface is classified as two-way.

Credit may be taken for information displayed by the SPDS that meets R.G. 1.97 provisions on the plant-specific R.G. 1.97 instrument list.

3.3.10. Specific R.G. 1.97 Application (Initial)/Iteration Interface - Line 6 The specific R.G. 1.97 application should consider all changes associated with EOPs, control room improvements, SPDS design, and ERF ~esign.

This iteration is an ongoing process as long as R.G. 1.97 design changes are made to any of the other program deliverables.

This interactive process provides essential

coordination between R.G. 1.97 application and other RJREG-0737, Supplement 1, elements.

Changes in any of these elements may impact the incorporation of R.G. 1.97 provisions.

jdm/0 56 SB/33

- '",:.*~

3.4 Safety Parameter Display System (SPDS)

The elements and interfaces described in this section are detailed on Figure 3-4.

CONTROL ROOM Line 2 SPDS PLAN Block A Line 1 J

HUMAN FACTORS ENGINEERING Block B CRITICAL SAFETY~- I TASK ANALYSIS/

FUNCTIONS

, 1 TECHNICAL I

L{ne 6

t,, Line 5 Block C

.. ire 3 GUIDELINES See Section 3.1 Line 4 r-See Section 3.2 1

SURVEY

~~~~~--L SPDS DESIGN Line 7 1 ERF CRITERIA l PLANT SPECIFIC Line R EOPS (INITIALS)

See Section 3.1 I

I./

IJ j dm/0565,R/ 34

~

r BASES (INITIAL)

~

(INITIAL) i Block D I See Section 3.5:

I ine 9 FIGURE 3-4

sPDS ELEMENTS AND INTERFACES

)

3.4.1 SPDS Plan - Block A A detailed SPDS plan is essential to the development and implementation of a Safety Parameter Display System.

The SPDS plan should provide. guidance for developing the design bases needed to provide a concise display of critical plant variables to the operating personnel to aid them in rapidly and reliably determining the safety status of the plant.

This plan should provide a description of each task involved, including definition of source documents, establishing a schedule, and specifying a method of configuration.

References for Safety Parameter Display System (Section 3.4)

"Guidelines for an Effeqtive SPDS Implementation Program", (INPO 83-003), SPDS IDTAC, January, 1983 "A Parameter Set for a N.tclear Plant Safety Console" (.NSAC/10), lBAC

Fundamental Safety Parameter Set for Boiling Water Reactors"

(.NSAC/21), N3AC "Verification and Validation of Safety Parameter Display Systems" ( lBAC/39),

N3AC HJ REG-083 5, "Human Factors Acceptance Criteria for Safety Parameter Display System", US1'RC, October 1981 (draft for comment) jdm/0565B/35

-~ -.-.----

.~-

3.q.2.

Human Factors Engineering - Block B A description or human factors principles should be* generated as a basis for developing and assessing SPDS designs.

Critical Safety Functions - Block C Critical safety functions should be developed based. on function-oriented EOPs if available. If not, the Technical Guidelines may be utilized.

3.4.4 SPDS Design Bases (Initial) - Block D The design basis should define the minimum information required to be displayed on an SPDS, the users, the location, and the availability.

3.4.5 SPDS Plan/Critical Safety Function Interface - Line 1 The SPDS plan should describe the task of determining critical safety functions as part of the development of a SPDS.

The* method for safety analysis should be defined.

3.4.6 SPDS Plan/Human Factors Engineering - Line 2 The SPDS Plan should consider human factors principles.

jdm/056 ~/36

e Critical Safety Function/Task Analysis (Technical Guidelines (EOPs)

The Technical Guidelines are utilized to determine critical safely functions if validated EOPs are not available.

3.4.8 Critical Safety Functions/SPDS Design Bases (Initial) Interface -

Lih*e 4 The defined critical safety functions will be included in the initial SPDS design bases.

Human Factors Engineering/SPDS Design Bases - Line 5 Human factors principles will be utilized in giving consideration to the goal of SPDS usability; 3.4.10 Control Room Survey/SPDS Design Bases (Initial) Interface - Line 6 Enhancements associated with the SPDS may be included in the control room survey (see Section 3.2.4).

This interface may be classified as one-way or two-way depending upon the intended use of the SPDS.

