Text

Amend 11 to TMI-1 Restart Rept.
	ML19257D373
Person / Time
Site:	Three Mile Island
Issue date:	01/31/1980
From:	; METROPOLITAN EDISON CO.
To:	;
Shared Package
ML19257D372	List: ML19257D371; ML19257D373;
References
	NUDOCS 8002040229
	Download: ML19257D373 (200)
	v • d • e

TABLE OF CONTENTS Pace

1.0 INTRODUCTION

AND REPORT ORGANIZATION 1-1 1.1 Introduction 1-1 1.2 Report Organizat ion 1-1 1.3 Abbrev iat ions 1-2 1.4 De f ini t ions 1-2 2.0 PLANT MODIFICATIONS 2.1 -1 2.1 General 2.1-1 2.1.1 Short-Term Modifications 2.1-1 2.1.1.1 Reactor Trip or Loss of Feedwater/ 2.1-1 Turbine Trip 2.1.1.2 Position Indication for PORV and Safety 2.1-3 Valves 2.1.1.3 Emergency Power Supply Requirements 2.1-5 for Pressurizer Heaters, PORV, Block Valve, and Pressurizer Level Indicat ion 2.1.1.4 Post LOCA Hydrogen Recombiner System 2.1-8 2.1.1.5 Containment Isolation Modifications 2.1-11 2.1.1.6 Instrumentation to Detect In ad equat e 2.1-17 Core Cooling 2.1.1.7 Auxiliary Feedwater Modifications 2.1-20 2.1.1.8 Leak Reduction Program For Systems 2.1-29a Outside Containment 2.1.2 Long-Term Modificat ions 2.1-30 2.1.2.1 Post Accident Monitoring 2.1-30 2.1.2.2 RCS Venting 2.1-31 2.1.2.3 Plant Shielding Review 2.1-31 2.1.2.4 Post Accident Sampling Capability 2 *.-33 2.1. . 5 Reactor Coolant Pump Trip on HPI 2.1-34 i869 097 i Am. 11 8002040 k 77

TABLE OF CONTENTS - Continued Page 2.1.2.6 Auxiliary Feedwater System 2.1-37 2.1.2.7 Increased Range of Radiation Monitors 2.1-38 3.0 PROCEDURAL MODIFICATIONS 3-1 3.1 General 3-1 3.1.1 Emergency Procedures 3-2 3.1.2 Administrat ive Procedures 3-2 3.1.3 Surveillance /Preventat ive Maintenance / Corrective 3-3 Maintenance Procedure 3.1.4 Operating Procedures 3-3 4.0 EHr".CENCY PLANNING 4-1 4.1 Introduction 4-1 5.0 THREE MILE ISLAND NUCLEAR STATION ORG ANIZATION 5-1 5.1 Gcneral 5-1 5.2 Station Organization 5-2 5.2.1 Vice President 5-3 5.2.2 Manager Unit 1 5-6 5.2.3 Supervisor of operations 5-7 5.2.4 Emergency Planning - Coordinator 5-10 5.2.5 Supervisor - Radwaste, Nuclear 5-12 5.2.6 Shift Supervisor 5-14 5.2.7 Shift Foreman 5-17 5.2.8 Control Room Operator 5-20 5.2.9 Auxiliary Operator 5-21 5.2.10 Superintendent of Maintenance 5-22 i869 098.

ii Am. 11

TABLE OF CONTENTS - Continued Page 5.2.1) Supervisor - Corrective Maintenance 5-22 5.2.12 Supervisor - Preventive Maintenance 5-24 5.2.13 Maintenance Foremaa 5-25 5.2.14 Lead Maintenance Foreman 5-26 5.2.15 Manager Plant Engineering 5-27 5.2.16 Ch air man , Plant Operation Review Committee 5-29 5.2,17 Lead Engineer Nuclear 5-31 5.2.18 Lead Electrical Engineer 5-33 5.2.19 Lead Instrument and Control Engineer 5-35 5.2.20 Lead Mechanical Engineer 5-37 5.2.21 Sh if t Technical Advisor 5-39 5.2.22 Supervisor Chemistry 5-42 5.2.23 Technical Analyst - Fire Protection 5-44 5.2.24 Manager Administration and Services 5-46 5.2.25 Manager Training 5-48 5.2.26 Supervisor Operator Training 5-50 5.2.27 Supervisor - Technician Training Section 5-52 5.2.28 Supervisor - Career Development Training 5-54 5.2.29 Manager Radiological Controls 5-56 5.2.30 Supervisor Radiological Controls 5-59 5.2.31 Radiological Controls Foremen 5-61 5.2.32 Radiological Controls Technicians 5-63 5.2.33 Supervisor Radiological Engineering 5-64 5.2.34 Radiological Engineers 5-66 iii Am. 11

TABLE OF CONTENTS - Continued Pace 5.3 St at ion Support organization 5-68 5.3.1 GPUSC Technical Functions Group 5-69 5.4 Quality Assurance Program and Procedural Control the TMI-l Restart 5-72 5.4.1 Introduction 5-6 5.4.2 Quality Assurance Department 5-72 5.4.3 Program 5-75 5.4.4 Procedures 5-79 5.5 St ation Organization Under Accident Conditions 5-79 6.0 OPERATOR ACCELERATED RETRAINING PROGRAM (OARP) 6-1 6.1 Introduction 6-1 6.2 Program Object ives 6-1 6.3 Topical outline 6-2 6.4 Program Rat ionale . 6-5 6.5 Instructional Procedure 6-6 6.6 Evaluation Procedure 6-9 6.7 Program Forrat 6-11 7.0 RADWASTE MANAGEMENT 7-1 7.1 General 7-1 7.2 Separation and Isolation of the Units 7-1 7.2.1 Radioact ive waste trans fer piping 7-1 7.2.2 Fuel Handling Building Environmental Barrier 7-3 7.2.3 Liquid Radwastes and Miscellaneous Waste Evaporator 7-3 7.2.4 Solid Waste Disposal 7-4

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iv Am. 11

TABLE OF CONTENTS - Continued Page 7.2.5 Sanitary Facility Drains 7-4 7.2.6 Radiation Protection and Decontamination Areas 7-5 7.2.7 Nuclear Sampling end Radiochemistry Laboratory 7-5 7.2.8 Industrial Waste Treatment Facilities 7-5 7.3 Supplemental Topics 7-6 7.3.1 Radwaste Capability 7-6 7.3.1.1 Liquid Radwaste Processing 7-6 7.3.1.2 Waste Gas System 7-9 7.3.1.3 Solid Waste System 7-11a 7.3.2 Plant Shielding 7-12 7.3.2.1 General 7-12 7.3.2.2 Design Review 7-12 7.3.2.3 Near Term Modifications 7-12 7.3.2.4 Long Term Modifications 7-12 7.3.3 Auxiliary Building Ventilation System 7-12 7.3.3.1 General 7-12 7.3.3.2 Testing Requirements 7-13 7.3.3.3 Implementation Schedule 7-14 7.3.4 Nuclear Sampling 7-14 7.3.5 Nuclear Sampling Capabilities 7-14 7.3.5.1 Po s t-Acciden t Sampling 7-14 7.3.5.2 Sample Drains 7-15 7.3.5.3 Improved In-Plant Radiodine Monitoring Instrumentation 7-15 7.4 Affect of TMI-2 Recovery on TMI-l Operation 7-15 1869 101 v Am. 11

TABLE OF CONTENTS - Continued Page 8.0 SAFETY ANALYSIS 8-1 8.1 Introduction 8-1 8.2 Areas of Investigation 8-1 8.2.1 Modifications Resulting from the 8-1 August 9, 1979 Order 8.2.2 Modification as Result of Order of May, 8-2 1978 8.2.3 Modification originating from within Met-Ed 8-2 8.2.4 I&E Bulletin 79-05C 8-2 8.3 Effect of Changes on Safety Analysis 8-2 8.3.1 Rod Withdrawal from St artup 8-3 8.3.2 Rod Withdrawal at Power 8-3 8.3.3 Mcdarator Dilution Accident 8-4 8.3.4 Cold Water Addition 8-5 8.3.5 Loss of Coolant Flow 8-5 8.3.6 Dropped Control Rod 8-6 8.3.7 Loss of Electric Power 8-7 8.3.8 Station Blackout (Loss of AC) 8-7 8.3.9 Steam Line Failure 8-8 8.3.10 Steam Generator Tube Failure 8-10 8.3.11 Fuel Handling Accident 8-11 8.3.12 Rod Ejection Accident 8-11 8.3.13 Feedwater Line Break Accident 8-12 8.3.14 Waste Gas Decay Tank Rupture 8-13 8.3.15 Small Break Loss of Coolant Accidents (LOCA) 8-13 8.3.16 Large Break Loss of Coolant Accidents (LOCS) 8-17 8.4 Summary and Conclusions 8-18 vi Am. 11

TABLE OF CONTENTS - Continued Page 9.0 DRAWINGS 9-1 10.0 CROSS REFERENCE TO ORDER RECOMMENDATIONS 10-1 10.1 Introduction 10-1 10.2 Short-Term Recommendations and Met-Ed Responses 10-1 10.3 Specific Responses to Recommendations 10-4 10.3.1 Response to IEB 79-05A, Item 2 10-4 10.3.2 Perf ormance Testing for PUR Relief and Safety Valves 10-5 11.0 TECHNICAL SPECIFICATIONS 11-1 11.1 Int roduction 11-1 11.2 Technical Specification Changes 11-1 11.2.1 Auxiliary (Emergency) Feedwater (AFW) 11-1 11.2.2 Reactor Trip on Loss of Feedwater or Turbine Trip 11-1 11.2.3 High Pressure Trip Setpoint Reduction 11-1 11.2.4 Containment Isolation Setpoints 11-1 11.2.5 H'ldrogen Recombiner 11-2 11.2.6 TMI-1/TMI-2 Separation 11-2 11.2.7 Administrative Controls 11-2 1869 103, vii Am. 11

Systems to be leak tested will include but may not be limited to:

1. Makeup and Purification (including RCS letdown)

2. Decay Heat Removal System

3. Waste Gas System Consideration will be given to the type of potential leak paths identified by NRC dated October 17, 1979 concerning the North Anna event.

A summary description of the leakage reduction program will be provided by January 1, 1980 and the program shall be imple-ment ed be f ore rest art .

2.1.1.9 Automatic Closure of the Pressurizer PORV Block Valve.

2.1.1.9.1 System Description A modification will be installed to automatically close the Pressurizer PORV Block Valve (RC-V2) on low Reactor Coolant System pressure. Its purpose is to prevent an excessive loss of reactor coolant inventory if the PORV (RC-RV2) f ails to close af ter opening on a high pressure excursion. RC-V2 will close automatically if the Reactor Coolant System pressure is below 1600 psig and the PORV is open. An open PORV will be detected by flow in the discharge line from the valve. The automat ic closure signal will be bypassed when the PORV mode selector switch is in the NDT protection position.

2.1.1.9.2 Design Bases The automatic closure of RC-V2 is designed to prevent excessive loss of reactor coolant inventory if the PORV f ails to close af ter opening on a high pressure signal. The modification will be designed so as not to degrade the existing protective functions of the PORV. It will also not prevent the operator from manually relieving through the PORV when the plant procedures call for him to do so. Means will be provided to retain the low temper-ature NDT protective function when the plant is shutdown. The design will preclude automatic cycling of the block valve.

2.1.1.9.3 System Design RC-V2 will be automatically closed if the Reactor Coolant System pressure goes below 1600 psig and the PORV is open. The low pressure signal will be derived from the ES Actuation System.

1869 104, 2.1-29b Am. 11

The "PORV open" signal will be derived from the PORV Flow Detector System. (See Section 2.1.1.2) the block valve will go fully closed when a close signal is gen-erated. There will be no automatic open signal to the valve. An alarm will be actuated when RC-V2 leaves its fully open position. The operator will be able to open the valve manually af ter the close signal has re-set.

The control circuits shall be supplied from on-site power of the same power train as the AC supply to RC-V2. The automat ic closure signal shall be by-passed when the PORV mode selector switch is in the NDT Protect ion posit ion.

2.1.1.9.4. Design Evaluation The design will provide a reliable means of preventing excessive reactor coolant inventory loss due to a malfunction of the pressurizer PORV. No operator inter-vention is required. It will not degrade any of the pro-tective or operational functions now provided by the PORV. The cont rol circuit will be supplied from on-site power sources. Any new components required will be specified to be suit able for the environment in which they will be located. No new components or wiring will be required inside containment.

i869 105 ,

2.1-29c Am. 11

APPENDIX 7A THREE MILE ISLAND NUCLEAR STATION UNIT 1 RADIATION PROTECTION PLAN REVISION 1 186910(.11

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Cnangc-- to this 6:2:1.Tcnt rcglire approval by these positionc.

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.a e-t i c] e . I',esponsibilities of V:orkers 5 Trti clt 3 Au'iits, B" vie..s and Reports on the 'H:I-l Radic]m;; cal Ccntrols Prcgra.u 7

.eticl e. Paf.jologicc3 Centrols Tretining 10 IrLicle 5 Control of Pxternal INpcsure 12 Icti r. lo 6 Contrul of Inte:nal D:posure 15 Artic]c 7 Ccntrol of Radicactive CentccrJnation 17 f.rticle E:

Control of Radioactive I4cterials 30 leticle 0 Oro.mirr. tion for Radiological Controls 20 DMl i/a j'bblij r.-ll0,3

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Tn:s ducc-e .t , t i r. ". "cn :ile Iclar5 Unit 1 Rc h tien Proce._ur Plur, scts fort). O.;- philo.sophi cs, basic policics and cbjectrces c .

'*ctrqulitan Edicen Cm ny and Ccneral Pc.blic Utili tics Corp:re'; ~

conce ning their 2 2-1 i:adio}ogical Ccntrols Puuc.i. Tne cbj e cc u .

tha radiologiccl cc .;rols prc; rr.- is to centrol rcdlation hct- i c -- _

ovcid accid,ntal raC etic.i cq csuces, to n.cintcin c.:pcaures .

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p':.ilascphi c:> , poli ciel. , and objectives tre based on cnd stem frcr thi-regulation of the ' ccl:r Rcgulatory Cc=.ission ("RC) as contt nci _r Title 10 of the Ccde of FeSaral Regulations, Parts 19, 20, 5 , and 1.

cn6 cppropriate Rcgu' ato.'f Guidc;, spccifical.ly 8.8 Rev. 3 (1975.;, 2.1 Rev. 1-R (1975) , 8.13 Rev. 1 (1975) , and 0.15 (1976) . Tne TC-1 -=.u ttier Protection Plan is hcrel on these referenccc, tiurefore thcy tre n::

rep ated throughout thc reaainder of this docrent.

Spacific detcils as to ho.: the 'ItE-1 Radiation Protecticn Pltn is implerented shall be prona.0 gated in the TC-1 Radiological Cont cls Procedurc.- Manual (R5G) and shall incluSe those cpplicable prcccircs addresned in Rog. Guide 1.33 Re'.'. 2(1978' , App. A, paragraph 7, a .i paragraph 8(aa) , (t:b) : further references to the TG-1 RCPM a e nc .

rcp?cted throughout this docu~cnt. Tne TC-1 R?M will consist Of revisions to procedres which evi cwa in the previous HPP 1600 E .6 l' :

series, applicable A:hinistrative prccedures, and additional precedres deened necc=sa y. Tnis 'IIC 1 wdiation "rotection Plan is the firs:

part at, the ua-1 _n m....

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M nn, but are ad.! rem;eJ in th.' Pnvi ren:rmtal Technical D> ci Mem ; a .

Vechat iin ccrplience with t he %C-1 ici. ' is nr rrlatory. In the evc:it a J .;i.rcci.ce co rr.ot lv fo] ] v. 03 cactly, wo ', und2r tirat. proccclure shall b" stopped e r1 J..ll rot cc.r:cncc again tu.tia th<' procedure r.c been correctt .

Tois. '271T-1 Rad ct1cn Protection Pler m:5 the nw 'I:2-1 TCi .' are b.' ire

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'1 : rr . "_"l Ur.i. 1. " . . _ _ , tt 7: r r t. c. - * .zirm '. c7plicac11 m c .1 t o 'IMI U;;i t 1. Proced tres sh:311 prcvj dc adcriu7te guidence an] cp?cify cpproprj ate treth:xis or tcchniques to irmtre that the perfornunce of each cctivity ir in ecendunne with counti rcra cicaical cont rol principles, and 2n cza 1;cnra vit'r cpp1m cle Icgelutcry provisions. 'Ihc RTM chall b2 pr e,m red , revic.te. , cpp: o'; 'J., c:.0 controllei as dccerib i in the R1W Jdrjnict rative procciiures.

The TMI-l Radiolcgical Controls Progr_t.. is to be fully integrated into each and evcry r%se of aparations at PG Unit 1. She 'HO-1 Radiclogical Ccntrolc Program whu. carried out as spacified will ascure that the. operaticn of Unit I will be parfonred with pcrsonnel t;ho v ork at the nite incurring radiation e:q mure en lua as can reasonably be achieved.

In order to la':t this object ive, the program nuct be carrj ed out by each person involved in the TG-1 activitics. Tnare is no grepp or person involved j n the 'nc-1 operations who cars r.ot have sona degree of rerponsibility for the Radio]ogical Ccntroh, Program. I'aihire of any person to recognize this respnsibility or to conply with issued proccattres will r.ot be tolerated.

I. re:'.ielogi.cally cafe opxaticn uill b" cc'. iev:d if each inividual carrit , out his ur hur resp;nsibility.

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Revision i Jar.uary 16, 1%3 Lu.L'Lc.l d . U : 'ihU i.trr.rHTi.:- .

Al thatub lercuryl specially trdr-J. in re3iological controb .or-nally ov ;:see. ratuoactiv2 work, t s ch inflvd.;E*' inval."r? in thu,' '-

I mt ecnita:.tly Ic.~-tir a'.iare of the potential radiological p' ;'.;ic E2ch ;ndiv m a:1 is teapontjble for trairPulnita h:n or her c: 7 ~ :. .. v J c.; ..r ; c:' s nchh . J en1} e. Each irfividt.cl '; articn.s directly af fc".

his t= cure, conttini. tion, an:] overali rafich acd prcbleu ass: cia' ..

v2th tha vori- 'lh. follo.ct:0 mle shell b: f ell o.s - 1 t y indivi6n ir to

..inuu . radiological problccc..

1. Ob. y promptly "stop-. ' rk" and "evruate" orders of radlolcgica' cuatrol personnel.

1 O'.1y Iw ted, oral,cnd written rcdiological control instructions and proceaures, including instructa ons en Radiation Work Pericit.s .

3. Ucar RD and self rcading dosieter wh re required by sic".s er by radio 3cgical control parscnnel. Rcrort locs or une: g cted exp3sure and offscale docimcter to Radiological Control D"partrcnt.

4. Keep traci of personal radiation expanure status and avoid ex-cocding exposure limits.

5. Renain in es la.e a radiation aren as prccticable to accorplish vorh.

6. D3 not loiter in Indiation arcs.

~1. D3 not Fr..DI'.e, eat , or che ' it cor.tinineted arena.

8. Kear anticontanination clothing nr.d respiratory protecticn pro .:rly and wherever rcquired by sig.s or radiological control personnel.

9. l'enove anticontacu nation clothi ' cr.d respiraioty protcetion p;op;ri'f to minirrize sprced of conte:ination.

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tu a cr r_Siolcgic,1 control point. i;atify h:=1, . .c:

Cu?.rclu personnel if conte.rir.e tion 2. founf.

31. For a Irc.. or p;saible reO osctiva spill, riniri.x its sprc d and not_f . rt.diolc;;ical control personnel prc. ptly.

Iz. 1:o at rr.c-cc.ssa--ily touch a contcrinc:cd srf ace cr .11r1 clattir- - ..c 13 . cr c. r - '~uirmsn* to do c:

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able scftres (for c>:srplc, sheet plastic) when not :.r :se t .d inside plastic b .gs w'.en work is finished.

14. Lirit the Enr;unt of raterial that has to be ciecentEnnsted er dispM e5 ci Es ccS cactive waste.

15. Notify Reho.logiccl Controls personnel of faulty or ala=.:;

radiat:Or protcction eIuipment.

16. Report the presence of open wounds to radiological c:ntrol cr.d rcdical parsonnel prior to work in areas where rcd .ca-- .ive contt --

inaticn exicts er.d irrediately if a wound occurs whi c in such an crea.

17. Notify Rr_diological Centrols personnel upon returning to t?" site after medicel cirinistration of radiophaaraceutit_'ic.

18. Assure := T entally alert and physically sound conditier fc:

parforrir; assigned work.

19. Ensure thut your activitics do not create radiolcgical prcble .s for others and be alcrt for the possibilitics that the activitics of others ray cher.ge the radiolo;dcal conditions to whic'r y:n are exposed.

6-1869 113 .

Revision 1 Janua ry 16, 1930 Article 3 - Pr0itcu Tgv i 07 , at:d Peggffi en tb ' 'rII- 1 P;dialorri.r;].

Cmi rol > prancam As indicated in Article 2, each individuel is responuib3e for trainta.ining his or h"r taSintion expostre as lcw as reascneble achiev-zSla v.hil , corsleting the scopa of werk they are required to parfo r.

Euch will be required to co: gly with the applicaale proced'rca of the

%.1-1 RO'. . end the specific radiological controls prescribsd for vari- 5:

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.- c& to enor. 2 "nat tWc re:p'irewau c e ba c.; ret rmf Oc ascict all site perconnel in understanding and ccuplying with these re-quircnents, the fo13caing audit and review procec ces shall be used:

1. Radiological. control technicians shal) nonitor and aid the perfor:rance ci each individual insofar as radiological ' ark practicos are cancerned.

2. Tne Radiological Dair.cerin;.1 chall review on a rcalar basis the perforence of the radiological control technicians.

Tnis review includes shift covemge on those jobs which are considered likely to have a high potential for radiological difficulties.

3. Radiological assess r.cnts shall be conducted throughout the Radio-logical Controls p.ogram on a continuous basis. '1his assessrent function shall report directly to the high:st level of rcier.ags-mont in 'the OMI-l organization and shall be outside the R3dic-logical controls D:3partnrnt. A written report of the findings of this assessmant shall be prepared and icsued every r: cath.

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, , . Otu litf 70. urance aviitr. shal l be ecc ..:ctc.3 cf the 2 ::-:

%3iological Contcels Progra.n by tech:d call.y gralified parco..s from cutt J2 the Radiclegical Controls I.'eprt ent. '?r_sc c .!its wil) La cond.:c':cd i:. accordance uith proceduras as cuti_nC j n the 3;C-1 Ocality .'.ssura.re Plan. 7 ne Quality J.sn. ; .c=

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audits shz ll cover the applicable portions of the ;cali f Assu-rance Plan, the 'D2-1 R;diation Protection Plan and a'l procede.res in the TC-1 RCE1 on at least an annual basic.

5. Toe Dicnt 0 rrations RevLe.e Cem,rattte shall rc. e. r.d c;nmnt en the TtC-1 Radiation Protection Plan and any cir.gcs thecto.

They also ' nave the responsibility to rcvic. those procei res requested oy the 'anager-P.adiological C;r.trols.

6. Periodically, the semces of an outside consultant ..t11 be rutairexi to provide evaluation and guidance c . rays to i. grove the T2-1 Radiological Controls Program.

7. In addition to these revic s and audits, a system chdl be c: ployed to identify radiological control deficiencies. A raiclcgical control c,en_ .clency is ca.,inco as either a vlontien c: an establiched prew5ure or a practico which could and shculd be inproved. Such deficiencies are recorded in a Radiological D2ficiency Report. This system shall be specific 4 in the itE-1 RCP:1 e-bodying the follueing concepts. A Ra:delecical Deficiercf Report Iray b initiated by any individun '.to 8.0 1869 115

Revision 1 Jcnuary 16, 1983 oLcerves a deviation from g >od radiological practices. Tnse reports shall be evaluated by RadioJojical Engineering for d:sirable or necessary corrective action. Tne purpa:<e of this systen i:s to identify all deficiencie:3, regardless of hc--

snall or incccacquential, correction of which veill result in an inprcm3 Radiciogical Cc :trels hogre.T. Radiological

4__: _: m . >' -

-m .-.:n:hl; r ,rt 7trnerizing the i .'..cl .;i t. 1 Le' t ri _.c_cy n.:~ :t fin'--

a.. Correrriw actio-taken.

8. Tne Euclear Regulatory Cecrission (hRC) el.:.o ins,m cts and revie. s the TIG-1 Radiologi cal Controls Progrem. Tne 'G-1 Raaicticr. Protecticn Plan cnd r.y cha". gas t'c.2rcto shall b' stimitted to the NRC for caninent.

9. In the cvent all the preceding measures fail to prevent a radiological incident, and investigation shall be ca. ducted to acte: Trine the catwes of the incident and to determine the (orrective actions and inprovements needed.

O" O h5)Yl b" f_ -

A dlmhML 9.0 1891b

3r .m .-_

r n. . . ,

fl " ;,' / .' 'J,N.. )-

@b;bMUW'IL)ah@ ] iip [J h 'O' U Dill[D)f abl O < b'r M'

" ' - ' ' ~ I c md I.-ti r b , - R'.J o: . c c l "- trr 2 "M ai n m-

1. Per.ic ; .c rcdiolorical control training sh all be gi'. ;r t-r ennur: es '.. :.crson 1:ndars ta..J.s the ra;.wlc;;ical ccr:'it; r .= to which h.- j s cxysed, understcr. 's his reipa:,sibility to "inirc .- u his c'.m e , caw.c to radletion, and tr. Car: tend his tvn r = S:r e . -

bilita= fn c::pl,inc ..ith rac. ole;;.. cal central p::.:. --._ e-r.crse nel ::ru;stienell,. c:m :: 6 to ra;ietier cal'. rrec_ . .

.:...>_.a.-

. . .. . - , +g._.

. .. ,...:,.u..

,c m

, -,. .. o ...-.-..n.c,...

- . . .-__-..+,.:.

,-; c ' _ . .

. . . . . u.-

with re .ation excosure.

2. Gancral rediolcrical irdoctrination ci. j l be given tc th: 2s not directly involved with radiation so that they tr.fers c. .d wt to --: _ .:: crian req iring TA and nr.t to crc =r rad _ f.c--

barriers. The indoctrination shall includa explcneticr of '.nc-radiologic;l cnviron. Tont in which they work.

3. Radiolo;;ical control trainir.g shall be given to pera .. .el requiring access to a restricted crea. Tnese personnel shall be regiar: 1 to pass a written exrtination, and they shall requalify bj .-critten cxaTinction at J eest annually.

4. In addition to the training and written c::aminations of parc-graph 3, th:ce who require eccess to areas controlled L_.

=

nadiaticr .:ork Parrits shall receive m2re extensivt. training and shall ce requirec. to pass a radiological exar.lirtric . t.'

their practical cbilities, including use of dosirretr/, frinking, anticonte_T.ination clothing, respirators, and recpor.sc to rn-itsual cituaticns. notraining, and bcth written end pra.ctical examinatien shcIl 'oe conducted at lea >t annually. Ir cddition, 10.0 1869 117

! g 'l' P [T.[ @

(() 0d J J@bi@$)a ,

, j gg

not cheAs reh 11 be rude that they retain the te.xiuircd kno.;

ledge during th p>riod hcLwen exa:riraticnr Special brief.1r rs and extra trainin ; includina me of nachu,rc w'.:ere applienble, siell be cc.:xiucu.d fcr wori: involvi.r.g highe r than usual c' a t.m to radiation and :.ndicactivity.

5. 1:2diolc4Ticui control ted.v.tcices and their forenen shali r -

_ , , , _ , - : n;

_2 3 --r -.;3.,ni ;, a +; ;, ning ic, ; _

.:-: ul 21'.E~. ' '. w . _ Tr d n?. . . 12 C1EC hi ~i'.*Cn eor C'.J'L'f'~- tO procedures, c: pip:.ent and progrur.s. Tney chall pus both critt -

c r and oral exc':J. nations, in which the passing grac,e tor torcr.cn shall be higher than the passing grada for tcdnicie .s. P2riedic practicai cl'.' and cral ch..'.'in shall be rcqur d for each tei-nician and f'm.;n. Annual rcqualif' eation chall be rcquired including both written and oral exrinatiom. Radiologicc.1 control technician assistants chall pm forin specific functicns under direction of a qualified technician er foreman and only af ter being qualified for the specific functier.

11.0 1869 118 .

r- -

h[ d g 7- j, - ' ., 1 d g" 'd d 3 ! .J F :'j l5, l '.l h 3

i,r <. c 1. c r3 ,_ pr. .. . .. . . . .-

p.y... ,,s. .a a .sy ,, u q.c CZ h ol ci re[ M_lon c:gcC.ure iG bO5 C CD thO CsStrrl:Eic.c .. 3 - : .f c.-,.._.

... .- .,1...'+.,.

.. .-.r. u. i. '.. ..,

uc..-.su. n

~ .v _i - ,1, . } w. . v or , ev, ,.-e_x_ ..-

. .._;-..h.._

in th2 OcCcptru 112.._; repM.90. Er C riO\ Droll CcTp3 red wit ".rl'_

}..- _.m. Q, g r _i _i c.-

. m.

C.

.. .- . , forF- "a - #. "V' l i y-"

.. w^ # " " " ' '

.- . ,"_C.l .it_un ~F . . .' '

s-- __.-.

m. -' . _e nn. ;,1_ - . ' .__,

3 $ 4

. ._; '_ .,. 6 -_

4 c._.;_ c.v ,.m, .. g,. 4,. .; ,,,r. . e,3, m,-., .4 , ..

, .; n, fa.,.

_. -g y ,. _ .. .. . _ ._.

jrLIVidl81S C' tC;E ' T M -rc~TS c.a l O.s' Es iS IOS:On3D3y aChi F _:_. ' .C' - 2 ' .

T. . . n.:_.,

1.

.~...--

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. , '1 A.m-...,.~..+..

- f . t, e. e. n.i_l r. : - - - -

- encb i .,. -7, m, , . - _-

-- w

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-r .. ms 4 . , y c__ .mv i n. 4.- _v-m

,r- 4

,_ w 6. 4vu -- C.' 'q". ^m. ' "." -- i 'ductio^4.

To aid in C?gC'3Cr0 TCdCChiCL, ad:dniStr2.t1Ye radiation c:g"5.a R3d.LatiOn D;n~ rem exp: =.=c g::.:

Control 30VC;1S Shall L3 CStEbll:.03.

c- .m.,,

m ._

m_ _m : . s. _, <..-. ~. .j._ _.-n._- .y q_. ._,f.- .-

j ou, s_. - w_ c-- c. g-4 v, t ., . . .

i_ . _ ..

.._-1..

s. .

radiction expont rc shall b2 .erc.ulare d. Mas ior e:r2xx3nrc .iobs =htll rc. irs .

th? rneiological centrols be ic~;rparnteS in the design, that .ritten precedure , by prcpared, end that pre-job briefing and rehearsale be c:ncccted prior to cacrencing ,icrk. A Rad'.ation Work Pc: Tait will be rcquired for any work or entr to restricted arean that would involve or create c .y cf the following: (e) :.1gh radiction crea, (b) airbo ne radicactiv".rj area, and (c) contaminated area, or (d) those radiation areas sp3cified ir applictble procedures.

Restrictcd car, used to ocntrol personnel access to radittien tn .

rz.dicactive materials shall be defined, access controlled, and p stcd ir_

eccordance with 10 Cr', 20.203 with the folla. ting rcodificaticns:

1. Each i:igh Radiation Arce shall be barricaded and ccns-icuousiv. z posted ca a High Radictier A-ea, end personnel desi:ir~ en_

trance shal.1 cStnir t Radiation Work Pc:' tit (R'@) . 2r/; indi-vidual entering a High Radiation Area sha . (C use r 6:FO 1.,_ . 0 1869 119

D97D Ej I

_M o d n

fn@dlioal@f\JM 7.

dudJ u .M [m!;q 6 Jam.T.o1.i y.

s, 2.

IaN conitoring device or (b) use c. radiation dose rate integating device which alarm at a preset dNe level, or (r) m..are that a ru.hc iojical coc; trol t echniciar pre eides par.i o.iic rndiation swv::illan c vith a dose rate ironitoring in.ctr c _ -* .

2. 7.ny area acce:ssible to p rrorc.cl where a Ir.njor portio: cf tha bully exdd rum .u -.y ee t hov a d se in c;;cr s of one thr' ' '

'. r~- shall b lc c' ? "' covent truuthorincA on t.n; .

' ' ,= ' r . s t: , m r'.-a ur ricades sh311 b rairtal-c' rn9r tha edn.inistrctive cc>citrol of the Radiolo',:ical Controls Forcr.2' on duty in accorde.nce with the RCT:..

Radiological Centrols personr.cl shall be e': rpt from the E9

.i tmus: - re virer.w t kr.tng the pirfc:rance of their assigc ed radiction protection duties providing they are follo. zing radiologicai_

c entrol procedurer for cmtry into High Radiation 1."cas.

To evaluate radiolc-;;ical conditions, radi.ation surveys shall b2 conductc5 for air activity, rc:.pvable surface conta:ninatien and cxtemal radiatio:1 at re.jular intervals. Surveys are perforned in order to (a) nonitor the suitability of control tr.easures, (b) cvaluate the needs for additional enntro]s, (c) cvaluate trcnds for IJ1m purinacs, and (d) cvalunto radiological conditions in cu; . 3 routirga; entered without radiation work parmit coverago.

Surveys in unrestrictea cr ea; are provided to insure the ef f1ctive control of radioactive natorial. Unu7ual conditions detected in the performnce cf cither a routine or spmial survey shall inmediately be brought to the attention of Radiological Centrols Ibrncp.;m . "ortable radiation nurvey instrwnts will b -

13.0 1869 120

.r>

. r., . . , '

i Janue, 13, 193

......cy- - ..,.;,...: -.

.,,1 -

e ..,1;3 .o u .. - u .....u.. .

.u., u.

....c., :.-

,.. t_. .r c,.,. c ..:- .c r

,,...u,c--..._ .....u..,

.a

\;: t will E.e cd.. crate? rn:e.dcr3y, to asc=a a ccr.sict . :t., relir? le end preuctmla respr.Ma to radi ation l=ve: . nccords of su're.:s ekl1 be r.cir.cair.c c.s file. [n Ivininistrative pro; ra:: '.ill b :

ex?. to verify :": cal 2.rc. tion of percor_.el cnd field ::cr.itori. ;

in--+:rn e n cy .

1869 121 .

14.0

r jp ;

f0

_f Of l Ph"jgghg';[hib/Jl' d UU k Rev ; ion l Janu ny 15, 1980 IMJ.cJ o 6 _ CanlrnLBi_lat5 mal _.DLTOW5.:

'1he policy of t'.etu cN1itan Diison Co:rpuny and Ccnaral Publi- U.a iitj a Cogcraticn i. not to have any cignificant intern:,1 e;:p;sure to per: cn el fraa radioacLivity av:.cciated with '"nrce L'.ile Island Unit 1. For personnel c>:poscS to radicactivi.ty during their i rk, this r.eans that no one ch^rld ut. nive frcw inteml reficartcWity more tlm cno tenth of their p Imittv3 emuel radintim e:caosw;e.

Crmtrals in ethn_ puts of this T'Z-1 Radiatie7 Protectic: Plan to n.inimice internal radioectivity, such as control of surface contarina-tion an0 control of uounds, are not repauted in this article. Tne following controls are to ninimize internal exposure from airborne retu.tuctivity :

1. Engir.eering controls and co:trols on personnel access shall be applied to th0 r.2ximum extent practicabic so that radioactive work does not incrcase the cou-ts of airbor .e radioactivity

- inhaled. Wen no other controle are practicable, respirators shall ba used. Those who ray need to use respiraters shull ba medically qualified, trained, tested for respirator eff.iciency, and roqualified in this respirator progra'n at least annually.

2. Itirborne radioactivity shall be rcasured regularly in areas where personnel ray be exposed. Continuous mnitoring repre-sent:five of air the parson is breathing shall be perforned to supplentat periodic ncasure~. ents during ucrk which has th potential to cause a worker to receive reasurable internal radioactivity.

Internal radioactivity shall be reasured at least annually in each person who worb, m an a wa requiring a radiation work permit; this 1869 122 15.0

Rev::irr, 1 Jan ar, iE, i9C incluJ_ cach parac '.'S urs :c7pirc. tory protection. Inter _1 .. .0-er'.tvj ty ciall be cr:a: ureci p:cgtly in each parson '.cho recen c:s rat: rc:ti co CC: *lL*:in - Llora Ca hiS hhi n Cr[ in CECh POrSC". '. hO 15 L '5pOctC'. C f ..i lir g LJ.f fiCi Orib rcKliOGOtiVity J. Crl'.L' : !..eafiuruhle ir. terr,nl rceior:=iv3 -. .

L2_n PtMO ; u_ ' Jnt Of 11/_2."".Q1 rc.'2 NCtivity abCyJC a lOVel T.OGr COL.:;~JId E } . ..: - l'. O FOVi(;!,lO tG d?tP ^ 2.".O tr.u C330 CT'd LG EEL A.It Art E.ir.1~ #_ iT' 4_--. O _em l }i lD tu y1

_g #j,jD ocJun hhq, a 6-1869 123

IMr r. ; ion 1 January 1 < , 1980 lat i r:le 7 - cor.t r ol of r.,tc:.c i.tve. r cr.t;. iru.tio ,

nadio2;tive srface car.ta rir.aticn shall be controlled in oda La ninimize p n.ib.1 irirchtien or in,yestion of radioact.ivity ar.d to n;. -

ndi e bui Ja up of radioactivity in th; envircnTat. .cesures to conte a radic e tivit;. and to ' nni:rsiz _ the nu%: ar.3 cxtent of areus ccatc: a .t : _

shall ho taken ir - -J ^r i n -in "r.i2 ' pctso"nel rcdiation expa.dr" to simpliEu subsc3.mt c:-rsorrel and area or facility decontamination, 2.r ?

to Ir.in On; ^ thm ne.M t o e]y on m

j contFTj natio 2 cl ottir.g.

Tn surf ace ce;1 twin-stj o: linlits for beta-ga:m.a activity shall be 1000 drc/100cn2 for loonc ccnterrir.ation a .5 0.1 nnfrz for total conte:r.-

ination. Yn: p':cferable re=.au cment techr.ique is with a pancake frishcr .

1000 chr L. spivalmt to 100 cm on the r.ancake frisker. For alpha activit,.

L're su-face conterrinatio:1 limit is 100 dpr'100:nn2 D phasis in planning, training and working shall be placed on ndni-rrlzir tiu r.unbers of occu rences and c cunts of radioactivity involved in occurrences of radioactive surface car.tcTination of a person's chin or of area.s nct control. led for radioactive surface conta rination. Pach r.uch occurrence shall be reviced in detail to detenrine hc*..' to correct deficiencies and irr. prove centrol of rcelcactivity.

e il c ,. 7-dldpngnnyp'ih& nL fd@

1869 124 17.0

D 1

@@p00{Qf.

D]D L _a v: t r, .$ 1M I?: 0- C - Co: .t r v . of T::dia chiv: :*v.ecicls Ir LCaiticn to the C.3finition of 10;.TR'_0, cny natoriel rin; c f me rat e ~ nsumi wit! e octu gr'ra sav:y .:'ter at 1 ir:ch c':cce31: r C .1 --! 'h or with nurfacc cr.zil.1 tion in cxeccs of the li : lits spacifin. ir Aru c' . 7 chS J : - M.%.5 n: rad i.cuct i s >. A rcaic.ctive nahrli.1 c'. :. trol syste n shull to entab2.i rNL to enstre rc,Jicact ive trateric. 2 :-

n d '_ lost a: 1:tisplv :0 it c lecction v:here pers:r el could u:Pr. <._:. .1; ' -:

eg; sex; to t u.uut tu, n_. tu prcm. une uncmtrollea spmna ci L. 2cm:;'

to areas where tr.e pt.ou c rr.gnt ~:>: afi~nted. '1his systc:a sh-ll c.;1 ;=

the follc< ting requirc.cntr :

1. 'the ntrier cf areca in which radicactive matcrials arc stcru .

shall be D.ini:.c.zcd.