If the SPDS is used only to fulfill minimum SPDS requirements, the interface is one-way.

If it is intended to enhance existing control boards and reduce inefficiencies by incorporating additional information on the SPDS, the interface is two-way.

3.4.11 SPDS Design Bases (Initial)/ERF Criteria (Initial) Interface - Line 7 Consideration may be given to locating an SPDS in one or more of the ERFs and the resulting impact on the ERF design.

jdm/056 5B/37

e e

3.4.12 Plant Specific EOPs {Initial)/SPDS Design Bases (Initial) Interface - Line 8 Conside~ation sho~ld_be given to how operating personnel utilize the SPDS in relation to EOPs (see Section 3.1.5). During the control room design review EOPs will be used to.verify the SPDS.

3.4.13 SPDS.Design B~ses {Initial)/Iteration Interface - Line 9 The SPDS design bases should be coordinated with changes resulting from other programs, such as EDP, R. G.. 1. 97, HED, and ERF. The iteration with the SPDS is an ongoing processas long as design changes are being made to any of the other basic elements.

3.5 Emergency Response Facilities (ERF)

The elements and interfaces described in this section are detailed on Figure 3-5.

L;ne 1

.ERF PLAN Block A Line 2 ine 3 TSC CRlTERIA I OSC CRITERIA I

£OF CRITERIA Block B Block.

Line 5 Line 6 SPDS DESIGN Line g ERF CRITERIA BASES *(INITIAL),___~.;.,;.;.;;;_..;...._ ____

"?!

&rnJPtL>

See S.e.t.U.o.n. 4 1--..:...:..;=r--__,

j dm/0565B/ 3&

FIGURE 3-5 ERF ELEMENTS AND INTERFACES Line 10 Line 7 NON UTILITY RESPONSE.

CAPAB.It. ITV Block

3. 5.1 ERF Plan - Block A A detailed ERF plan is important to the design and implementation of the emergency re~ponse facilities.

This plan should provide a description of each task. involved and the administrative guidance required to perform these tasks.

This would include defining source documents, determining manpCMer requirements, specifying the vendors involved, establishing a schedule, and specifying a method of configuration control.

References for Emergency Response Facilities "Guidelines for Implementation of Emergency Response Facilities", draft ( I NPO 83-0XX HJ TAC), ERC HJ TAC "Code of Federal Regulations", Title 10, Part 50, Section 47 (Emergency Plans), op cit "Code of Federal Regulations", Title 10, Part 50, Section 50 (Conditions of Licensees), op cit "Code of Federal Regulations", Title 10,. Appendix E (Emergency Planning and Preparedness for Production and Utilization Facilities), op cit HJREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency.

Response Plans and Preparedness in Support of N.iclear Power Plants", usmc, lbvember 1980.

jdm/0 56 !:a/39

Supplement 1 to llJ REG-0737, ":Requirements for Emergency :Respo~se Capability",

us~c, December 1982 RJREG-0696, "Functional Criteria for Emergency Response Facilities", usmc HJREG-0814, "Methodology for Evaluation of Emergency :Response Facilities" (Draft :Report for Comment), USM:iC, August 1981 HJREG-0818, "Emergency Action Levels for Light Water Reactors" (Draft :Report for Comment), usmc, October 1981 Eisenhut letter to licensees, 9/13/79, request for committment to meet requirements Denton letter to licensees, 10/30/79, clarification of requirements Eisenhut letter to licensees, 2/18/81, Commission-approved guidance on location, habitability, and staffing for ERFs.

Request and deadline for submittal of conceptual designs.*

COMJA-80-37, 1/21/81, Commission-approved guidance on EOF location and habitability Secretary Memorandum 81-19, 2/19/83, Commission approval of.RJREG-0696 as general guidance only Utility Emergency Plans N:>n-Utility Emergency Plans jdm/05658/40

3. 5.2 Technical Support Center (TSC) Criteria - Block B e

TSC criteria have been developed and incorporated into the design through review of the station emergency plan and the EPIPs.

Data requirements are yet to be developed but will be developed in the same manner.

3. 5.3 Operational Support Center (OSC) Criteria - Block C OSC criteria are provided in the present Emergency Plan and no change in intended.