2. Any new radicactive materia] storage area shall be appror:-d before use by the Mrlagar-Radiological Controls.

3. Tne nir:bers of radioactive items and the ancunt of radic-activity in storage shall be rainimized.

4. Rulicactive itcas sh:111 be identified as radioactive befera renoving ik.C fro.n a restricted area.

5. Radioactive raterials raroved fmn the Protected Securit; Area or rcrroved from a restricted area outside the Protected Sect;rity Arca shall be centrolled in acccrdance with an acco.:ntability proe- lure inich ensures the nuterials are not lost or igr :- _

crly har.dled during transfer or subject to unauthorized rcrc', _1.

Tnis accouM ability procedure shall require inventory of radioactive :raterials which ic: rain ouhide such creas.

6. F:ach inccclng or outgoirg shipre.nt of radioactive cat'.cri '

Shall be hc-d. led in strict cc=pliance with detailed writtcn procedurcs.

18.0 1869l125

Revir. ion 1 Janutry 16, 1900 Each e"x' O in Vimch redicactiv, nm-'e.1al ~,. e le Cf C300C<=tud nr chall br revig -

3 in dat311 to 6%'-~< m tb

' patOntial rea'etton a g:;.gg...q,

^C=C1 m.ic:ht u;-;". ' '-

-,,1,* .11y receiVO . to corrOct ~dn'i-D'CIC3,

~ " and to iTrrova contro,' o'c "occhive i.eteri.313, T

D l O AJh l) fiD U[. li,'d

[d,IdlI kiti!ud! L ~ij fL ul.L 1869 126 19.0

R' Vlai V g ,. . , .

,i:-

f ,..

[*." b.)' -* (irC " 2:1 i( l [ f'. r '.. O10. .' Cdk .~ I 19 r ." '

h lo3 OlC(iCi1 Co'~.I'.~l prOCrC*'l Can"Oh t2 StrCng and C150'~il-^; i~

}t_.. _c,._s...-

.~

_ s 4 g 4..h. n_ ')- - A. .i g.7. m m-~.. p; . { s-'g. ci [.m r

7 .-a T'. .,s a m'- m.

. , _.s.e..-4 .

g.i ,m. ...'. c - , . ~

.3..,.

m, _

.7;,,..

s ~_ } .... , , ,_ ; , v,,..m..

. , . _ .; .4 4..y _r..< . ,... _. ., ,1. : .y._ _i ic , ' gg, g. pg ,7 ., g-.n. . _ ,

tha at er.i La t ten f c. t'.'. .c cntir., f.ree 7.1.!O Ic1and Unit 1 repre s 5 --- r E.

v.. ..um .

s/ c.....L.-. u . . v~ ~y..__

, _.- s w.,- a . Lv

u. . . .

c.- +-w

.s :. . . . - . -. c _ %. .. . ,:_, e. ' e. ~. _s .- <--< . } r_<,. .*-- - ,m_1S .i. e.,. r,

, m. e. ..d. _ W,. _ .*_- .~ _- . - . _

....w..u. ... .. . t.e

., ,. . . .,. _, .u . _ . . .~._....w.~

. _ . , , , , . , , ~ a. ,r.q. . ,. , . 3.., rr. o ., ,

4. 5. . ,..._....;

c.n. - .e1.m. .: ...# e m ~. i . ._j . .r. t ia' ' . ' - u - s', .~ '. .r.,' i

. b ' l i 'v "..,O# t} Lo Pe'.A. d' ol "m y~i'9.c '. L".-.~.

D;U "tl:2nt to CVal'citO radiological ConditicQs and reco:r'end p!d76; i: T ,-

nea: . ires. '1D cssist the idar.cger-Radiol qical Ccatrols, a Radiolo:;'.rr.1

._ a .m, y.

. v_ ..s.,,..._ _ ..-

e v., . s. ..g< .. - c. .as .cn,.., u. -

In F.1cr..Ic2 s

.t .

Ist tim's when da :rls in. :r the nadiological Control iD2 crtren: m n n . .

sit :.1clenL1 h arcv. to rea.uire a tcr.oorery increase in sta.gr, ... c.ua3 .:.c.

cc,ntractor eracr;e1 ai]l be u.ead. Tnese personnel will be #c.'1; t i..:e-grated into the depar Tent under the cltrection of the F.anagr-R=dio'rp.r:1 .

Controlm Qualifications for t.he key radiological rcanagers in GC RO;nlit;q Guide 1.8, Rev. 1-R, (19Td will be n.ct 6s far as practicable. ~. ' .e re the cc.winatLon of strona : r.acer anc ex.cerlence in radiolocicel cer:r ._.s cannot precticcl0.'; 'ce obtt.inM2 in the uru person, credit for ex c.x .;

nt. clear rc.cer plant cperations, Ircintenr.ce, or engineering ', rill 'ce ;_ . .

One portion of th-2 'HII-1 radiolo;;ical controla program is ti.e TC.J.:J.

p ogram for person .e,1 radiation e:gostnes oc be as low as rcecen ,c.;

achievab}e. 'Ib acon:rplish this each engine:er involved with ' :.E-1 ' .c_-

to have radiological cngin:-rin; as part of his assigna.cnt. "' v x : cs-radiological cngine-ring functic's are .rarfonrcd in enc;inacrin. crc. :.: -

20.0 1869 127 .

Revision 1 January 16,19D0 1 Izdit logical Controla Departent.

c- --

9 r})(~)j 0 f !cm o , 7:

0;;' c, r.'

ll gin j

Jo h:o.a escny j!;.e.cn1 e,_eo;,l:,\!,! rd Pl n.o ~ 1869 ,

128

. t Vice res iden j Net-Ed I
i 1
Manager -
Radiological Controls t
..___.L....
,W n;il!.ra'iJe )
,o . :1 C1<:r i ea -
Supl ort
-~
.l_ . . . . . .- ! .
I Supervisor . _ .
? wl"? rv i s e-P.adiological l Radialenici Control Technicians _
l _ r gin ne <.:r i ng na
" ~ ~
Radiological ( . .
,d a a. .lologicci Control Technicians Encineers Foremen a
Radiological M ,
O Control e O !
Technicians b e M' a
- Figure 1 Tril-I Radiological Controls Depart. ment Organization ( J ,.,.,
N '
m FRD) - ,. .
rn-_ i. . . ,
c- ...~
539 (.n --.*
_C 223
[I.3 - . ,
b c.'. )
c.
APPENDIX 7B REPORT ON THE STATUS OF THE CORRECTIVE ACTION IN RESPONSE TO AUDIT BY NUS CORPORATION 1869 130 Am. 11
Report on the Status of Recommendations of NUS Audit Report of March 20, 1979 The following is the status as of December 21, 1979 on each of the recommendations of tne NUS Audit Report of March 20, 1979. Each recom-mendation is identified and reference is made to the specific paragraph and page number from the NUS Audit Report.
1. ... the pool of Health Physics / Chemistry Technicians be divided into separate groups for the two disciplines." (Ref: Section 2.1, page 2-2)
Status: Met-Ed intends to propose a split in job classification during the spring of 1980. The change would result in separate Radiation Protection Technicians and Chemistry Technicians. On January 1, 1980, Met-Ed has temporarily split the group by specifically assigning individuals to each of the disciplines until the contract negotiations are finalized.
Additionally, since July of 1979, Technicians have been rotated between the disciplines on a twelve week c3 ;1e, resulting in an increase in consis-tency in job performance and awareness of specific changes or anticipated operations within each discipline. The twelve week rotation results in each individual beginning his cycle during one of his training weeks, thereby allowing a refresher program to insure awareness of current procedures, maintenance and operations functions and plant conditions.
The recommendation has been incorporated to the extent possible and will, dependent on contract negotiations, be fully implemented.
2. " Designate one of the three Chemistry Foremen as a Chemistry Supervisor."
(Ref. Section 2.1, Page 2-3)
Status: Reorganization since the Unit-2 incident has changed the reporting requirements of the chemistry supervisors to others, outside the Health Physics organization. A separate chemistry organization has been developed for each Unit. This has resulted in the Health Physics Department being able to focus strictly on radiological concerns.
The recomendation has been incorporated in the reorganization.
3. "Re-arrange the Radwnte function so that the two foremen do not report directly to... (Supervisor Radiation Protection and Chemistry)".
Status: The Radwaste function is now the responsibility of the Operations Department. A Radwaste Supervisor has been appointed who reports directly to the Supervisor of Operations, Unit-1.
The recommendation has.been incorporated in the reorganization.
4. "A crew of personnel, such as Utility Workers, should be permanently assigned to Health Physics for the specific purpose of tool, equipment and respirator decontamination." (Ref: Section 2.2, page 2-4)
Status: A crew of Utility Workers will be assigned permanently to assist in radwaste and Health Physics functions. These individuals are not as yet in place, but will be available as the open Utility Worker positions are 1869 131
filled through hiring of new employees. Their function will not be limited to decontamination of tools and equipment. At present, a specific group of contractor personnel have been assigned to decontaminate tools and equipment A second group is responsible to decontaminate respirators in accordance with a respirator cleaning procedure.
The long-term solution to this recommendation has not been resolved.
5. "There is inmediate necessity for two additional full-time clerical personnel in the Radiation Protection / Chemistry Department." (Ref:
Section 2.3, page 2-6)
Status: The reorganization of the Unit-1 Radiological Controls Department since the Unit-2 incident has resulted in the addition of an Administrator and two c.lerical personnel in addition to the one existing clerical position.
The reconmendation has been incorporated into the reorganization.
6.
"One individual radiation should dosimetry be assigned program." (Ref: full-time responsibility)for Section 2.4, page 2-6 the Status: A full-time dosimetrist has been hired and in-place since July of 1979. He has developed a staff whose full-time function is the issuing, collection, processing of dosimeters and dosimetry record control.
Action has been implemented in accordance with the recommendation.
7. " Adjustments to the Technician staffing should include the provision to re-implement the training shif t on a continuing basis." (Ref: Section 3.1, page 3-2)
" Consideration should be given to including the Health Physics Foremen in selected phases of the continuing re-training' program " (Ref:
Section 3.1, page 3-2)
Status: A revised training program is in the final stages of development and will be implemented in January of 1980. The program text is complete and is designed to instruct both Technicians and Foremen. The training will be incorporated into the six week shift rotation schedule resulting in one week of training each six weeks for each technician and foreman. The con-tents of the program will include academic training in Health Physics prin-ciples, practical training including a review of site procedures, " hands on" laboratory and field sessions, and unusual events. Additional Technicians are being 6dded such that the training program can be conducted without direct impact on daily activities. Instructors in Health Physics are being added to the Training Department to insure that the burden of preparing and presenting the Health Physics program will fall on those supervisors having daily in-plant responsibilities.
Action will be completed in accordance with the recommendation.
8. " Personnel who are responsible for initiating RWP's should be clearly appraised of the necessity to properly describe their intended scope of work. All personnel must also understand that no changes in work scope which may invalidate the adequacy of RWP protective specifica-tions shall be made without prior Health Physics notification."
(Ref: Section 4.2, page 4-2) ]ggg}
Status: Training programs have placed emphasis on the need for complete description of work scope on RWP's such that the Radiological Controls personnel may provide adequate protective specifications. The technicians have been instructed to insure that they fully understand the work scope prior to approving the RWP's. The initiator is contacted when insuffi-cient information is provided. To assure that the radiological communi-cations via the RWP is effective, verbation compliance with the procedures is required and a Health Physics Audit Program, including audits from a Radiological Controls Inspector, the Radiological Engineering Group and contracted Health Physics Group, continuously monitors the work and the compliance with good Health Physics Practices. The recomendation has been incorporated in the Training Program, Audit Program and Health Physics Procedures.
9. "An immediate goal at TMI must be to insure that the Health Physics Technicians and Foremen are adequately trained. Once a reasonable degree of competence achieved, the decisions of these individuals must be given full support." (Ref: Section 3.2, page 3-3)
Status: The training program is discussed in item #7. To insure that decisions related to radiological control practices are based on com-plete information and in accordance and consistent with Health Physics Procedures and not reversed by other department supervisors, Radiological Control Technician Foremen have been assigned to shifts. The result has been a more direct involvement in off-shift operations and maintenance and a more complete awareness of all factors involved in the radiation protection aspects thereby allowing for greater consistency in applying radiation protection practices. The recommendation has been incorporated by means of improved training programs and shift Radiation Protection Technician Foremen.
10. "A concerated effort should be made by all parties to improve the day-to-day communications among all the members of the Health Physics organization." (Ref: Section 4.3, page 4 3)
Status: Comunications is recognized as the major factor in an effective radiological control program. The reorganization has resulted in a specific radiological control group without additional responsibilities such as chemistry and radioactive waste. This change has resulted in more direct communications within Health Physics. Additionally, a formalized shift turn-over followed by a shift briefing has been implemented with shift briefings being held to insure proper awareness of plant operations. Shift turnovers include a comunication with operations Shift Foreman. The Unit planning meetings, (i.e. Plan of the Day Meetings, Outage Planning Meetings are attended by the Manager-Radiation Protection or his designee, followed by a department staff meeting with the intent of improving overall communications. Appropriate information is presented to technicians during shift briefings by foremen. This recommendation has been incorperated into daily operations. i869 i33.
"A method should be implemented to ensure that all technicians are aware of each TCN and procedure revision." (Ref: Section 4.4, page 4-3) Status: A document review sheet system has been implemented which requires that technicians review, each procedure change and sign the review sheet. The recommendation has been incorporated by means of the document review sheet.
11. " Consideration should be given to making a complete re-evaluation of the TLD calibration method." (Ref: Section 5.1, page 5-2)
Status: Cnanges to the TLD calibration system are in place which reflect the use of a SR-90 Calibration as well as periodic quality assurance checks similar to those in effect before the Unit-2 incident. The creation of a dosimetry group as described in item #6 of this document has resulted in consistency of calibrations. The recommendation has been incorporated into existing procedures and the TLD calibration program is considered to be adequate.
12. "...a full-time individual with a job classification such as Technical Analyst should be assigned to perform surveillance of the overall radiation dosimetry program." (Ref: Section 5.2, page 5-3)
"To the extent which is achievable, all TLD's should be processed and read by the same technician, ..." (Ref: Section 5.3, page 5-3)
Status: A full-time dosimetrist has been hired and in-place since July of 1979. He has developed a staff whose full-time function is. the issuing, collection, processing of dosimeters and dosimetry record control. Action has been implemented in accordance with the recommendation.
13. "The continuous air monitoring system should be supplemented with pro-curement and use of portable CAMS." (Ref: Section 6.1, page 6-2)
"A total of five or six portable CAMS should be provided at TMI. These CAMS should be of the moving filter type and should have the capability to monitor both particulate and radioiodine activity." (Ref: Section 6.1, page 6-2)
Status: Approximately twelve portable air monitors (Eberline AMS-3) are available and in use at TMI-1. These monitors are of the fixed filter type which monitor for particulate activity. Portable monitors for rad-ioiodine activity have been evaluated. Due to the inherent problems with field monitoring for radioiodine, portable monitors have not been purchased. However sampling and lab analysis have been upgraded and procedures are in effect which define a complete air monitoring program for both particulate radiciodine, and gaseous activity. The upgraded procedures impose strict requirements for knowing the air activity prior to and during work functions. The recommendation has been incorporated through the purchasing of new equip-ment and upgrading of the air sampling procedural requirements. i869 , 134
14. "The schedules for cbtaining and analyzing air samples should be re-evaluated." (Ref: Section 6.2, page 6-2)
Status: The air sampling program has been re-evaluated and procedures upgraded. Presently, requirements exist for air sample results to be available within twelve (12) hours prior to beginning work in any area with potential for airborne activity. Continuous sampling is required for work in areas with air correntrations greater than 10% of MPC. The recommendation has been completed and steps taken to upgrade procedures.
15. "...the schedules for performing radiation and contamination surveys should be re-evaluated." (Ref: Section 6.3, page 6-3)
Status: Procedures which establish the survey requirements prior to work commencing in any area have been reviewed and upgraded. The result has been an increase in the number of areas surveyed for specific RWP's as well as increased frequencies of routine surveys for the areas routinely entered for maintenance or operations. The recommendation has been completed and procedures upgraded as required.
16. " Surveying and releasing of tools and equipment from the decon area should be performed by Health Physics Technicians. Procedures should provide guidance, techniques, criteria and limits for surveying and releasing these items." (Ref: Section 7.0, page 7-2)
Status: All tools and equipment which have been decontaminated are surveyed by Health Physics Technicians or other individuals specifically trained in that function. Procedures for control of contaminated material are in place. These procedures include monitoring requirements and limits for material leaving the controlled area as well as leaving the security " Protected Area." The recomendation has been incorporated into procedures. 1869 135.
8.0 SAFETY ANALYSIS
8.1 INTRODUCTION
Design changes affecting the acceptance criteria for the TMI-l FSAR safety analyses arise from several sources. First is the TMI-l " Order and Notice of Hearing" (Reference 19) which contains NRC staff recommendations that certain changes be made to the plant. This order encompasses recommendations made in NRC bulletins 79-05 A, B and C and the TMI-2 Lessons Learned Task Force NUREG-0578 (Reference 20). Most of the changes listed below are being made in response to this order. Prior to the TMI-2 accident, B6W 177 FA plants received orders requiring modifications to the high pressure injection system to accommo-date certain small break LOCA's. These changes are being evalu-ated as well. A third source of changes has originated f rom plant upgrades that Metropolitan Edison believes would improve plant performance. Some of these modifications were being evaluated prior to the TMI-2 accident on March 28, 1979. 8.2 AREAS OF INVESTIGATION The plant modifications which are being investigated are sum-marized below. They are grouped according to tneir origin. 8.2.1 Modifications Resulting from the August 9, 1979 Order
1. The reactor protection high pressure trip setpoint has been changed to 2300 psig from 2390 psig. This lower trip set-point in conjunction with the higher power operated relief valve (PORV) setpoint of 2450 psig results in a lower like-lihood of PORV operation.

2. A ccmplete loss of feedwater flow will initiate a reactor trip.

3. A turbine trip will initiate a reactor trip.
4 The emergency feedwater system will be modified before re-start to allow:
a. safety grade automatic unit ration of the steam and motor driven EFW pumps upon loss of all 4 reactor coolant pumps or a loss of both main feedwater pumps.

b. loading of EFW pumps on the diesel generators and dele-tion of the blackout start interlock.

c. alternate manual control for the EFW control valves.

d. negative feed to steam differential pressure.

e. loss of both main FW pumps.
1869 136 8-1 Am. 11
5. A long-term modification will provide safety grade actuation of the EFW pumps on the low cteam generator level. This is a long-term item since further engineering is required.
Plant safety therefore will be discussed with and without this feature. 8.2.2 Modification as Result of Order of May, 1978 Modifications to the high pressure injection system. The HPI injection lines have been cross connected to assure acceptable results from a break in a high pressure injection line. Cavitat-ing venturis have been added to provide the proper flow split in the event of an HPI line break. 8.2.3 Modification originating from within Met-Ed
1. Post accident instrument and valve operator availability will be improved by the addition of heat shrink tubing.

2. The switchover of the ECCS system suction supply from the borated water storage tank (BWST) will be accomplished automatically rather than by operator action.

3. The reactor building spray system will be modified to delete sodium thiosulfate. Sodium hydroxide will be retained.
This change will provide equal drawdown of the BWST and NaOH tanks for a large spectrum of single failures.
4. Letdown Flow will be automatically isolated af ter a reactor trip.

5. Cavitating venturis are being added to the emergency Feed-water system to prevent pump runout and to limit maximum flow to each OTSG, 8.2.4 I6E Bulletin 79-05C Met-Ed has committed to install an automatic reactor coolant pump trip to be initiated on a SFAS coincident with an indi-cation of a large (in excess of 10-20%) void fraction. (See section 2.1.2.5) 8.3 EFFECT OF CHANGES ON SAFETY ANALYSIS Following are summaries of the accidents listed in Table 8-1.
Table 8-1 indicates where FSAR analyses took credit for non-safety grade equipment, or where mitigation is dependent on a specific operating / emergency procedure or design margin. These conclusions will continue to be revised to account for plant design changes. The event description and mitigating equipment are for the plant design before modification. The modifications discussed in the previous sections were considered in the review of each accident. If a modification affected that analysis, then a note as to its safety significance was made under the " conclusions" section. 8-2 Am. 11 J
8.3.1 Rod Withdrawal from Startup (FSAR Section 14.1.2.1)
1. Description Uncontrolled reactivity excursion starting from a subcritical condition of 1% ok/k at hot standby.