3. 5.4 Emergency Operating Facility (EOF) Criteria - Block D The Station and Corporate emergency plans will be utilized to develop the EOF criteria.

jdm/05693/41

e

3. 5. 5 N:>n-utility Response Capability - Block E Criteria must be developed to enable personnel to identify and assess

.interactions between utility and non-utility personnel during emergency conditions.

These criteria may impact the design or manning requirements of the emergency response facilities.

The bases for these criteria should include me: regulatory requirements and utility and non-utility emergency plans.

3. 5.6 Emergency Response Facility CERF) Criteria (Initial) - BlockF The criteria developed for the TSC,

OSC, EOF, and non-utility response capabilities should be evaluated to ensure the integration of all emergency response facilities.

3. s. 7 ERF Plan/TSC Criteria Interface - Line 1 This activity has been completed for th.e facility but needs development for data displays.

3. 5.8 ERF Plan/OSC Criteria Interface - Line 2 This is defined in the current emergency plan.

jdm/056SB/42

3. 5.9 ERF Plan/EDF Criteria Interface - Line 3 This has been completed based on Corporate and Station Emergency Plans.

3._5-.19 ERF Plan/ N:m-Utili ty Response Capability Interface - Line 4 This is defined in the Emergency Plans specifying ho,, we interface with state and local governments and the ~c.

3.5.11 TSC Response Capability Criteria/ERF Criteria (Initial) Interface -

Line 5 TSC criteria should be used to determine initial integrated ERF criteria, including resolution of deficient criteria, resources available, and planning objectives.

3. 5.12 DSC Response Capability Criteria/ERF Criteria (Initial) Interface -

Line 6 DSC criteria should be used to determine initial integrated ERF criteria, including resolution of deficient criteria, resources available, and planning objectives.

jdm/0 56 SB/ 43

e

' e 3.5.13 EOF Response Capability Criteria/ERF Criteria (Initial) Interface -

Line 7 EOF criteria should be used to determine initial integrated ERF criteria, including resolution of deficient criteria, resources available, and planning objectives.

3.5.14 N:>n-Utility Response Capability Criteria/ERF Criteria (Initial)

Interface - Line 8 N:>n-utility criteria should be used to determine initial integ~ated ERF criteria, including resolution of deficient criteria, resources available, and planning objectives.

3.5.15 SPDS Design Bases (Initial)/ERF Criteria (Initial) Interface - Line 9 Design bases for SPDS should be considered in the design and layout of the ERFs, as appropriate.

3.5.16 ERF Criteria (Initial)/Iteration Interface - Line 10 The ERF criteria should consider changes resulting from control

room improvements, incorporation of upgraded EOPs, addition of accident monitoring

~nstrumentation, and SPDS installation

jdm/05658/44

3.6 IMPLEMENTATION The elements and interfaces described in this section are detailed on Figure 3-6. Although all of the lines illustrated in this figure are classified as interfaces, their descriptions are evident from the information provided for each element.

As a result, this section provides separate guidance on only two Jn~erfaces that deal with an iteration process {see lines 1 & 2).

! PLANT SPECIFIC SPECIFIC R.G. 1.97 I SPDS DESIGN I ERF CRITERIA EOPS (INITIAL)

HEDS APPLICATION (INITIAL)!, BASES {INITIAL)

! {INITIAL)

See Section 3.1 See Section 3.2 See Section 3.3 See Section 3.4 l See Section 3.

I l

I

I L-------.....1.~------t----------i..---------

HUMAN FACTORS

IMPROVEMENTS &

UPGRADES Block A J,

INTEGRAT lON/

~--------~------------~-----------------------,i~

ITERATION

/CR/SPDS I TRAINING FINAL SPDS Block D Block B

.I, I

'1'"----

I jFINAL EOPS i -lCR MODIFICATIONS '.FINAL R.G.1.97].. ERF CRITER1 Block C I

Line 1 I-Line 2 I

I I

I INSTRUMENTATION' !

, Block E I Block F Block G

! I Block H I

I I

=:J J,

COMPONENT J VERIFICATION Block I REVIEW Block J T---

IMPLEMENTATION Block L SYSTEM VALIDATION Block M

OTHER NRC i.--~--

ISSUES

' *Block K I

COMPLETED Figure 3-6 IMPLEMENTATION IMPLEMENTATION Block N j dm/05658/ 45

\\

3.6.1 Human Factors Improvements and Upgrades - Block A After.the HEDs have been identified as a result of the Control Room Survey and initial efforts are complete for the EOPs, R.G. 1.97, SPDS and ERFs, human ra.ator improvements and upgrades should be identified to resolve HEDS.

In addition to physical improvements in the Control Room, HEDs may be resolved through use of procedures, R.G. 1.97 instrumentation or the SPDS.

References for Human Factors Improvements and Upgrades

.MJREG-0700, "Guidelines for Control Room Design Review", US?EC, September 1981

.MJREG-0801, "Evaluation Criteria for Detailed Control Room Design Reviews",

usmc (draft for comments)

"Control Room Design Review Implementation Guideline", draft (INPO 83-0XX NJ TAC), CRDR NJ TAC 3.6.2 Integration/Iteration - Block B All the initial criteria (i.e., plant specific EOPs, specific RG 1.97 application, SPDS design bases, and ERF criteria) and all identified human factor improvements and upgrades*(see Section 3.6.1) should be reviewed in an integrated fashion.

This review should further assure that all criteria are consistent with each other, all required human factor improvements and upgrades are being resolved, and all necessary interfaces between the EDP,

CRDR, R.G. 1.97, SPDS, and ERF plans have occurred.

This review should identify all necessary changes to the respective initial criteria: in order to finalize these criteria.

jdm/056SB/46

(

e This review may occur more than once as a result of problem areas identified by verification (see Section 3.6.9) or other tf{C issues.

EOP/CR/ERF/SPDS Training - Block C Training. programs. for operating personnel should be developed based on EOPs/CR/SPDS Criteria~

References for EOP/CR/SPDS Training "Senior Control Room Operator and Shift Supervisor Qualification",

(I.NPO 82-008), I.NPO, September 1982 "N.iclear PCMer Industry Training System Development", I.NPO, draft December 1982 "The Accreditation of Training in the NJ clear PCM er Industry", ( I.NPO 82-011),

I.NPO, May 1982 "Task Analysis Data Collection Procedure", TA-1, I.NPO, December 1982 "Emergency Operating Procedure Implementation Guideline", (I.NPO 82-016), EOPIA Review Group, June 1982 "Component Verification and System Validation Guideline", draft (INPO 83-0XX IIJTAC), ERC.IJTAC "Emergency Operating Procedures Generation Package Guideline", (INPO 83-007),

EOPIA Review Group, February, 1983 jdm/05658/47

e "Control Room Design Review Implementation Guideline", dra~

(I ffl>O 83-0XX IIJTAC), CRDR RJTAC "Guideline for an Effective SPDS Implementation Program", (INPO 83-003), SPDS RJ',rAG., January, 1983 IIJREG-0660, me Action Plan Developed as a Result of the TMIII Accident",

usmc, May 1980 NJ REG-0899, "Guidelines for th~

Prepar'ation of Emergency Operating Procedures", usmc, August 1982 Supplement 1 to IIJREG 0737, "Requirements for Emergency Response Capability",

usmc, December, 1982 3.6.4 Final SPDS - Block D Design modification identified during the Integration/Iteration process should be incorporated to determine final SPDS design criteria.

3.6.4.2 References for Safety Parameter Display System

. See Section 3.4 jdm/0 56 SB/48

3.6.s Final EOP - Block E e

Procedure modifications identified during the Integration/Iteration.Process should be incorporated to determine final EOPs.

e References for Emergency Operating Procedures See Section 3.1.

3.6.6 Control Room Modifications - Block F Control room modifications as identified based on results from the Integration/Iteration process should be incorporated to determine final control room improvements.

References for Control Room Modifications See Section 3.2.

3.6.7 Final R. G. 1.97 Instrumentation - Block G Design modifications identified during the Integration/Iteration Process should be incorporated to determine final accident monitoring instrumentation.

References for R. G. 1.97 Instrumentation See Section 3.3.

jdm/0 56 513/49

3.6.8 ERF Criteria - Block H Emergency Response Facility modifications identified by the results from the Integration/Iteration process should be incorporated to determine final ERF des.ign criteria.