2. Acceptance Criteria

1. Limit power to design overpower (112%)

11. RCS pressure not to exceed code allowable of 2750 psig.

3. Mitigation

1. RPS trip on high pressure for fast power rises.

11. Pressurizer code safety valves lift and peak pressure is limited to 2515 psia.
iii. Doppler coefficient provides a negative reactivity addition. 4 Conclusion The FSAR analysis still bounds the modified TMI-1 plant design. The RCS high pressure trip is lower and safety margins are increased. Since no credit was taken for opera-tion of the PORV, raising the valve setpoint does not change the analysis results. As discussed in Ref. 2, the PORV would lift for the worst case rod withdrawal accident which was analyzed in the FSAR. Nevertheless, the probability of occurrence has been decreased so that safety margins have been improved and lifting of the PORV is not likely for a broad spectrum of rod withdrawal accidents. 8.3.2 Rod Withdrawal at Power (FSAR Section 14.1.2.3)
1. Description Accidental withdrawal of a control rod group at normal rated power, without ICS control and a 1% shutdown margin.

2. Acceptance Criteria

1. Limit power to design overpower of 112%.
ii. RCS pressure not to exceed code allowable (2750 psig).
3. Mitigation

1. RPS trips on high pressure for slow transients and high neutron flux for fast transients.
ii. Doppler and moderator coef ficients provide negative reactivity addition. 8-3 Am. 11
4 Conclusions The FSAR analysis bounds the modified TMI-1 plant design. Lowering of the reactor trip setpoint *ncreases safety margins for this event. Credit was not taken for PORV operation. As discussed in Reference 2, some low worth rod withdrawals can result in PORV actuation. Nevertheless, the probability of such an occurrence has been greatly decreased by the changes in the PORV and high pressure trip setpoints. 8.3.3 Moderator Dilution Accident (FSAR Section 14.1.2.4)
1. Description Diluted makeup water is inadvertently added to the reactor coolant system at a rate of 500 gpm beginning at normal power. RCS boron concentration is at its highest initial value. The result is a reactivity insertion, increased power, pressure and temperature. The addition of one makeup tank volume of unborated water changes the shutdown margin by
.8% Ak/k-
2. Acceptance Criteria

1. Reactor power will be limited to less than the design overpower (112%).
ii. Reactor coolant system pressure will be limited to less than code allowable 2750 psig. iii. The minimum shutdown margin will be at least 1% Ak/ k -
3. Mitigation

1. Righ pressure or high temperature trip.
ii. Termination of deborated water to makeup tank on reactor trip. iii. Termination of makeup flow on high pressurizer level.
4. Conclusion The FSAR analysis bounds the modified TMI-l plant design.
Lowering of the high pressure trip setpoint increases the safety margins for this accident. Operation of the PORV was not assumed in the original analysis, and peak pressure is 2435 psia. Therefore, the PORV setpoint will not be reached during this transient. Reactor power is limited to 107.3%, and the final shutdown margin is greater than 1% Ak/k even with the most reactive rod stuck out of the core all of the acceptance criteria for this accident are met. 1869 139 8-4 Am. 11
8.3.4 Cold Water Addition (FSAR Section 14.1.2.5)
1. Description Startup of one or more idle reactor coolant pumps can cuase excess heat removal from the primary coolant system. This cooldown can cause positive reactivity insertions, which result in a power rise. The worst case event is the startup of two reactor coolant pumps from 50% power. A tripped rod worth of 1% ok/k is used in the analysis.

2. Acceptance Criteria

1. Limit overpower to less than the maximum design overpower (112%).

3. Mitigation

1. RPS trip on high pressure for slow power increases or power / flow mismatch for rapid power increases.
ii. RC pump / power monitor limits initial conditions under which event can occur.
4. Conclusion Lowering of reactor trip setpoint increases safety margins for th'is event. The FSAR analysis was performed without taking credit for PORV. Peak pressure did not exceed 2400 psia, hence the PORV will not lift during this event.
The FSAR analysis bounds the modified IMI-l plant design. 8.3.5 Loss of Coolant Flow (FSAR Section 14.1.2.6)
1. Description Fuel rods experience a limiting DNB transient when all four reactor coolant pumps trip on loss of offsite power or when one pump experiences a locked rotor resulting in an instan-taneous loss of flow. The loss of flow analysis is performed from 114% normal power, nominal reactor coolant pump flow, a
+2 F core inlet temperatu e error and a -65 psi error in pressure. Reactor trip delay is assumed to be 620 ms, and a 1% ok/k suberitical margin is assumed at hot standby. The event is analyzed past the time that the minimum DNBR occurs.
The locked rotor accident is performed from an initial power level of 102% power, with a rampdown in flow from 100% to 75% in 100 ms. Temperature ar d pressure were the same as for the loss of flow accident. Reactor trip delay is assumed to be 650 ms.
2. Acceptance Criteria

1. DNB is greater than 1.3 for a loss of coolant flow.

11. DNBR is greater than 1.0 for a locked rotor accident.
8-5 Am. 11
m, m e ACCIDENTS /TRAllSIENTS CONSIDERED FOR RE-ANALYSIS TABLE 8-1 Don't Consider Depend on Non-Safety Failure of Analyses Need Affected by Operator Equipment flon-Safety to be more Plant Changes Action Used _E Suipment Realistic Startup Accident X X . Dilution Accident X X X Cold Water X X loss of Coolant Flow X X Dropped Rod X Ios 3 of AC X X X Loss of Elec. Load X X X Steam Line Failure X X X Steam Cencrator X X Tube Failure Fuel llandling Accident X Rod withdrawal at X X Power Rod Ejection Accident X Small Break LOCA X X X X m EVENTS NOT ANALYZED IN FSAR e EIN Inadvertent X X X X
- In!tlation .s Loss of Feedwater X X - Feed Line Break X llP1 Line Break X X X loss of Offsite Power X X X
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TABLE 9.2-1
!!c s t ar t Modification Addit ional Loads Addit ional Load MCC or Items Title in KW Dist. Panel Bus Upgrade Decay Heat System Vent Valve A 0.7 DC Panel IE Vent Valve B 0.7 DC Panel IF Vibration Monitor 0.6 MCC IB Bus IS 2.1.1.1 Reactor Trip Neg. Vital Busses Each operator IA will be connected IB to it s associated IC channel ID 2.1.1.6.3.1 Incore Thermocouples No requirement 2.1.1.5 Containment Isolation less than 1 Later !
2.1.1.2 Valve Position Indicat ion 0.23 PNI ATB Inverter IE Computer Le s s than 0.5 Inverter IE 2.1.1.2 Power Operated Relief Valve Posit ion Indicat ion Less than 1 Later 2.1.1.4 H2 Recombiner A bus heater & blowers 45 MCC 1A IP ___ B bus heater & blowers 45 MCC IB IS CD space heater 0.8 PNL CT-E IS
& isolation valves 0.53 DC PNL lA e isolation valves 0.53 DC PNL IB - position indication 0.025 Swing ~
DC PNL IM W (PNLS lA and IB) LD 2.1.1.7 Emergency Feedwater Less than 1 Pump A VBA Inverter IA Pump B Vl'B Inverter B s Changes to HPI System to Accommodate Small Break LOCA 0.05 Inverter IE C
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11.0 TECHNICAL SPECIFICATIONS
11.1 INTRODUCTION
A considerable number of plant modifications are being accomplished in response to TMI-2 Lessons Learned (NUREG-0578), the TMI-I Order and Notice of Hearing - August 9, 1979, IE Bulletins, and Met-Ed's review of the TM;-2 accident. The hardware modifications are described in Sectior. 2.0 of this report. In some instances, Technical Specification changes are appropriate to account for systems and changes to systems that play a significant role in mitigating the consequences of accidents or transients. These new draf t Technical Specifications are discussed in Section 11.2. Formal requests to modif y the TMI-l Technical Specifications will be forwarded to the NRC at an appropriate time following PORC and GRC review of the Technical Specifications which must be completed before final submittal. 11.2 DRAFT TECHNICAL SPECIFICATIONS This section contains evaluations of those proposed modifications for which changes to the Technical Specifications will be requested. Draft Technical Specifications pages in the TMI-l format are con-tained in Appendix llA. 11.2.1 Reactor Trip on Loss of Feedwater or Turbine Trip Introduction Item B.5 of IE Bulletin 79-05B requires licenses to " Provide for NRC approval a design review and schedule for implementation of a safety grade automatic anticipatory reactor scram for loss of feed-water, turbine trip, or significant reduction in steam generator level. Item B.7 requires the submittal of Technical Specifications for design changes including those changes associated with Item B.S. Evaluation The reactor trips on loss of feedwater or turbine trip are designed as anticipatory reactor trips which respond to equipment situations which would produce significant primary system pressure transients. By tripping the reactor on the anticipation of a pressure transient, (1) the probability of an overpressure trip is reduced and (2) the challenge rate to the pressurizer power operated (Electromatic) re-lief valve and the pressurizer code safety valves is reduced. The design of the reactor trips on loss of feedwater or turbine trip incorporate a 2-out-of-4 logic, are fully testable, and meet the single failure criterion of IEEE-279. A description and evaluation of these reactor trips are contained in Section 2.1.1.1 of " Report in Response to NRC Staff Recommended Requirements for Restart of Three Mile Island Nuclear Station Unit 1." 1869 157 11-1 Am. 11
Since the reactor trip on loss of feedwater and the reactor trip on turbine trip are of the same safety grade as other Reactor Protection System trip functions, the Limiting Conditions for Operation in draf t Technical Specifications 3.5.1.1 and the Surveillance Requirements in draf t Technical Specification 4.1.1 (Items 45 and 46 of Table 4.1-1) have been chosen to be consistant with other Reactor Protection System functions, a " check" being required each shift, a " test" each month, and a " calibration" during each refueling outage. Bypass of the loss of feedwater trip below 10% power, and turbine trip below 20% power is per-mitted to allow normal reactor startup oeprations. Conclusions In conclusion, we have determined that, with regard to the reactor trip on loss-of-feedwater and the reactor trip on turbine trip: (1) The probability or consequences of accidents previously evaluated have not been increased. The proposed trips are anticipatory and have not been taken credit for in the accident analyses. The reactor trip on overpressure and the pressurizer Code Safety Valves remain the principal means of mitigating pressure transients. (2) No accidents, other than those previously considered, will be introduced. The additional reactor trips have been de-signed so as not to introduce additional failure modes into the Reactor Protection System or other safety equipment. Moreover, by anticipating significant pressure transients, the challenge rate to the overpressure trip and pressure relieving capacity has been reduced. (3) No safety margins have been reduced. Since the additional reactor trips scram the reactor on anticipation of signifi-cant pressure transients, the peak pressure associated with these transients can be expected to decrease which would re sult in an increase in safety margins as a result of postulated turbine trip or loss-of-feedwater transients. For these reasons, we conclude that implementation of the design for additional reactor trips, and adoption of associated Technical Specifications, does not involve unreviewed safety questions with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). 11.2.2 Position Indication of PORV and Safety Valves, Setpoints Introduction Section 2.1.3 of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," July 1979, contains NRC recommendations on installations of valve position indications for safety and relief valves. The guidance on safety / relief valve position is to, " Provide in the control room either -a reliable, direct position indication for the valves or a reliable flow in-dication device downstream of the valves." 11-2 , Am. 11
Evaluation 9 In response to the above recommendations, a system of indirect safety and relief valve position indications has been designed. The safety / relief valve position indication system, described in Section 2.1.1.2 of " Report in Response to NRC Staf f Recommended Requirements for Restart of Three Mile Island Nuclear Station Unit 1", consists of the following: (1) delta pressure taps at discharge piping elbows downstream of the safety and relief valves, and (2) acoustic monitors (accelerometers) mounted on the pressur-izer power operated relief valve. The above sensors are in addition to the tailpipe the rmocouples which are presently installed. Technical Specifications, Limiting Conditions for Operation and Surveillance Requirements, will be proposed for the safety / relief valve position instrumentation. The Limiting Conditions for Operation, contained in draft Technical Specification 3.5.5, requires the three delta pressure monitors and an accoustic monitor to be operable. The remedial action to be taken if one or more delta pressure monitors or the acoustic monitor becomes inoperable is to fix the monitors prior to startup follcwing the next cold shutdown. This requirement is based upon the following considerations: (1) The sensors for the delta pressure and accoustic monitors are located inside containment and would most likely require containment entry for repair or replacement. (2) Extended periods of operation without the delta pressure or acoustic monitors will not preclude the reactor operator detecting a leaking or stuck-open safety or relief valve. Several indications of safety or relief valve discharge flow include: (a) reactor coolant drain tank (pressurizer quench tank), level, temperature, and pressure, (b) safety and relief valve tailpipe temperature. In addition to delta pressure and accoustic monitor Technical Specifications, Limiting Conditions for Operation are proposed to establish " dual" setpoints for the pressurizer power-operated (Electromatic) relief valve. Draft Technical Specification 3.1.1.4a requires a setpoint of 2450 psig + 1% when the reactor coolant system temperature is above 275'F. The increase in this setpoint (from 2255 psig to 2450 psig) conforms to the NRC guidance contained in IE Bulletin No. 79-05B. No operability or remedial action is associated with draft Technical Specification 3.1.1.4a since the Electromatic relief valve is not safety-grade and thus is not credited in the safety analyses. Draft Technical Specifica-11-3 1869 159 , a.11
tion 3.1.1.4b provides the second of the " dual" setpoints as 485 psig + 1% when the reactor coolant temperature is below 275'F. The 485 psig setpoint is associated with considerations relating to overpressurization of the reactor coolant system under " cold" conditions. Overpressurization is addressed in our Marci. 13, 1978 submittal, Technical Specification Change Request No. 74. An additional aspect of IE Bulletin 79-05B, item B.3, involved a decrease in the RPS high pressure trip setpoint to reduce the challenge rate to the pressurizer Electromatic relief valve. The recommendations of B&W, in an April 20, 1979 communication, indicated that the RPS high pressure trip setpoint should be reduced to 2300 psig. The decrease in the RPS high pressure trip, from 2390 to 2300 psig is incorporated in TMI-l draft Technical Specification 2.3.1 (see Section 11.2.12). The Surveillance Requirements associated with the delta pressure and acoustic safety and relief valve monitors are contained in draft Technical Specification 4.1.1 (Items 47 and 48 in Table 4.1-1); these monitors are to be checked each shift and tested / calibrated each refueling period. The " check" and " test" surveil-lance need not be performed when TAVG is below 200*F since the reactor would be shutdown and this safety function unnecessary. Surveillance that is not performed due to a reactor shutdown greater than one month should be performed prior to startup. This requirement has been presented in draft Technical Specification 4.1.1 and made applicable to all surveillance requirements in Table 4.1-1. Accessibility considerations, as noted previously are significant with regard to the delta pressure and accoustic monitors and are the determining factor in the test / calibration interval. A Surveillance Requirement (setpoint test) for the pressurizer Electromatic relief valve is incorporated in draft Technical Specification 4.1.2; a refueling period interval has been se-lected to be consistent with the pressurizer safety valve surveil-lance interval. Conclusion In conclusion, with regard to the additional requirements for the delta pressure and acoustic monitors, and the setpoint requirements for the pressurizer Electromatic relief valve: (1) The probability or consequences of accidents, previously evaluated have not been increased. The requirement for operability and surveillance of the safety and relief valve monitors increases the probability that misoperation of the relief or safety valves will be detected, remedial action taken, and thus reduces the consequences associated with certain small-break loss-of-coolant accidents. 11-4 Am. 11 1869 160
(2) No accidents, other than those previously considered, will be introduced. The delta pressure and acoustic instrumentation have no automatic functions and therefore cannot change the course of any accident or transient; sufficient confirmatory information is available in the control room to detect improper functioning of these monitors. With regard to the pressurizer Electromatic relief valve, the requirement for periodic testing of the setpoint will enhance the availability of this equipment. (3) No safety margins have been reduced. Although the setpoint of the pressurizer Electromatic relief valve has been increased, no credit was taken for this equipment in the safety analysis. The decrease in the RPS high pressure trip setpoint will cause the reactor to trip earlier in the course of significant pressure transients and thus reduce the peak pressure during the transient. For the reasons presented above, implementation of the design changes associated with the delta pressure and acoustic monitors and associ-ated Technical Specifications , including those addressing the setpoint of the pressurizer Electromatic relief valve, do not involve any un-reviewed safety considerations with regard to the criteria of 10CFR Pa rt 50, Section 50.59(a)(2). 11.2.3 Emergency Power Supply Requirements - Pressurizer Heaters Introduction Section B.I.b of lE Bulletin 79-05B requires licensees of operating reactors to develop procedures and train personnel to
"... assure availability of adequate capacity of pressurizer heaters, for pressure control and maintain primary system pressure to satisfy the subcooling cirterion for natural circulation."
Section 2.1.1 of NUREG-0578, "TMI-2 Lessoas Learned Task Force Status Report and Short-Term Recommendations," goes further in that it recomaends that reactors, " Provide redundant emergency power for the minimum number of pressurizer heaters required to maintain natural circulation conditions in the event of loss of offsite power. Also provide emergency power to the control and motive power systems for the power-operated relief valves and associated block valves and to the pressurizer level indication instrument vhannels." Evaluation Section 2.1.3 of " Report in Response to NRC Staf f Reccomm. ded Requirements for Restart of Three Mile Island Nuclear Station Unit 1" describes design changes , and operator actions, that are required to supply 126 KW of pressurizer heaters from each of two independent engineered safeguard power sources in the event offsite power is lost. The manual transfer of po~er from the normal (balance of plant) to the back up (engineered safeguards) power source involves the use of a " Kirk Key" system that assures proper transfer of power from the normal to the back-up power 11-5 . Am. 11
source. In the event that an engineered safeguards actuation signal is received while the pressurizer heaters are powered from the diesel generators, the pressurizer heater loads are automatically shed from the diesel generators to prevent overloading of the diesels. Upon existence of an engineered safety features actuation signal (indicating a LOCA), primary system pressurization is no longer a consideration. To assure proper load shedding (breaker operation) of the pres-surizer heaters, from the diesel generators upon an engineered safety feature actuation signal, a test of the engineered safety features pressurizer heater supply breaker will be undertaken on a periodic basis. Technical Specification 4.6.1.b requires a test of the diesel generators, during each refueling shutdown, to determine proper automatic response under loss of normal AC power conditions concurrent with an engineered safety features actuation signal. Draft Technical Specification 4.6.1.b would also include a require-ment to confirm proper operation of the engineered safety features pressurizer heater supply breakers upon receipt of an engineered safety features actuation signal. With regard to the remaining requirements of NUREG-0578, Section 2.1.1, to supply emergency power capability for the PORV, the block valve, and pressurizer level instrumentation, Sections 2.1.1.3.2, 2.1.1.3.3, and 2.1.1.3.4 of the Restart Report indicate that these requirements are satisfied by existing equipment. Conclusion In conclusion, with regard to the provisions for transfer of pressurizer heater loads from normal to back-up power supplies: (1) The probability or consequences of accidents, previously evaluated, have not increased. Periodic testing of the engineered safety features pressurizer heater supply breaker will assure that, in the event that the pressurizer heaters are powered by the diesels generators when an engineered safety features actuation signal is received, the pressurizer heaters will be shed from the diesel generators supply busses. (2) No accidents, other than those previously considered, will be introduced. The manual transfer of the pressurizer heater supply loads, in the correct manner, is assured by the Kirk Key system. Periodic testing of the engineered safety features pressurizer heater supply breaker will prevent diesel overloading in the event that the pressurizer heaters are powered by the diesels-generators when an engineered safety features actuation signal is received. (3) No safety margins have been reduced. The availability of the pressurizer heaters on loss of offsite power provides additional assurance that the primary system subcooling margin can be maintained such that natural circulation will be enhanced. 11-6 }8h9 }b2 Am. 11
Based upon the above, we conclude that plant modifications necessary to allow manual transfer of selected pressurizer heater loads, f rom normal backup power sources, and adoption of associated Technical Specifications, do not involve an unreviewed safety question with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). 11.2.4 Post-LOCA Hydrogen Recombiner System Introduction Section 2.1.5 of "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," NUREG-0578, July 1979, contains a Task Force minority opinion that, "...all operating reactors, which do not already have the capability, be required to provide the capability to add, within a few days af ter an accident, a hydro-gen recombiner system for post-accident hydrogen control." Section 2.1.4 of " Report in Response to NRC Staff Recommended Requirements for Restart of Three Mile Island Nuclear Ststion Unit 1" contains a description and evaluation of design modifications that are re-quired to: (1) Install at TMI-l a hydrogen recombiner that was purchased for TMI-2, and (2) Provide structural, piping, and electrical facilities such that a second hydrogen recombiner can be installed after an accident, within the time period available before it is required to be operational. Evaluation Present NRC Policy, as reflected in " Standard Technical Specifica-tions for Babcock and Wilcox Pressurized Water Reactors", NUREG-0103, requires Technical Specifications for installed hydrogen recombiners. The proposed Technical Specifications for the hydrogen recombiners have been adopted f rom the Technical Specifications for TMI-2, specifically, TMI-2 Technical Specification 3/4 6.4.2. The hydro-gen recombiner Technical Specifications for TMI-2, based upon NRC's Baocock and Wilcox Standard Technical Specifications, contain: (1) Limiting Conditions for Operation requiring one operable hy-drogen recombiner during startup and power operation, (2) Surveillance Requirements for the following inservice inspection program: (a) A recombiner functional test once per 92 days, and (b) The following surveillance every 18 months: Channel calibration of recombiner instrumentation, visual examin-ation, heater functional test, and heater electrical circuit integrity. 11-7 Am. 11
In addition, the hydrogen recombiner is required to undergo surveillance prior to startup following on extended outage. The Limiting Condition for Operation for the hydrogen recombiner is incorporated in draft TMI Technical Specification 3.6, " Reactor Building"; the Surveillance Requirements for the hydrogen recom-biner are contained in TMI-l draf t Technical Specification 4.4.4,
" Hydrogen Recombiner System."
Conclusion In conclusion, with regard to the installed hydrogen recombiner and associated Technical Specifications: (1) The probability or consequences of accidents previously evaluatec have not increased. The use of hydrogen recombiners at TM1-1 would result in lower off-site doses, in the event of a LOCA, when compared with other post-accident hydrogen control tech-niques requiring containment venting.* (2) No accidents, other than those previously considered, will be int rod uc ed . The design and installation features of the hydro-gen recombiner are designed so as to preclude the compromising of containment integrity or other safety features. (3) No safety margins have been reduced. The hydrogen recombiner is a post-accident system that is not operated under normal con-ditions and thus is not involved in consideration of any safety margin. Based upon the above, we conclude that plant modifications needed for installation of the hydrogen recombiner(s), and associated Technical Specifications, do not involve any unreviewed safety questions with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). 11.2.5 Containment Isolation Modifications Introduction Section 6 of IE Bulletin 70-05A required licensees of operating B&W facilities to, " Review the containment isolation design and proce-dures, and prepare and implement all changes necessary to cause containment isolation of all lines whose isolation does not degrade core cooling capability upon automatic initiation of safety in-jection.' Section 2.1.4 of "TMI-2 Lessons Learned Task Force Status
*The recombiner cooling air is vented directly to the environment.
An evaluation involving failure of this cooling air system indicates that the resulting off-site doses are not significant (see Question 91, Supplement; Part 2, Restart Report) 11-8 Am. 11 1869 164
Report and Short-Term Recommendations ," NUREG-0578, July 1979, ex-panded the requirements of I&E Bulletin 79-05A, Section 6, as follows:
" Provide containment isolation on diverse signals in conformance with Section 6.2.4 of the Standard Review Plan, review isolation provisions for non-essential systems and revise as necessary, and modify containment isolation designs as necessary to eliminate the potential for inadvertent reopening upon reset of the isolation signal."
Evaluation Section 2.1.1.5 of " Report in Response to NRC Staff Recommended Requirements for Restart of Three Mile Island Nuclear Station Unit 1" provides the details and evaluation of a redesigned contain-ment isolation system with the following new features: (1) Containment isolation on reactor trip, (2) Containment isolation on 30 psig building pressure (3) Specific line isolation on high radiation With regard to the revised containment isolation design, this design meets the NRC's requirements in that: (1) The system initiates automatically on safety injection (IE Bulletin 79-05A) - The reactor trip signal is utilized to obtain a diverse isolation signal. Since the RPS trips the reactor on low pressure (1800 psig)* prior to the safety injection signal (1600 psig), an RPS trip signal on low pressure will always preceed a safety injection signal. The reactor trip signal, therefore, isolates the containment more quickly than a safety injection signal. (2) The system is diverse (NUREG-0578) - The redesigned containment isolation system provides containment isolation on the follow-ing signals: (a) Reactor trip (b) High radiation (individual line isolation) (c) Pipe break (individual line islolation) (d) The 1600 psig safety features actuation signal (e) The 30 psig contair. ment signal (f) The 4 psig containment signal (to be eventually removed)
Section 11.2.12 herein proposes an increase in the low pressure trip setpoint from 1800 psig to 1900 psig. ,.
11-9 1869 . 163 Am. 11
(3) Following isolation, lines should not be vulnerable to inad-verten reopening (NUREG-0576). Overriding the containment isolation signal does not open the containment isolation valves, deliberate operator action it required to reopen selected in-dividual valves. Draft Technical Specifications, described herein, provide Limiting Conditions for Operation and Surveillance Requirements for the additional containment isolation functions. Limiting Concitions for Operation for containment isolation on the RPS Trip and the 30 psig containment pressure have been incorporated into TMI-l draft Technical Specification 3.5.1 1 (Items 3.c and 3.d of Table 3.5-1). The minimum channel operability for containment isolation on RPS trip, and on Reactor Building 30 psig, have been chosen to be the same as the existing containment isolation functions; this would require a minimum of two channela to be operable or place the reactor in hot shutdown. With regard to Surveillance Require-ments, surveillance for containment isolation on RPS trip, and on Reactor Building 30 psig, have been incorporated into TMI-I draft Technical Specification 4.1.1 (Items 19.c and 19.d of Table 4.1-1) and chosen to be the same as the existing containment isolation system surveillance requirements the Reactor Building 4 psig signal; this requires a channel check each shift, testing each month, and calibration each refueling period.* A surveillance requirement for the manual containment isolation function has also been included (Item 19.b of Table 4.1-1) requiring a check each shift and a monthly test. Surveillance Requiements for line isolation on high radiation are presently provided for in Technical Specification 4.1.1 (Item 28, " Radiation Monitoring Systems," Table 4.1-1). The " check" and " test" surveillances are required to be performed only when containment integrity is required. This provision deletes surveillance requirements during extended outages when containment isolation may not be needed. Conclusion In conclusion, with regard to the revised containment isolation design and associated Technical Specifications: (1) The probability or consequences of accidents previously evaluated have not increased. The increased diversity of the containment isolation signals increases the proba- , bility and timeliness of containment isolation.
*This containment isolation function is not calibrated since no analog function is involved.
11-10 'l8/9 J 166 Am. 11
(2) No accidents, other than those previously considered, will be introduced. The revised containment isolation design does not in any way hamper the function of systems designed to mitigate the consequences of postulated accidents. Supurious initiation of any of the additi a al containment isolation signals would not isolate any components that would not also be isolated by a spurious initiation of the existing 4 psig building pressure signal. (3) No safety margins have been reduced. The plant safety features required to mitigate the consequences of postulated transients and accidents are not impacted by the revised containment isolation design. Based upon the above, we conclude that the modifications associated with the revised containment isolation design, and associated Technical Specifications, do not involve any unreviewed safety questions with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). 11.2.6 Instrumentation to Detect Inadeauate Core Cooling Introducticn Section 2.1.3b of Appendix A to "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations ," NUREG-0578, July 1979, requires that:
"...each PWR shall install a primary coolant saturation meter to provide on-line indication of coolant saturation condi-tion (SIC). Operator instruction as to use of this meter shall include consideration that is (SIC) not to be used exclusive of other related plant parameters."
Section 2.1.1.6 of " Report in Response to NRC Staff Recommended Requirements for Restart of Three Mile Island Nuclear Station Unit 1" contains a description of a saturation margin meter which is proposed for installation at TMI-1. Evaluation The saturation margin meter will display, in the control room, the margin between the actual primary plant temperature (T H) and the saturation temperature (Tsat) for the existing plant pressure. The Tsat will be computed using primary system pressure measure-ments and compared to the wide range Tg instrument reading. The temperature margin will be displayed in the control room. An alarm will be initiated if the margin falls below a pre-set value. Redundancy will be provided by computing Tsat margin independently 11-11 Am. 11 1869 167
for each reactor coolant loop. The lower tempert ure for each loop will automatically be selected for the compu itions. Satura-tion pressure margin is also computed in a similar manner so that the operator has the option of displaying the saturation margins in terms of temperature or pressure. The equipment used for tr.ese computations will be safety grade and seismically qualified. In addition, the plant computer, using the same inputs, will indepen-dently compute Psat and Tsat margin for logging, trending, and alarm. Draft Technical Specifications Limiting Conditions for Operation and Surveillance Requirements are presentad, herein for the saturation margin meter. Draft IMI-l Technical Specification 3.5.6, " Saturation Margin Meter," requires the saturation margin meter and alarm to be operable during start-up and power operation. If the saturation margin meter is not operable, the reactor operator is to have a procedure available for calculation of saturation temperature. This remedial action is appropriate since (1) no automatic actuations of safety features are associated with the saturation meter and (2) saturation temperature is easily calculated using reactor coolant system measurements and " steam ta ble s . " Surveillance Requirements for the saturation margin meter are incorporated in DHI-l draft Technical Specification 4.1.1 (Item 49 of Table 4.1-1). The proposed surveillance requires the saturation margin meter to be checked each shif t, tested monthly, and calibrated each refueling period. The " check" surveillance is only required when TAVG is above 200' such that this requirement is deleted during extended outages when the saturation margin meter is not needed. The proposed surveillance schedule is consistent with surveillance schedules for other safety grade instrumentation at TMI-l and is sufficient to assure reliable performance from *he saturation meter. Conclusion in conclusion, with regard to the saturation margin meter and associated Technical Specifications: (1) The probability or consequences of accidents previously evaluated have not inc re a se d . The saturation margin meter is not required for the prevention or mitigation of accidents, or transients, previously considered. (2) No accidents, other than those previously considered, will be introduced. No automatic actuations of safety features are associated with the saturation margin meter nor is the saturation margin meter capable of effecting any safety features. (3) No safety margins have been reduced. The saturation margin meter is not associated with any safety margins; both low pressure and high temperature RPS trips protect the reactor's thermal margins. Based upon the above, we conclude that the installation and use of the saturation margin meter, and the associated Technical Specifications , do not involve any unreviewed safety questions with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). 11-12 Am. 11
11.2.7 Emergency Feedwater System Modifications Introduction By letter dated June 28, 1979, Met-Ed presented NRC with recomman-daticas for modifications to TMI-l which would be completed pricc to restart of IMT-1. The June 28, 1979 letter recommended tne following changes to the emergency feedwater system and associated procedures:
1. Automatic initiation of the motor driven AFW pumps upon loss of both feedwater pumps or loss of four (4) Reactor Coolant Pumps.

2. Modification of the AFW control valves such that they fail open on loss of control air.

3. Automatic block loading of the motor driven AFW pumps on the emergency diesel generators.

4. Incorporation of AFW in the TMI-l Technical Specifications as specified in IE Bulletin 79-05A, item 8. Verification that Technical Specification requirements of AFW capacity are in accordance with the accident analysis will be conducted.

5. Provide indication in the control room of AFW flow to each Steam Generator.

6. Provide procedures and training to assure that AFW is available and properly applied when required. Procedures will identify the need to verify proper operation when AFW is initiated.

7. To assure that AFW will be aligned in a timely manner to inject on all AFW demand events when in the surveillance test mode, procedures will be implemented and training conducted to pro-vide an operator at the necessary location in communications with the control room during the sut seillance mode to carry out alignment changes necessary upon AFW demand events.

8. Design review and modifications, as necessary, will be conducted to provide control room annunciation for all auto start condi-tions of the AFW system.
On August 9, 1979 the NRC issued an " Order and Notice of Hearing" which addressed modifications to the TMI-1 facility. With regard to those changes proposed for the emergency feedwater system in the June 28, 1979 letter, the August 9, 1979 Ordtr directed that these changes should be made. A description and evaluation of changes to the emergency feedwater systra. invol nng equipment t modifications (items 1,2,3,5 and 8, r. ecst:1 bed above) are con-tained in Section 2.1.1.7 of "Repefe de Reshtnse to NRC Staff Recommended Requiremea*s for Re , s' .' Three Mile Island Nuclear Station Unit 1." Draft Technik q ',ications for the modified emergency feedwater system are :tecssseo herein. 11-13 Am. 11
Draft IMI-l Technical Specification 3.4 provides Limiting Conditions for Operation for the emergency feedwater system. Guidance on oper-ability of the emergency feedwater System is contained in IE Bulletin 7 9-C ' e, 7 tem 8, as follows:
" Prepare and implemant immediately procedures which assure that two independent steam generator auxiliary feadwater flow paths, each with 100% flow capacity, are operable at any time when heat removal from the primary system is through the steam generators. When two independent 100% capacity flow paths are not available, the capacity shall be restored within 72 hours or the plant shall be placed in a cooling mode which does not rely on steam generators for cooling within the next 12 hcurs.
When at least one 100% capacity flow path is not available, the reactor shall be made suberitical within one hour and the facil-ity placed in a shutdown cooling mode which does not rely on steam generators for cooling within 12 hours or at the maximum safe shutdown rate." The guidance contained in IE Bulletin 79-05A has been incorporated in Draft TMI-l Technical Specification 3.4.1 with the exception tha t the restoration time for the emergency feedwater system has been reduced from 72 hours to 48 hours as a result of subsequent requirements from the NRC. Existing Technical Specifications 3.4.3 and 3.4.6 have been rewritten to incorporate remedial action in the event that the condensate storage tanks and/or the main steam safety valves are inoperable. For both the condensate storage tank and the main steam safety valves, remedial action has been proposed that is consistent with NRC guidance as reflected in the B&W Standand Technical Specifications. With regard to surveillence requirements, draft Technical Specifica-tion 4.9, " Emergency Feedwater System," has been modified as follows: (1) Existing Technical Specificatic, 4.9.1.1 which requires testing of the emergency feedwater pumps every three months, as modified, would require pump testing every 31 daye and also require verification of specific pump flow values during the testing. The flow testing would be based the requirements of the ASME Boiler and Pressure Vessel Code, Section XI, Article IWP-3220, and would confirm that the emergency feedwater system can deliver at least 500 gpm to either steam generator flow path. . (2) Draft Technical Specification 4.9.1.2 would require valve line-up vs ification for valves in the flow paths of the emergency feedwater system, every 31 days. (3) Draf t Technical Specificat ion 4.9.1.3 would require a test, each 18 months, of automatic pump start logic and automatic valve lineup following an emergency feedwater actuation signal. In addition, the operability of the manual control valve station would be verified. 11-14 1869 170 ^=- 11
(4) Item 10F of NRC's October 26, 1979 letter to Mr. R. C. Arnold requires a, "... Technical Specifications to assure that prior to plant startup following an extended cold shutdown, a flow test would be performea to verify the normal flow path from the primary EFW system water source to the steam generators. The flow test should be conducted with EFW system valves in their normal alignment." This test is incorporated in draft Technical Specification 4.9.1.5 where the term " extended cold shutdown" is interpreted as "a cold shutdown of longer than 30 days' duration." (5) Existing Technical Specification 4.1.2 (Table 4.1-2), would require a functional test of the Backup Instrument Air Supply System (backup air supply for the emergency feedwater control valves), every refueling period. (6) Existing Technical Specification 4.1.1 (Item 50 in Table 4.1-1), as modified, would require a check each shif t, monthly testing, and calibration each refueling period, for the emergency feedwater flow instrumentetion. The " check" and " test" surveillances would not be required when TAVG is less than 200*F since the reactor would be shutdown anc this safety function not needed. (7) Existing Technical Specification 4.5.1.1, as modified would incorporate the motor driven feedwater pumps into the list of equipment whose operation is verified during the testing of the emergency diesel generators. In this case , only operation of the interlock would be verified since the pumps do not actually start on loss of AC power (the actual start signal is on loss of main feedwater or loss of reactor coolant pumps.) Conclusion In conclusion, with regard to the modifictions to the emergency feedwater systems and associated Technical Specifications: (1) The probability or consequences of accidents previously evaluated have not increased. The more conservative Limiting Condition for Operation and Surveillance Requirements for the emergency feedwater system provide increased assurance that the system will operate properly, when required. (2) No accidents, other than those previ:usly considered , will be introduced. The modifications to the emergency feedwater system could only iffect the non-operation or spurious operation of the system; both of these conditions have been previously evaluated. The only aspect of the emergency feedwater system modification with the potential for ef fecting other systems involves the loading of the motor driven feedwater pumps on the emergency diesel generators. An anslysis of the diesel generator loading indicates that the 11-15 Am. 11 1869 _ 17\
maximum load, with the emergency feedwater pumps is 2817 Kw compared to the 2000 hour rating of 3000 Kw. The proper diesel generator loading sequence with the emergency feedwater pumps will be verified prior to startup and every 18 months thereafter. Other aspects of the emergency feedwater system will be tested prior to startup, and periodically thereaf ter. (3) No safety margins have been reduced. The modifications to the emergency feedwater system did not involve any changes which resulted in a decrease in capacity of this system to perform its designed function. Based upon the above, we conclude that the modifications to the emergency feedwater system, and associated Technical Specifications, do not involve any unreviewed safety questions with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). 11.2.8 Post Accident Monitoring Introduction Section 2.1.8 of, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," NUREG-0578, July 1979 makes the following recommendations with regard to the increased range of radiation monitors:
" Provide high range radiation monitors for noble gases in plant effluent lines and a high-range radiation monitor in the contain-ment. Provide instrumentation for moitoring effluent releases lines capable of measuring and identifying radiciodine and par-ticulate radioactive ef fluents under accident conditions." In addition, the NRC recommended that facilities " Provide in.ntrumen-tation for accurately determining in plant airborne radioiodine concentratians to minimize the need for unnecessary use of res-piratory equipment. In an August 13, 1979 ACRS memorandum to the NRC, the ACRS recommended the following additional post-accident instrumentation: (1) containment pressure, (2) containment water level, and (3) on-line monitoring of hydrogen concentration in the containment.
The post-accident monitoring instruments to be installed at TMI-l are responsive to the recommendations of the NRC and the ACRS. Evaluation Sec tion 2.1.2.1 of "Repor t in Response to NRC Staf f Recommended Requirements for Restart of Three Mile Island Nuclear Station Unit 1" describes the post-accident monitoring instrumentation to be installed at TMI-1. The post-accident instrumentation, in conformance with Regulatory Guide 1.97, consists of the following: (1) Containment Pressure - the range will be - 5 psig to three times the containment design pressure; 11-16 Am. 11 1869 172
(2) Containment Water Level - a narrow range monitor will measure containment sump level while the wide range monitor will measure from the bottom of the containment to a 10 ft. level; (3) Containment Hydrogen Indication - continuous reading of the concentration of hydrogen in the containment, from 0 to 10%, will be available in the control room; (4) High Range Containment Radiation Monitor - two monitors with a range to 107 R/hr will be provided; (5) High Range Effluent Monitors: (a) Undiluted Containment Exhaust - 105 C1/cc (b) Diluted Containment Exhaust - 104 Ci/cc (c) Auxiliary and Fuel Handling Building Exhaust 103 Ci/cc (d) Condenser Off Gas - 102 C1/cc (e) High Range Effluent Radio Iodine 6 Particulate Sampling and Analysis - silver zeolite cartridges. Although the above instrumentation does not actuate safety equipment, nor is it required by safety analyses, it is appropriate to provide Surveillance Requirements to assure reliable post-accident performance of the instrumentation. Surveillance Requirements for post accident monitoring instrumentation is incorporated into TMI-l Draft Technical Specification 4.1.1 (Table 4.1-1) as follows: (1) Item 13 of Table 4.1-1, "High Reactor Building Pressure," is provided with a footnote to include the post-accident instru-mentation in the existing containment pressure instrumentation surveillance program; (2) Item 28 of Table 4.1-1, " Radiation Monitoring Systems," is provided with a footnote to include the post-accident instru-mentation, described in item (5)(a) thru (5)(d) above, in the existing radiation monitor system surveillance program; (3) I t e. 37 of Table 4.1-1 " Reactor Building Sump Level" has been changed to " Reactor Building Sump and Containment Level." A foot-note has also been added to include the post-accident intrumentation in the sump level instrument surveillance program. (4) A new item 52, " Reactor Building Hydrogen Concentration," has been added to address Surveillance Requirements for the reactor building hydrogen concentration instrumentation. The
" check" and " test" surveillance need not be performed when TAVG is less than 200*F since the reactor would be shutdown and this safety function not needed.
1869 I73 11-17 Am. 11
Conclusion With regard to TMI-I post-accident monitoring instrumentation, and associated Technical Specifications, since the instrumentation does not actuate safety equipment, nor is it required by the safety analysis: (1) The probability or consequencts of accidents previously evalu-ated have not increased. (2) No accidents of a type, not previously evaluated, will occur, and (3) No safety margins have been reduced. Based upon the above, we conclude that the post-accident monitoring instrumentation, and associated Technical Specifications, do not involve any unreviewed safety issues with regard to the criteria of 10CFR. Part 50, Section 50.59 (a)(2). 11.2.9 Reactor Coolant Pump Trip on Coincident ESFAS and Coolant Voiding Introduction The IE Bulletin Nos. 79-05C and 79-06C, July 26, 1979 states that, "Recent preliminary calculations performed by Babcock & Wilcox, Westinghouse and Combustion Engineering indicate that, for a certain spectrum of small breaks in the reactor coolant system, continued operation of the RCPs can increase the mass lost through the break and prolong or aggravate the uncovering of the reactor core. The damage to the reactor core at TMI-2 followed tripping of the last operating RCP, when two phase fluid was being pumped through the reactor coolant system. It is our current understanding that all three of the nuclear steam system suppliers for PWRs now agree that an acceptable action under LOCA symptoms is to trip all oper-ating RCPs immediately, before significant voiding in the reactor coolant system occurs." With regard to reactor coolant pump trip, lE Bulletin Nos. 79-05C and 79-06C recommends the following long-term action:
" Propose and submit a design which will assure automatic tripping of the operating RCPs under all circumstances in which this action may be needed."
Section 2.1.2.5 of " Report in Response to NRC Staff Recommended Response to NRC Staf f Recommended Requirements for Restart of Three Mile Island Nuclear Station Unit 1" contains a description of the reactor coolant pump trip that is proposed for TMI-1. i869 174 11-18 Am. 11
Evaluation The logic for the reactor coolant pump trip receives inputs f rom the High Pressure Injection (HPI) signal f rom the ESFAS, and redundant pump current sensors from each of four reactor coolant pumps. The pump trip will occur on concurrent HPI and low current in two of the four reactor coolant pumps. Non-operating reactor coolant pumps have effectively tripped current sensors. With only two reactor coolant pumps operating, therefore,these pumps will trip on HPI. The output of the trip logic provides a trip signal to each reactor coolant pump. Limiting Conditions for Operation and Surviellance Requirements for the Reactor Coolant Pump Trip are addressed below. Limiting Conditions for Operation for the Reactor Coolant Pump Trip are presented in TMI-l draft Technical Specification 3.5.7. The draft Technical Specification requires that an HPI actuation channel and one pump current channel from each operable pump be available to the reactor coolant pump trip logic. In the event that the reactor coolant pump trip is inoperable, hot shutdown must be achieved within 24 hours. The draft Technical Specifica-tion only allows continued reactor operation if the pump trip is in a condition which assures reliable operation. The remedial action is specified to allow a reasonable time to restore the reactor coolant pump trip to operability or achieve an orderly shutdown. The Surveillance Requirement for the reactor coolant pump trip is contained in TMI-l draft Technical Specification 4.1.1 (Table 4.1-1). A new item, number 51, proposes a pump trip channel check each shift, a test each month, and a calibration each refueling period. The draft Surveillance Requirement for the pump trip is consistent with the surveillance for other safety instrumentation channels. The " check" and " test" surveillances need not be performed when TAVG is less than 200*F since the reactor is shut down and this safety function is not needed.
Conclusion With regard to the reactor coolant pump trip, the logic is designed to provide high assurance that the reactor coolant pumps will be triped when required. Any single failure within the reactor coolant pump trip logic will result in only a single reactor coolant pump being tripped. The draf t Limiting Condition for Operation for the reactor coolant pump trip prevents extended reactor operation if the reactor coolant pump trip is significantly degraded. The draft Surveillance Requirement for the reactor coolant pump trip provides assurance of reliable operation.
11.2.10 TMI-1/TMI-2 Separation Introduction Item II.4 of the NRC's August 9, 1979 " Order and Notice of Hearing," requires that, "The licensee shall demonstrate that decontamination and/or restoration operations at TMI-2 will not affect safe operations at TMI-1. The licensee shall provide separation and/or isolation of TNI 1/2 radioactive liquid transfer lines. Fuel handling areas, vent-ilation systems, and sampling lines. Effluent monitoring instruments shall have the capability of discriminating between effluents resulting from Unit 1 or Unit 2 operations." 11-19 Am. 11 1869 175 .
Section 7.2 of " Report in Response to NRC Staf f Recommended Require-ments for Restart of Three Mile Island Nuclear Station Unit 1" de-scribes a plan to separate TMI-1/TMI-2 interfaces that have the potential of transferring significant quantities of contamination as a result of restoration activities at TMI-2. Evaluation The two major pathways for potential transfer of contamination from TMI-2 to TMI-l are the waste management interconnections and the common air space of the Fuel Handling Building. The following E4I-1/TMI-2 waste management interfaces have been identified: (1) Unit 2 Reactor Coolant Bleed Holdup Tank - Unit 1 Reactor Coolant Waste Evaporator. (2) Unit 1 Miscellaneous Waste Evaporator - Unit 2 Evaporator Condensate Test Tank. (3) Unit 2 Nuetralizer Tanks, Contaminated Drain Tanks, Reactor Coolant Bleed Holdup Tanks, Auxiliary Building Sump Tanks and Miscellaneous Waste Holdup Tanks - Unit 1 Liquid Waste Disposal System. (4 ) Unit 1 Evaporator Concentrate - Unit 2 Evaporator Concentrate. (5) Unit i Spent Ion Exchange Resin - Unit 2 Spent Ion Exchange Resin. Draft TMI-l Technical Specification 4.1.2 (Table 4.1-2, Item 13) requires the isolation devices (valves, blank flanges, etc.) on the above tielines to be verified to be isolated, by visual in-spection, on a monthly basis. Draft TMI-l Technical Specification 3.19 requires that, if an isolation device is found to be open with-out prior NRC authorization, a " Thirty Day Written Report" must be prepared per TMI-1 draft Technical Specification 6.9.2.B(5). In addition, TMI-l draft Technical Specification 3.19.2 requires NRC approval prior to creation of additional B4I-1/TMI-2 system inter-ties that can transfer potentially significant quantities of con-tamina tion. With regard to the separation of the air space in the Fuel Handling Building, the details of this modification have not been finalized. Additional evaluations and preparation of draft Technical Specifica-tions will be undertaken, if appropriate, following finalization of the design details of the Fuel Handling Building isolation system. Conclusion The draf t IMI-l Technical Specifications 3.19 and 4.1.2 for the TMI-1/ TMI-2 interties provide assurance that: (1) System intenties that could potentially transfer significant quantities of contamination from TMI-l to TMI-2 will remain closed. 1869 176 11-20 Am. 11