References for Emergency Response Facility See Section 3. 5.

3.6.9 Component Verification - Block I A verification should be included in the implementation program to ensure that each element achieves its design objective.

All element (SPDS, EOPs, accident monitoring instrumentation, control room,

training, and ERFs) should be involved in a component verification process.

jdm/0 56 :B/ ~

e

.-.... ~,....,:.~** *, ',*

References ror Component Verification

. Emergency _Operating Procedure Verification Guideline,* "(INPO.83-0XX) ", EOPIA Review Group, i983 "Verification and Validation of Safety Parameter Display Systems", ( IBAC/39) fBAC "Component Verification and System Validation Guideline", *draft (INPO 83-oxx HJ TAC), ERC HJ TAC 3.6.10 Review - Block J This review should compile verification findings and identify new me issuances and other regulatory guidance that affect the SPDS, control room, EOPs, ERF, Reg. Guide 1. 97 ins trumen ta tion, and training.

As a result of these findings and. issuances, a methodology for revising the affected elements should be determined.

References for Review Design Documentation.

New Regulatory Guidance 3.6.ll Other NRC Issues - Block K Future me issues that effect Supplement l to IIJREG 0737 provisions should be considered in the review process.

jdm/05658/51

3.6.12 Implementation - Block L Implementation should include the purchase,* installation,* and testing of new eq1,,1ipment resulting from the Supplement l upgrades.

Plant personnel should be fully trained on this equipment as well as EOPs during.implementation.

Reference for Implementation "Guidelines for an Effective SPDS Implementation Program", (INPO 83-003), SPDS RJTAC, January 1983.

3.6.13 System Validation - Block M A system validation should be performed to determine that all elements together achieve the objective of mitigating the consequences of emergency.

conditions.

All

elements, SPDS,.

EOPs, control

room, Reg.

Guide 1.97 instrumentation, ERF, and_ training should be included in the validation.

References for System Validation "Emergency Operating Procedure Verification Guideline", (I NPO 83-oxx), EOPIA Review Group, 1983 "Verification and Validation _of Sa_fety Parameter Display Systems", OEAC/39) tBAC "Component Verification and System Validation Guideline"-, draft* (INPO 83-0XX HJTAC), ERC HJTAC.

jdm/056SB/2

/ *-,.,

e e

3.6.14 Completed Implementation - Block N The implementation plan for Supplement 1 t_o

HJREG-0737

should be completed wh~ _- all equipment has been installed in. its permanent lo~ation, system

validation has sho,,n that. the plant modifications achieve* the objective of.

mitigating the consequences of emergency conditions, and plant personnel have been fully trained on EOPs and hardware.

3.6.15 Review/Integration Interface - Line 1 The revisions identified, if any, in the revi6l process should be incorporated_

in the final design criteria for the SPDS, control room modifications, R.G.

1.97 instrumentation and ERF.

.Final EOPs and the EOP/CR/ERF /SPDS training program should also be upgraded to eliminate _ discrepancies determined during:

the review.

Another component verification should be -.performed. once the affected element has been revised.

References for Interface See Section 3.6.2 through 3.6.14 r:

jdm/056 5B/ 53

e L

3.6.16 System Validation/Integration/~~!ration Interface - Line 2 The discrepancies. ideritifie~. during s:ystem validation, if any'

~houl~ be classified according to their. cause, EOPs, SPDS~. control ropm, ERF, traioipg,

... t

~ -

or -a

combination or.these element~.

Resolutions should*_ b~ :dev1?lppei1

a11d

-*~

appropriate

modifications. focorpora~ed.

Al}C')ther

~oinpC')pent

verifica~ipn, revie., and validation should be.Perf~rmeci if ~ignifican~. modific~tions a~e required to resolve the discrepancies~

References for Interface See Sections 3.6.2 through 3.6ci5 jdm/0 56 53/ 5IJ

ML18130A390

Text

1.2 BACKGROUND

Response

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