(2) If permission is received from the NRC to open system interties, these interties will be used in accordance with plant procedures. (3) No new system interties, with the potential for transferring significant quantities of contamination from TMI-2 to TMI-1, will be created without prior NRC approval. The above controls limit releases from TMI-l to materiaAs under control at TMI-l and thus to previously evaluated quantities and concentrations of contamination. 11.2.11 Low Reactor Coolant System Pressure Channel for HPI/LPI Initiation Introduction The Low Reactor Coolant System Pressure Channel setpoint, which is used as input to the ESFAS logic, is determined based on a generic LOCA analysis. The generic LOCA analysis for TMI, referenced as "ECCS Analysis of B&W's 177-FA Lowered-Loop NSS," BAW-10103, has referenced the Low Reactor Coolant System Pressure setpoint as 1600 psig compared with the Technical Specification value of 1500 psig. The setpoint a tually used in the BAW-10103 calculations, however, was 1350 psi,. Evaluation The TMI-l Technical Soecification 3.5.3.1, " Engineered Safeguards Protection System Ae uation Setpoints," requires the Low Reactor Coolant System Pressure HPI/LPI initiation setpoint to be > 1500 psig. Draft TMI-l Technical Specification 3.5.3.1 would require the Low Reactor Coolant System Pressure HPI/LPI initiation setpoint to be raised to > 1600 psig. In the event of a LOCA, the only impact of the 100 psig increase in the minimum Low Reactor Coolant System Pressure setpoint would be to initiate actions, based on this signal, at an earlier time in the accident (e.g. , in conjunc-tion with the 4 psig High Reactor Building Pressure, both HPI and LPCI pumps would start earlier in the accident.) Conclusion With regard to the 100 psig increase in the minimum Low Reactor Coolant System Pressure HPI/LPI initiation setpoint: (1) The probability or consequences of accidents previously evalu-ated have not increased. The potential initiation of engineered safety feature equipment at an earlier time in a LOCA is not expected to have a significant impact on peak clad temperature and other LOCA limits (any changes would be expected to be in direction of a less severe accident). (2) No accidents of a type not previously evaluated will occur. The proposed change in the Low Reactor Coolant System Pressure Setpoint would have only a small impact on the severity of the LOCA, in the conservative direction, rather than change the nature of the accident. 1869 177 11-21 Am. 11
(3) No safety margins have been reduced. The applicable LOCA calculations continue to be those for which the Low Reactor Coolant System Pressure HPI/LPI initiation setpoint is 1350 psig; operationally, raising the minimum setpoint to 1600 psig would slightly increase the LOCA margins. Based upon the above, we conclude that raising the minimum Low Reactor Coolant Pressure setpoint from 1500 psig to 1600 psig does not involve any unreviewed safety questions with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). 11.2.12 Raising the Reactor Protection System (RPS) Trip Setpoint from 1800 psig to 1900 psig Introduction The TMI-l Technical Specification 2.3.1 (Table 2.3-1, Figure 2.3-1) provides a value of 1800 psig for the RPS Low Reactor Coolant Pressure trip setpoint. The B&W generic ECCS analysis, "ECCS Anal-ysis of B&W's 177-FA Lowered Loop NSS," BAW-10103, Rev. 2, April 1976, referenced a value of 1900 psig for the Low Reactor Coolant Pressure Trip setpoint. Draft Technical Specification 2.3.1 would increase the Low Reactor Coolant System Pressure Trip Setpoint from 1800 psig to 1900 psig. Evaluation The principal reason for the Low Reactor Coolant System trip set-point is to maintain thermal margins for the fuel by preventing the minimum DNB ratio from decreasing below the safety limit of 1.3; the transient analysis for TMI-l is based on an 1800 psig Low Reactor Coolant System trip setpoint. The Low Reactor Coolant System trip setpoint is also credited in the ECCS analysis since a reactor trip is part of the assumed LOCA scenario. By increasing the Low Reactor Coolant System Pressure setpoint from 1800 psig to 1900 psig, the reactor would trip earlier in t he LOCA scenario and thus the decay heat would be slightly less when the ECCS functions. Increasing the Low Reactor Coolant System trip setpoint also has the effect of increasing the margin to DNB following a trip on low pressure; the reactor would trip earlier on low pressure and thus the final minimum DNB would be higher (more conservative) than if the reactor tripped at 1800 psig. Conclusion With regard to increasing the Low Reactor Coolant System trip se tpoint from 1800 psig to 1900 psig: (1) The probability or consequences of accidents previously con-sidered have not increased. For any accident that involves a pressure decrease, the reactor will trip earlier in the trans-ient and thus the result of the accident will be more conser-vative. I869 178 11-22 Am. 11
(2) No accident of a type not previously evaluated, will occur. The increasing of the Low Reactor Coolant System trip setpoint will not have any effect other than tripping the reactor at an earlier time in pressure reduction transients. (3) No safety margins have been decreased. It is expected that for pressure reduction transients, DNB following the reactor trip will be higher (more conservative) and for the IDCA, the peak clad temperature and other system parameters will be more favorable. Based upon the above, we conclude that increasing the Low Reactor Coolant System trip setpoint from 1800 psig to 1900 psig does not involve unreviewed safety cuestions with regard to the criteria of 10CFR Part 50, Section 50.59(a)(2). 11-23 Am. 11 1869 179.
I1A. DRAFT TECHNICAL SPECIFICATIONS The following sections contain the draft 'IMI-1 Technical Specifications that are referenced in Section 11. 1869 \80_
Draft Technical Specifications Corresponding to Section 11.2.1 1869 I81 .
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reactor r oolant temperature instrument channels, f our reactor coolant flow instrument channels, four reactor coolant pressure instrument channels, four pressure-temperature instrument channels , f our flux-imbalance flow instrument channels, f our power-number of pumps instrument channels, and four high reactor building pressure instrument channels. The reactor trip, on loss of feedwater, may be bypassed below 10% reactor power. The reactor trip, on turbine trip, may be bypassed below 20% reactor power. The safety features actuation system must have two analog channels functioning correctly prior to startup. Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column "B" (Table 3.5-1). This is in agreement with redundancy and single failure criteria of IEEE 279 as describec in FSAR Section 7. There are four reactor protection channels. Normal trip logic is two out of four. Required trip logic for the power range instrumentation channels is two out of three. Minimum trip logic on other instrumentation channels is one out of two. The four reactor protection channels were provided with key operated bypass switches to allow on-line testing or maintenance on only one channel at a time during power operation. Each channel is provided alarm and lights to indicate when that channel is by passed. There will be one reactor protection system bypass switch key permitted in the control room. Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypas switch is being used. Power is normally supplied to the control rod drive mechanisms from two separate parallel 460 volt sources. Redundant trip devices are employed in each of these sources. If any one of these trip devices fails in the untripped state on-line repairs to the failed device, when practical, will be made, and the remaining trip devices will be tested. Eight hours is ample time to test the remaining trip devices and in many cases make on-line repairs. REFERENCE FSAR, Section 7.1 3-28
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Channels subject only to " drift" errors induced within the instrumen-tation itself can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at the intervals of each refueling period. Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures. Thus, minimum calibration frequencies set forth are considered acceptable. Testing On-line testing of reactor protection channels is required once every four weeks on a rotational or perfectly staggered basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood s# the same systematic test errors being introduced into each redundant channel. The rotation schedule for the reactor protection channels is as follows: Channels A, B, C, & D Before Startup, when shutdown greater than 24 hours Channel A One Week After Startup Channel B Two Weeks After Ltartup Channel C Three Weeks After Startup Channel D Four Weeks After Startup The reactor protection system instrumentation test cycle is continued with one channel's instrumentation tested each week. Upon detection of a failure that prevents trip action in a channel, the instrumentation asso-ciated with the protection parameter failure will be tested in the remaining channels. If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting. The protection channels coincidence logic and control rod drive trip breakers are trip tested every four weeks. The trip test checks all logic combinations and is to be performed on a rotational basis. The logic and breakers of the four protection channels shall be trip tested prior to startup when the reactor has been shutdown for greater than 24 hours. Discovery of a failure that prevents trip action requires the testing of the instrumentation associated with the protection parameter failure in the remaining channels. 4-2 1869 185
For purposes of surveillance, reactor trip on loss of feedwater and reactor trip on turbine trip are considered reactor protection system channels. The equipment testing and system sampling f requencies specified in Table 4.1-2 and Table 4.1-3 are considered adequate to maintain the equipment and systems in a safe operational status. REFERENCE (1) FSAR, Section 7.1.2.3.4 4-2a jbibh bb
Draft Technical Specifications Corresponding to Section 11.2.2 4 1869 187
3. LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM 3.1.1 Operational Components Applicability Applies to the operating status of reactor coolant system components.
Objective To specify those limiting conditions for operation af reactor coolant system components which must be met to ensure safe reactor operations. Specification 3.1.1.1 Reactor Coolant Pumps
a. Pump combinations permissible for given power levels shall be as shown in Specification Table 2.3.1.

b. Power operation with one idle reactor coolant pump in each looo shall be restricted to 24 hours. If the reactor is not returned to an acceptable RC pump operating combination at the end of the 24-hour period, the reactor shall be in a hot shutdown condition within the next 12 hours.

c. The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.
3.1.1.2 Steam Generator
a. Both steam generators shall be operable whenever the reactor coolant average temperature is above 250*F.
3.1.1. 3 Pressurizer Safety Valves
a. The reactor shall not remain critical unless both pressurizer code safety valves are operable with a lift setting of 2500 psig 11%.

b. When the reactor is suberitical, at least one pressurizer code safety valve shall be operable if all reactor coolant sy r *.em openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code, Section III.
3.1.1. 4 Pressurizer Electromatic Relief Valve
a. The setpoint for the pressurizer Electromatic relief valve shall be 2450 psig 11% when reactor coolant system temperature is greater than 275*F.

b. The setpoint for the pressurizer Electromatic relief valve shall be reset within one hour to 485 psig +1% when reactor T
coolant system temperature is less than 275 F. 1869 188 .
Bases The limitation on power operation with one idle RC pump in each loop has been imposed since the ECCS cooling performance has not been cel-cuiated in accordance with the Final Acceptance Criteria requirements specifically for this mode of reactor operation. A time perioc of 24 hours is allowed for operation with one idle RC pump in each loop to effect repairs of the idle pump (s) and to return the reactor to an acceptable combination of operating RC pumps. The 24 hours for this mode of operation is acceptable since this mode is expected to have considerable margin for the peak cladding temperature limit and since the likelihood of a LOCA within the 24-hour period is considered very remote. A reactor coolant pump or decay heat removal pump is required to be in operation before the boron concentration is reduced by cilution with makeup water. Either pump will provide mixing which will pre-vent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One decay heat removal pump will circulate the equivalent of the reactor coolant system volume in one-half hour or less. The decay heat removal system suction piping is designed for 300*F and 370 psig; thus, the system can remove decay heat when the reactor coolant system is below this temperature. (2,3) Both steam generators must be operable before heatup of the Reactor Coolant System to insure system integrity against leakage under normal and transient conditions. Only one steam generator is re-quired for decay heat removal purposes. One pressurizer code safety valve is capable of preventing over-pressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, anc reactor decay heat. (4) Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpressure for a rod withdrawal or feedwater line break accidents. (5) The pressurizer code safety valve lift setpoint shall be se t at 2500 psig +1% allowance for error and each valve shall be capable of relieving 280,800 lb/h of saturated steam at a pressure not greater than three percent above the set pressure. The purpose of the pressurizer electromatic relief valve is to re-duce the challenge rate of the code safety valves. No credit is taken for the pressurizer relief valve in the design overpressure transient. A duel setpoint is specified; the lower setpoint is associated with considerations relating to the potential for over-pressurization of the reactor coolant system under cold conditions. References (1) FSAR, Tables 9-10 and 4-3 through 4-7 (2) FSAR, Sections 4.2.5.1 and 9.5.2.3 (3) FSAR, Section 4.2.5.4 (4 ) FS AR, Sections 4. 3.10. 4 and 4. 2.4 (5) FSAR, Section 4.3.7 3-2 1869 189
3.5.5 Pressurizer Safety and Relief Valve Monitors Applicability Applies to the operability requirements for the delta pressure and acoustic instrumentation that monitors the status of the pressurizer code safety valves and the pressurizer Electromatic relief valve. Objective To provide reactor operators with a reliable means of detecting anc monitoring pressurizer safety and relief valve discharge flow. S pecifica tion 3.5.5.1 The three (3) delta pressure monitors, one for each of the two (2) pressurizer code safety valves and one for the pressurizer Electro-matic relief valve, shall be operable during STARTUP and power OPERA-TION. If one or more of the delta pressure monitors in inoperable, it shall be returned to operation prior to startup following the next cold shutdown. 3.5.5.2 An acoustic monitor for the pressurizer Electromatic relief valve shall be operable during STARTUP and power OPERATION. If the acoustic monitor is inoperable, it shall be returned to operation prior to startup following the next cold shutdown. Bases Discharge flow f rom the two (2) pressurizer code safety valves anc the Electromatic relief valve is measured by differential pressure transmitters connected across elbow taps downstream of each valve. A delta pressure indication from each pressure transmitter is avail-able in the control room to indicate safety or relief valve line flow. An alarm is also provided in the control room to indicate that discharge fioa a pressurizer safety or relief valve is occur-ing. In addition, an acoustic monitor is provided to detect flow in the relief valve oischarge line. An alarm and a flow indication is provided in the control room for the acoustic monitor. In the event that a delta pressure monitor or the acoustic monitor becomes inoperable, access to the containment would most likely be equired; however, a reactor shutdown to allow containment acce ss for this repair is not justifiable due to the existence of alternate means of detecting and monitoring safety.or relief valve discharge flow. The following indications are available to the reactor operator to monitor safety or relief valve discharge flow: (1) Reactor coolant drain tank level, pressure, and temperature (2) Safety and relief valve tailpipe tempe ra tures Based upon the existence of these means of monitoring safety and relief valve discharge flow, continued operation until the next cold shutdown is acceptable in the event that a delta pressure or acoustic monitor is inoperable. 3-40a 1869 190, 4
4 SURVEILLANCE STANDARDS Specified intervals may be adjusted plus or minus 25 percent to accommcdate normal test schedules. 4.1 OPERATIONAL SAFETY REVIEk' Applicability Applies to items directly related to safety limits and limiting conditions for operation. Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions. Specificaion 4.1.1 The minimum frequency and type of surveillance required for re-actor protection system and engineered safety feature protec-tion system instrumentation when the reactor is critical shall be as stated in Table 4.1-1. Surveillances in Table 4.1-1, not performed due to reactor shutiown greater than one month, shall be performed prior to STARTUP. 4.1.2 Equipment and sampling test shall be performed as detailed in Tables 4.1-2 and 4.1-3. Bases Check Failures such as blown instrument fuses, defective indicators, or faulted amplifiers which result in " upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance. Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking f re-quency stated is deemed adequate for reactor system instrumentation. Calibration Calibration shall be performed to assure the presentation and acquisition of accurate information. The nuclear flux (power range) channels ampli-fiers shall be checked and calibrated if necessary, every shif t against a heat balance standard. The frequency of heat balance checks will assure that the difference o'etween the out-of-core instrumentation and the heat balance remains less than 4%. 4-1 1869 I9I
TABLE 4.1-1 (Continued) CilANNEL DESCRIPTION CllECK TEST CAllBRATE REF1 ARKS
38. Steam Generator Water Level W NA R

39. Turbine Overspeed Trip NA R NA

40. Sodium Thiosulfate Tank Level NA NA R Indicator

41. Sodium ilydroxide Tank Level NA NA R Indicator

42. Diesel Generator Protective NA N R Relaying i 43. 4 KV ES Bus Undervoltage Relays NA M(1) R (1) Relay operation will be checked (Diesel Start) by local test pushbuttons.

44. Reactor Coolant Pressure S(l) M R (1) When reactor coolant system is Dil Valve Interlock Bistable pressurized above 300 psig or Taves is greater than 200*F.

45. Loss of Feedwater Trip S(l) M(1) R (1) When reactor >10% power

46. Turbine Trip / Reactor Trip S(l) M R (1) When reactor >20% power CO
,-ys 47. Pressurizer Code Safety Valve and S(l) R R (1) When TAVG is greater than 2OO*F sg) Electromatic Relief Valve delta P/ flow 3 48. Pressurizer Electromatic S(l) R R (1) When TAVG is greater than 200*F PN) Relief Valve - acoustic / flow S - Each Shift T/W - Twice per week R - Each Refueling Period D - Daily B/M - Every 2 months NA - Not Applicable W - Weekly Q - Quarterly B/W - Every two weeks P - Prior to each startup M - Monthly if not done previous week
TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency
1. Control Rods Rod drop times of all Each refueling shutoown full length rods

2. Control Rod Movement of each rod Every two weeks, when reactor Movement is critical

3. Pressurizer Safety Setpoint* 50% each refueling period Valves

4. Main Steam Safety Setpoint 25% each refueling period Valves

5. Refueling System Functional Start of each refueling period Interlocks

6. Main Steam (See Section 4.8)
Isolation Valves
7. Reactor Coolant Evaluate Daily, when reactor coolant System Leakage system temperature is greater than 525*F

8. Charcoal and high DOP test on REPA filters, Each refueling period and at efficiency filters freon test on charcoal any time work on filters for Control Room, filter units could alter their integrity and RB Purge Filters

9. Spent Fuel Cooling Functional Each refueling period prior to System fuel handling

10. Intake Pump House (a) Silt Accumulation- Each refueling period Floor Visual inspection of Intake (Elevation 262 Ft. Pump House Floor 6 in.) (b) Silt Accumulation Quarterly Measurement of Pump House Flow

11. Pressurizer Setpoint Refueling period Electromatic Relief Valve
The setupoint of the pressurizer code safety valves shall be in accoraance with ASME Boiler and Pressurizer Vessel Code, Section III, Article 9, Winter, 1968.
4-8 Amendment No. 30 (5/11/77) 1869 193
Draft Technical Specification Corresonding to Section 11.2.3
'1869 194
4.6 EMERGENCY PCWEE 3YSTEM PERIODIC TESTS Applicability Applies to periodic testing and surveillance requirement of the emergency power system. Objective To verify that the emergency power system will respond promptly and properly when required. Specification The following tests and surveillance shall be performed as stated: 4.6.1 Diesel Generators
a. Manually-initiated start of the diesel generator, f ollowed by manual synchronization with other power sources and assumption of load by the diesel generator up to the nameplate rating (3000 kw). This test will be conducted every month on each diesel generator. Norcal plant operation will not be affected.

b. Automatic start of each diesel generator and restoration to operation of particular vital equirsent, initiated by an actual loss of normal a-c station service power supply to-gether with a simula ed Engineered Safeguards Actuation Signal.
Following input of the Engineered Safeguards Actuation Signal, it shall be verified than the circuit breakers, supplying power to the manually transferred loads for pressurizer heater Groups 8 and 9, have been tripped. This test will be conducted durinb reactor shutdown for refueling to assure that the diesel-gen-erator will start assuming load in ten seconds and assume the load of all safeguards equipment listed in 4.5.1. lb within 60 seconds af ter the initial starting signal. ,
c. Each diesel generator shall be given an inspection at least annually in accordance with the manuf acturer's recommendations for this class of stand-by service.
4.6.2 Station Batteries
a. The voltage, specific gravity, and liquid level of each cell will be measured and recorded monthly.

b. The voltage and specific gravity of a pilot cell will be measured and recorded weekly.

c. Each time data are recorded, new data shall be compared with old to detect signs of abuse or deterioration.
4-46 1869 . 19cD
d. The battery will be subjected to a load test at a frequency not to exceed refueling periods. The battery voltage as a function of time will be monitored to establish that the battery performs as expected during this load test.
Bases The tests specified are designed to demonstrate that one diesel gener-ator will provide power for operation of safeguards equipment. The y also assure that the emergency generator control system and the control systems for the safeguards equipment will function automatically in the event of a loss of normal a-c station service power or upon the receipt of an an engineered safeguards Actuation Signal. The automatic tripping of manually transferred loads, on an Engineered Safeguards Actuation Signal, protects the diesel generators from a potential over-load con-dition. The testing frequency specified is intended to identify anc permit correction of any mechanical or electrical deficiency before it can result in a system failure. The fuel oil supply, s tar ting circuits, and controls are continuously monitored and any faults are alarmed and indicated. An abnormal condition is these sys tems would be signaled without having to place the diesel generators on test. Percipitous failure of the station battery is extremely unlikely. The surveillance specified is that which has been demonstrated over the years to provide an indication of a cell becor.ing unserviceable long bef ore it fails. REFERENCE (1) FSAR, Section 8.2 4-47 1869 196.
Draft Technical Specification Correspondin8 to Section 11.2.4 1869 197
3.6 REACTOR BUILDING Applicability Applies to the containment integrity of the reactor buildinb. Objective To assure containment integrity. Specification 3.6.1 Containment integrity, as defined in Section 1.7, shall be main-tained whenever all three of the following conditions exist:
a. Reactor coolant pressure is 300 psig or greater.

b. Reactor coolant temperature is 200 "F or greater.

c. Nuclear fuel is in the core.
3.6.2 Containment integrity shall be maintained when both the reactor coolant system is open to the containment atmosphere and a shut-down margin exists that is less than that for a refueling shut-down. - 3.6.3 Positive reactivity insertions which would result in a reduction in I shutdown margin to less than 1% k/k shall not be made by control rod motion or boron dilution unless containment integrity is being maintained. 3.6.4 The reactor shall not be critical when the reactor building internal pressure exceeds 2.0 psig or 1.0 psi vacuum. 3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all manual containment isolation valves which should be closed are closed and are conspicuously marked. 3.6.6 If, while the reactor is critical, a reactor building isolation valve is determined to be inoperable in a posit 1sn other than the required position, the other reactor building isolation valve in the line shall be testec to insure operability. If the inopera-ble valve is not restored within 48 hours, the operable valve will be closed or the reactor shall be brought to the cold shutdown condition within an additional 24 hours. 3.6.7 One hydrogen recombiner shall be operable during STARTUP cad POWER OPERATION. With the hydrogen recombiner inoperable, rescore the recombiner to operable status or bring che reactor tL hot standby within seven (7) days. 3-41 .
Bases The Reactor Coolant System conditions of cold shutdown assure that no steam will be formed and hence no pressure will build up in the containment if the Reactor Coolant System ruptures. The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in any occurrence. A condition requiring integrity of containment exists whenever the reactor coolant system is open to the atmosphere and there is insufficient soluble poison in the reactor coolant to maintain the core one percent subcritical in the event all control rods are withdrawn. The reactor building is designed for an internal pressure of 55 psig, ano an exteri.al pressure 2.5 psi greater than the internal pressure. The operability of the hydrogen recombiner ensures that this eaaipment will be available to maintain the hydrogen concentration within containment be-low its flammable limit during post-LOCA conditions. The recombiner unit is capable of controlling the expected hydrogen generation associateo with
1) zirconium-water reactions, 2) radiodlytic decomposition of water and

3) corrosion of metals within containment. The recombiner is designed in accordance with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Followinh a LOCA", March 1971, the acceptance criteria of S.R.P. 6.2.5., and NUREG-0578, July 1979. In addition to the installed hydrogen recombiner a second hydrogen recombiner can be installed arter an accident within the time period available before it is needed; all piping, electrical, and structural provisions for the second recombiner are available.
The hydrogen mixing is provided by the reactor building ventilation system to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit. REFEREN CES FSAR Section 5.2.2.4.3 3-41a 1869 199
4.4.4 Hydrogen Recombiner System Applicability Applies to the testing of the hyorogen recombiner and associated controls. Objective To verify that the hydrogen recombiner and associated controls are operable. 4.4.4.1 Specification
a. At least once per 92 days, during STARTUP or POWER OPERATION, perform a hydrogen recombiner system functional test to d emons t ra te that the minimum heater sheath temperature in-creases to 2,700*F within 90 minutes and is maintained 2,700*F for at least 2 hours. This test shall also be performed prior to STARTUP following a reactor outage greater than 90 days.

b. At least once per 18 months, perform the following surveillance:

1. A channel calibration of all reccm'siner instrumentation and control circuits.

2. Verify through a visual examination that there is no evi-dynce of abnormal conditions within the recombiners (i.e. ,
loose wiring or structural connections, d eposits of foreign materials, etc.)
3. Verify during a recombiner system functional test that the haater sheath temperature increases to 2,1200*F within 5 hours and is maintained 2,1200*F for at least 4 hours.

4. Verify the integrity of the heater electrical circuits by performing a continuity and resistance to ground test following the above required functional test. The resistance to ground f or any heater phase shall be 2, 10,000 ohms.
Bases The surveillance program described above provides high assurance that the hydrogen recombiner system will be available to perform its post-LOCA function of reducing the containment hydrogen con-centration to below 4.1 volume percent. 4-38a i869 200
' ~
Draft Technical Specifications Corresponding to Section 11.2.5
'n 1
1869 201
, , e. '
3 4
TABLE 3.5-1 Cont ,ed INSTRUMENTS OPERATING CONDITIONS Functional Unit (A) (B) (C) Minimum Minimum Engineered Safeguards Operable Degree of Operator Action if Conditions Channels Redundancy of Column A cannot be met (a)
3. Reactor Building 1 solation and Reactor Building Cooling System

a. Reactor Building 4 psig Instrument Channel 2 I Hot Shutdown

b. Manual Pushbutton 2 1 Hot Shutdown

c. RPS Trip 2 1 Hot Shutdown

d. Reactor Building 30 psig 2 1 Hot Shutdown (a) If minimum conditions are not met within 24 hours, the unit shall then be placed u, in a cold shutdown condition.
N' (b) Also initiates Low Pressure injection.
4. Reactor Building Spray System

a. Reactor Building 30 psig Instrument Channel 2 (b) 1 Hot Shutdown CD cys b. Spray Pump Manual Switches (c) 2 i Hot Shutdown si?
PN) (a) If minimum conditions are not met within 24 hours, the unit shall then be placed C'] in a cold shutdown condition. N (b) Two out of three switches in each actuation channel operable. (c) Spray valves opened by manual pushbutton listed in item 3 above.
TABLE 4.1-1 Cont ' led INSTRUMENTS OPERATING CONDITIONS CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS
19. Reactor Building Emergency Cooling and Isolation System Channels

a. Reactor Building S(l) M(1) R (1) When CONTAINHENT INTEGRITY is required 4 psig Channels

b. Manual Pushbutton S(l) M(1) NA ( 1) When CONTAINMENT INTEGRITY is required

c. RPS Trip S(l) M(1) NA ( 1 ) When CONTAINMENT INTEGRITY is required

d. Reactor Building 30 psig S(l) M(1) R (1) When CONTAINMENT INTEGRITY is required

20. Reactor Building Spray NA Q NA System Logic Channel i
'a 21. Reactor Building Spray System Analog Channels
a. Reactor Building NA M R 30 psig Channels

22. Pressurizer Temperature S NA R Channels

23. Control Rod Absolute Position S(l) NA R (1) Check with Relative Position Indicator

24. Control Rod Relative Position S(l) NA R (1) Check with Absolute Position Indicator CO cp, 25. Core Flooding Tanks W
. a. Pressure Channels S(l) NA R (1) When Reactor Coolant system pressure rs) is greater than 700 psig CD L/4 b. Level Channels S(l) NA R
26. Pressurizer Level Channels S NA R

27. Makeup Tank Level Channels D(1) NA R (1) When Makeup and Purification System is in operation
Draf t Technical Specification Corresponding to Section 11.2.6 1869 204
3.5.6 Saturation Margin Meter Applicability Applies to the Saturation Margin Meter and associated alarm. Objective Provide the reactor operator with a reliable indication of the margin that exists between the water properties of the reactor coolant system and saturation conditions. Specification 3.5.6.1 The Saturation Margin Meter shall be operable during startup and power operation. If the Saturation Margin Meter is in-operable, a procedure for calculation of saturation pressure margin and saturation temperature margin shall be made readily accessible to the reactor operator and the saturation meter returned to service at the earliest practicable time. Bases The Saturation Margin Meter provides a quick and reliable means for determination of saturation temperature and satur-ation pressure margins. The hand calculation of saturation pressure and saturation temperature margins can be easily and quickly performed since it only requires knowledge of recircu-lation loop temperatures and system pressure, and the use of steam tables; accordingly, hand calculation provides a suitable backup for the Saturation Margin Indicator. 1-40b 1869 205
TABL{
.1-1 (Continued)
CilANNEL DESCRIPTION CHECK TEST CALIBRATE ret 1 ARKS
38. Steam Generator Water Level W NA R

39. Turbine Overspeed Trip NA R NA

40. Sodium Thiosulfate Tank Level NA NA R Indicator

41. Sodium flydroxide Tank Level NA NA R Indicator

42. Diesel Generator Protec tive NA NA R Re laying

43. 4 KV ES Bus Undervoltage Relays NA M(1) R (1) Relay opreation will be checked (Diesel Start) by local test pushbuttons.

44. Reactor Coolant Pressure S(l) M R (1) When reactor coolant system is ul Valve Interlock Bistable pressurized above 300 psig or Taves is greater than 200*F.

45. Loss of Feedwater Trip S(l) M(1) R (1) When reactor > 10% power.

46. Turbine Trip / Reactor Trip S(l) M(1) R (1) When reactor > 20% power.

47. Pressurizer Code Safety Valve - delta P/ flow S(l) R R (1) When TAVG is greater than 200oF.
48 Pressurizer Electromatic Relief Valve - acoustic / flow S(l) R R (1) When TAVG is greater than 2000F. c33 49. Saturation Margin Meter S(l) R R (1) When TAVG is greater than 2000F. '43 S - Each Shift T/W - Twice per week R - Each Refueling Period
) D - Daily B/M - Every 2 months NA - Not Applicable Ch W - Weekly Q - Quarterly B/W - Every two weeks M - Monthly P - Prior to each startup if not done previous week
Draft Technical Specification Corresponding to Section 11.2.7 1869 207
3.4 DECAY HEAT REMOVAL - TURBINE CYCLE Applicability Applies to the operating status of equipment that functions to remove decay heat, utilizing the secondary side of the steam generators. Objective To define the conditions necessary to assure immediate availability of the auxiliary feedwater system and main steam safety valves. Specification 3.4.1 With the reactor coolant system temperature greater than 2500F, three independent steam generator emergency feedwater pumps and and associated flow paths shall be OPERABLE with:
a. Tuo emergency feedwater pumps, each capable of being powered from an OPERABLE emergency bus, and

b. One emergency feedwater pump capable of being powered from an OPERABLE steam supply system.
With one emergency feedwater pump or flow / path inoperable, restore the inoperable pump or flow path to OPERABLE status within 48 hours or be in COLD SHUTDOWN within the next 12 hours. With more than one emergency feedwater pumps or flow path inoperable, restore the inoperable emergency feedwater pumps or flow paths to operable status or be suberitical within I hour, in at least HOT SHUTDOWN within the next 6 hours, and in COLD SHUTDOWN within the following 6 hours.
c. Four of six turbine bypass valves are OPERABLE.
3.4.2 The condensate storage tanks (CSTS) shall be OPERABLE with a mini-mum of 150,000 gallons of condensate available in each CST. With a CST inoperable, restore the CST to operability within 4 hours or be in at least HOT STANDBY within the next 6 hours, at least HOT SHUTDOWN within the next 6 hours, and COLD SHUTDOWN within the next 24 hours. 3-25 1869 208
3.4.3 With the reactor coolant system temperature greater than 2500F, all eighteen (18) main steam safety valves shall be operable or, if any are not operable, the maximum overpower trip setpoint (see Table 2.3-1) shall be reset as follows: Maximum Number of Maximum Overpower Safety Valves Disabled on Trip Setpoint Any Steam Generator (% of Rated Power) 1 92.4 2 79.4 3 66.3 With more than 3 main steam safety valves inoperable, restore at least fif teen (15) main steam safety valves to operable status within 4 hours or be in et least hot standby wit.hin the next 6 hours and in cold shutdown within the following 30 hours. Bases A reactor shutdown following power operation requires removal of core decay heat. Normal decay heat removal is by the steam generators with the steam dump to the condenser when system temperature is above 250*F and by the decay heat removal system below 250*F. Core decay heat can be continuously dissipated up to 15 percent of full power via the steam bypass to thee condenser as feedwater in the steam generator is converted to steam by heat absorption. Normally, the capability to return feedwater flow to the steam generators is provided by the main feedwater system. The main steam safety valves will be able to relieve to atmosphere the total steam flow if necessary. If main steam safety valves are inoperable, the power level must be reduced, as stated in Technical Specification 3.4.3, such that the remaining safety valves can accommodate the decay heat. In the unlikely event of complete lo__ of of f-site electrical power to the station, decay heat removal is by either the steam-driven emergency feedwater pump, or two half-sized motor-driven pumps. . Steam discharge is to the atmosphere via the main steam safety valves and controlled at-mospheric relief valves, and in the case of the turbine driven pump, from the turbine exhaust.(1) Both motor-driven pumps are required initially to remove decay heat with one eventually sufficing. The minimum amount of water in the condensate storage tanks, contained in Technical Specification 3.4.2, will allow cooldown to 250*F with steam being discharged to the atmosphere. After cooling to 250*F, the decay heat removal system is used to achieve further cooling. An unlimited emergency feedwater supply is available from the river via either of the two motor-driven reactor building emergency cooling water pumps for an indefinite period of time. 3-26 1869 209
The requirements of Technical Specification 3.4.1 assure that before the reactor le heated to above 250*F, adequate auxiliary feedwater capacity is available. One turbine driven pump full capacity (920 gpm) and the two half-capacity motor-driven pumps (460 gp=, each) are specified. However, only one half-capacity motor-driven pump is necessary to supply auxiliary feedwater flow to the steam generators in the onset of a small break loss-of-coolant accident (Reference 2). The requiremen'_s of Technical Specification 3.4.1 assure that at least 920 gpm is available at all times to both steam generators giving re-dundant capacity except for a limited time of 72 hours to allow for com-ponent maintenance. Further degradation of the emergency feedwater system requires the reactor to be suberitical within I hour. The feedwater line break accident performed for TMI-2 (Reference 3) shows satisfaction of core thermal power limits and reactor coolant system pressure limits assuming f ull auxiliary feedwater flow within 40 seconds. The Technical Specification 3.4.1 provides assurance that this flow will be available with automatic initiation following loss of both main feedwater pumps. REFERENCES (1) FSAR, Section 10.2.1.3 (2) " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant," Volume I and II, Babcock and Wilcox, May 7, 1979. (3) Three Mile Island Nuclear Station - Unit 2, Final Safety Analysis Report, USNRC Docket No. 50-320. 3-26a 1869 210
TABLF ,1-1 (Continued) CilANNEL DESCRIPTION CllECK TEST CALIBRATE REMARKS
38. Steam Generator Water Level W NA R

39. Turbine Overspeed Trip NA R NA

40. Sodium Thiosulfate Tank Level NA NA R Indicator

41. Sodium liydroxide Tank Level NA NA R Indicator

42. Diesel Generator Protective NA N R Relaying

43. 4 KV ES Bus Undervoltage Relays NA M(1) R (1) Relay operation will be checked (Diesel Start) by local test pushbuttons.

44. Reactor Coolant Pressure S(l) M R (1) When reactor coolant system is Dil Valve Interlock Bistable pressurized above 300 psig or Taves is greater than 200*F.

45. Loss of Feedwater Trip S(l) M(1) R (1) When reactor > 10% power.

46. Turbine Trip / Reactor Trip S(l) M(1) R (1) When reactor > 20% power.

47. Pressurizer Code Safety Valve and Electromatic Relief Valve delta P/ flow S(l) R R (1) When TAVG is greater than 2000F.

48. Pressurizer Electromatic
__, Relief Valve - acoustic / flow S(l) R R (1) When TAVG is greater than 200 F. CO ,gs 49. Saturation Margin Meter S(l) M(1) R (8) When T Ayg is greater than 2000F.
50. Emergency Feedwater Flow Instru- NA M(1) R (1) Emergency Feedwater is not normally IN) mentation in operation.
m
--^
S - Each Shift T/W - Twice per week R - Each Refueling ?criod D - Daily B/M - Every 2 months NA - Not Applicable W - Weekly Q - Quarterly B/W - Every two weeks M - Monthly P - Prior to each startup if not donc previous week
TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frecuency
1. Control Rods Rod drop times of all Each refueling shutdown full length rods

2. Control Rod Movement of each rod Every two weeks, when reactor Movement is critical

3. Pressurizer Safety Setpoint* 50% each refueling period Valves

4. Main Steam Safety Setpoint 25% each refueling period Valves

5. Refueling System Functional Start of each refueling period Interlocks

6. Main Steam (See Section 4.8)
Isolation Valves
7. Reactor Coolant Evalua te Daily, when reactor coolant System Leakage system temperature is greater than 525*F

8. Charcoal and high DOP test on HEPA Each refueling period and at efficiency filters filters, freon test any time work on filters for Control Room, on charcoal filter could alter their integrity and RB Purge units Filters

9. Spent Fuel Cooling Functional Each refueling period pric: to System fuel handling

10. Intake Pump House (a) Silt Accumulation- Each refueling period Floor Visual inspection of (Elevation 262 Ft. Intake Pump House Floor 6 in.) (b) Silt Accumulation Quarterly Measurement of Pump House Flow

11. Pressurizer elec- Setpoint Each refueling period tromatic relief valve

12. Back-up instrument Functional Each refueling period air supply system
The setupoint of the pressurizer code safety valves shall be in accordance with ASME Boiler and Pressurizer Vessel Code, Section III, Article 9, Winter, 1968.
1869 212 4-8
4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 EMERGENCY LOADING SEQUENCE Applicability Applies to periodic testing requirements for safety actuation systems. Objective To verify that the Emergency loading sequence and automatic power trans-fer is operable. Specifications 4.5.1.1 Sequence and Power Transfer Test
a. During each refueling interval, a test shall be conducted to demon-strate that the emergency loading sequence and power transfer is operable.

b. The test will be considered satisfactory if the following pumps and fans have been successfully started and the following valves have completed their travel on preferred power and transferred to the emergency power as evidenced by the control board component operating lights, and either the station computer or pressure /
flow indication or, in the case of the Motor Driven Emergency Feedwater Pumps, the pump Interlock.
- M. U. Pump - D. H. Pump and D. H. Injection Valves and D. H. Supply Valves - R. B. Cooling Pump - R. B. Ventilators - D. H. Closed Cycle Cooling Pump - N. S. Closed Cycle Cooling Pump - D. H. River Cooling Pump - .. S. River Cooling Pump - D. H. and N. S. Pump Area Cooling Fan - Screen House Area Cooling Fan - Spray Pump. (Initiated in coincidence with a 2 out of 3 R. B. 30 psi Pressure Test Signal.) - Motor Driven Emergency Feedwater Pump Interlock.
4.5.1.2 Sequence Test
a. At intervals not to exceed 3 months, a test shall be conducted to demonstrate that the emergency loading sequence is operable, this test shall be performed on either preferred power or emer-gency power.

b. The test will be considered satisf actory if the pumps and fans listed in 4.5. lb have been successfully started and the valves listed in 4.5.1.lb have completed their travel as evidenced by the control board component operating lights, and either the station computer or pressure / flow in t tgon.} } }
4-39
Bases The Emergency loading sequence and automatic power transfer controls the operation of the pumps associated with the emergency core cooling system and Reactor Building cooling system. A successful test of the emergency loading sequence and automatic power transfer is a prerequisite to any system test of the emergency core cooling system or reactor building cooling system. References (1) FSAR Section 7 (2) FSAR Section 1.4 1869 2I4
4.9 EMERGENCY FEEDWATER SYSTEM PERIODIC TESTING Applicability Applies to the perodic testing of the turbine driven and two motor-driven emergency feedwater pumps, assocaited actuation signals, and valves. Objective To verify that the auxiliary feedwater system is capable of performing its design function. Specification 4.9.1 TEST 4.9.1.1 Whenever the Reactor Coolant System temperature is greater than j 2500F, the emergency feedwater pumps shall be tested in the recirculation mode in accordance with the requirements and acceptance criteria of ASME Section XI Article IWP-3000. The test frequency shall be at least every 31 days +7 days of plant operation at Reactor Coolant Temperature above 2500F. 4.9.1.2 At least once per 31 days each valve (manual, power operated, or automatic) shall be verified to be in its correct position. This applies to valves C0 V10 A and B, EF VI A and B, EF V2 A and B, EF V10 A and B, EF V16 A and B, and MSV 2A and B. 4.9.1.3 At least once per 18 months, during shutdown, verify that: (a) each emergency feedwater pump starting logic actuates upon receipt of an auxiliary feedwater actuation signal, and (b) valves in the emergency feedwater flow paths actuate to their l correct position on an emergency feedwater actuation signal i and that the manual control valve station functions properly. I 4.9.1.4 On a quarterly basis , the valves which are a part of the emer-gency feed system discharge (EFV-30A and 30B) will be checked for proper operation by cycling the valve over its full stroke. 4.9.1.5 Prior to start-up, following a cold shutdown of longer than 30 days' duration, conduct a test to demonstrate that the motor driven emergency feed pumps can pump water from the CST to the steam generators. 4.9.2 ACCEPTANCE CRlTERIA These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly. Bases The 31 day testing f requency will be sufficient to verify that the turbine driven and two motor-driven emergency feedwater pumps are operable and that the associated valves are in the correct alignment. ASME Section XI Article IWP-3000 specifies requirements and acceptance standards for the testing of nuclear safety related pumps. Compliance with the normal 4-52 lbb
acceptance criteria of IWP-3000 assures that the rmergency feedwater pumps are operating as expected. The test frequency of 31 days (nominal) has been demonstrated by the B&W Emergency Feedwater Reliability Study to assure an appropriate level of reliability. If testing under Article IWP-3000 indicates that the flow and/or pump head for a particular pump is not within the normal acceptance standard, Article IWP-3000 requires that an evaluation of the pump performance shall be completed within 96 hours or the pump declared inoperable. For the case of the emergency feedwater system, the system shall be considered operable if under the worst case single pump failure, a minimum of 500 gpm of emergency feedwater can be delivered when steam generator pressure is 1050 psig and one steam generation is isolated. A flow of 500 gpm, at 1050 psig head, ensures that sufficient emergency feedwater, demonstrated to be acceptable for plant cooling requirements under transient and accident conditions, can be delivered to either steam generator flow path. The 18 month surveillance requirements ensure that the overall emergency feedwater system functional capability is maintained comparable to the original design standards. 4-52a i869 216
Draft Technical Specifications Corresponding to Section 11.2.8 1869 217
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TABLE 4.1-1 (Continued) CllANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS
28. Radiation Monitoring Systesm (1) W(2) M Q(3) (1) Includes po s t accident monitoring instrumentation (a) liigh Range-Containment (b) Undiluted Containment Exhaust (c) Auxiliary and Fuel Handling Building Exhaust (d) Condenser Off Gas (2) Using the installed check source when background is less than twice the expected increase in cpm which would y result from the check source alone.
[ Background readings greater than this value are sufficient in themselves to show that the monitor is function-Ing. (3) Except area gamma radiation monitors RM-G6, RM-G7, and RM-G8 which are lo-cated in high radiation areas of the C3D Reactor Building. These monitors will C3% be calibrated quarterly or at the next
'43 scheduled reactor shutdown following the py quarter in which calibration would
__, normally be due, if a shutdown during sy) the quarter does not oc c u r .
29. liigh and Low Pressure Injection NA NA R Systems: Flow Channels
TABLE 4.1-1 (Continued) CHANNEL DESCRIPTION CllECK TEST CALIBRATE REF1 ARKS
30. Borated Water Storage Tank Level W NA R Indica to r

31. Boric Acid Mix Tank

a. Level Channel NA NA R

b. Temperature Channel M NA R

32. Reclaimed Boric Acid Storage Tank

a. Level Channel NA NA R i b. Temperature Channel M NA R

33. Con ta inmen t Tempe ra ture NA NA R

34. Incore Neutron Detectors M(1) Rt NA (1) Check functioning; including functioning of computer read-out or recorder readout when reactor power is greater than 15%
CX? 35. Emergency Plant Radiation M(1) NA R (1) Battery check C3' Instruments W
36. Strong Motion Accelerometer Q(1) NA Q (1) Battery check
)
IV C3 37. Reactor Building Sump and NA NA R (1) Includes post-accident moni-Containment Level (1) toring Instrumentation (a) Narrow Range (sump) (b) Wide Range (Containment)
F F 0 0 0 0 0 0 2 2 n n a a h h t t r r e e t t a a e e r r g g s s i i g G y V S A A K T T R A n n M e e E h h R W W
) )
1 1 ( ( E T A R B I R R L A C ) ) ) d T 1 1 e S ( ( u E M M n T i t n o C ( ) ) K l l 1 C ( (
- E S S 1 l l . C 4
E L B A T p n i e r g N T o O p r I d T m ly P u f I P R g C t n S n i E a d n D l l o o ii L o ut E C B a N r N r rt A l o on t t e Cl cp cc ai an er eo RT RC 1 2 5 5 a AFG NN-Ifa
Draft Technical Specification Corresponding to Section 11.2.9 1869 222
3.5.7 Reactor Coolant Pump Trip Applicabilit,y Aoplies to the operational status of the reactor coolant pump trip. Objective To specify requirements for operability of the reactor coolant pump trip. S pe ci f ica tion 3.5.6.1 The reactor coolant pump trip, with a minimum of a high pressure injection actuation channel input and one pump current channel input per OPERABLE reactor coolant pump, shall be OPERABLE during reactor STARTUP and POWER OPERATION. With the reactor coolant pump trip inoperable, place the reactor in hot shut-down within 24 hours. Bases Analysis has shown that, for a certain range of small primary breaks, inacceptable clad temperatures may result if the reactor coolant pumps are not tripped at a time when the Reactor Coolant System void fraction has achieved a high level. To prevent these detrimental consequences, the reactor coolant pump trip will promptly trip the reactor coolant pumps when system conditions indicate that a small break in this range may be in progress. The system actuates when High Pressure Injection has been initiated and the reactor coolant system void fraction has reached a nominal value, as sensed by the reactor coolant pump current monitors, which indicates that a high void fraction may develop. 3-40c
Table 4.1-1 (Continued) I Channel Description Check Test Calibrate Remarks
51. Reactor Coolant Pump S(l) M(1) R (1) When TAVG is greater than 2000F, Trip

52. Reactor Building S(l) M(1) R (1) When TAVG is greater than 2000F.
flydregen Concentration I e u N N 4
Draft Technical Specification Corresponding to Section 11.2.10 1869 225
3.19 Separation of TMI-l and TMI-2 Applicability Applies to interconnections between TMI-l and .MI-2 which have the potential for transferring significant quantities of con-tamination between units. Objective To control the transfer of radioactivity from TMI-1 to TMI-2 via system interties. Specification 3.19.1 The isolation devices for the system interties described in Table 3.19-1 shall remain isolated unless written approval has been received from the NRC. If approval for use of interties is received, use shall proceed under preestablished plant procedures. If a system intertie, listed in Table 3.19-1 is found defeated without prior NRC approval, a " Thirty Day Written Report" shall be prepared and submitted as required by Technical Specification 6.9.2.B.(5). 3.19.2 No additional TMI-1/TMI-2 interties, with the potential of trans-ferring significant quantities of radioictivity, shall be created without prior NRC approval. Bases Interties exist between TMI-l and TMI-2 that have the potential for trans-ferring contamination to TMI-l as a result of restoration activities at TMI-2. These interties should remain isolated unless approval for their use is received from the NRC. 3-95
}O/93 226
Table 3.19-1 TMI-1/TMI-2 Interties (1) Unit 2 Reactor Coolant Bleed Holdup Tank - Unit 1 Reactor Coolant Waste Evaporator (2) Un' t 1 Miscellaneous Waste Evaporator - Unit 2 Evaporator Con-densate Test Tanks (3) Unit 2 Neutralizer Tanks, Contaminated Drain Tanks, Reactor Coolant Bleed Holdup Tanks, Auxiliary Building Sump Tanks and Miscellaneous Waste Hold-up Tanks - Unit 1 Liquid Waste Disposal System (4 ) Unit 1 Evaporator Concentrate - Unit 2 Evaporator Concentrate (5) Unit 1 Spent lon Exchange Resin - Unit 2 Spent Ion Exchange Resin 3-96 1869 227
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/
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6.9.2 REPORTING REQUIREMENTS (cont'd) Note: This item is intended to provide for reporting of potentially generic problems B. Thirty Day Written Reports. 1/The reportable occurrences discussed below shall be the subject of written reports to the Director of the tppropriate Regional Office within thirty days of occurrence of the event. The writcen report shall include narrative material to provide complete explanation of the cause of the event, circum-stances surrounding the event, any corrective action, and component failure data. (1) Reactor protection system or engineered safety feature instru-ment settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems. (2) Conditions leading to operation in a degraced mode permitted by a limiting condition for operation or plant shutdcwn re-quired by a limiting condition for operation. Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configura-tions as described in items 6.9.2.B(1) and 6.9.2.B(2) need not be reported except where test results themselves re-veal a degraded mode as described above. (3) Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems. (4) Abnormal degradation of systems other than those specified in item 6.9.2.A(3) above designed to contain radioactive material resulting from the fission process. Note: Sealed sources or calibration sources are not ingluded under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifi-cations need not be reported under this item. (5) Observed defeat of an isolation device, which separates a tieline between Units 1 and 2, without prior NRC approval. 6-17 1869 229
Draft Technical Specification Corresponding to Section 11.2.11 1869 230
3.5.3 ENGINEERED SAFEGUARDS PROTECTION SYSTEM ACTUATION SETPOINTS Applicability This specification applies to the engineered safeguards protection system actuation setpoints. Objective To provide for automatic initiation of the engineered safeguards protection system in the event of a breach of Reactor Coolant System integrity. Specification 3.5.3.1 The engineered safeguards protection system actuation setpoints and permissible bypasses shall be as follows: Initiating Signal Function Setpoint High Reactor Building Reactor Building Spray f,30 psig Pressure (1) High-Pressure Injection f,4 psig Low-Pressure. Injection f,4 psig Start Reactor Building Cooling & Reactor Building Isolation f,4 psig Low Reactor Coolant High Pressure Injection >1600(2) and System Pressure >500(3) psig Low Pressure Injection >l600(2) and l
>500(3) psig (1) May be bypassed for reactor building leak rate test.
(2) May be bypassed below 1750 psig and is automatically resinstated above 1750 psig. (3) May be bypassed below 900 psig and is automatically resinstated above 900 psig. Bases High Reactor Building Pressure The basis for the 30 psig and 4 psig setpoints for the high. pressure signal is to establish a setting which would be reached in adequate time in the event of a LOCA, cover a spectrum of break sizes and yet be far enough above normal operation maximum internal pressure to prevent spurious initiation. Low Reactor Coolant System Pressure The basis for the 1600 and 500 psig low reactor coolant pressure setpoint for high and low pressure injection initiation is to establish a value which is high enough such that protection is provided for the entire spectrum of break sizes and is far enough below normal operating pressure to prevent spurious initiation. Bypass of HPI below 1750 psig, and LPI below 900 psig, prevents }' ECCS actuation during normal system cooldown. I 3-37 )bb9 23l
Draft Technical Specifications Corresponding to Section 11.2.12 1869 232
c. Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit for high reactor coolant system pressure has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. (6)Due to calibration and instrument errors, the safety analysis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting.
The high pressure trip setpoint was subsequently lowered from 2390 psig to 2300 psig. The lowering of the high pressure trip setpoint and raising of the setpoint for the pressurizer Electromatic Relief Valve, from 2255 psig to 2450 psig, has the effect of reducing the challenge rate to the pressurizer Electromatic Relief Valve while maintaining ASME Code Safety Valve capability. (7) The low pressure (1900 psig) and variable low pressure (11.75 TOUT _ 5103) trip setpoint were established to maintain the DNB ratio greater than or equal to 1.3 fore those design accidents that re-sult in a pressure reduction (3,4). The B&W generic ECCS analysis however, assumed a low pressure trip of 1900 psig and is therefore the basis of low pressure reactor trip. Figure 2.3-1 shows the high pressure, low pressure, and variable low pressure trips.
d. Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperature in the operating range.
The calibrated range of the temperature channels of the RPS is 520 to 620' F. The trip setpoint of the channel is 619'F. Under the worst case environment, power supply perturbations, and drift, the accuracy of the trip string is llF. This accuracy was
rived at by summing the worst case accuracies of each module. Thir _s a conservative method of error analysis since the normal procedure is to use the root mean square method.
Therefore, it is assured that a trip will occur at a value no higher than 620*F even under worst case conditions. The safety analysis used a high temperature trip set point of 620*F. The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drif t, linearity, repeatability, etc. This does not imply that the equipment is restricted to operation within the calibrated range. Additional testing has demonstrated that in fact, the temperature channel is fully operational approximately 10% above the calibrated range. 2-7 i869 233
Since it has been established that the channel will trip at a value of RC outlet temperature no higher than 620*F even in the worst case, and since the channel is fully operational approximately 10% above the calibrated range and exhibits no hyst asis or foldover charac-teristics, it is concluded that the insh nent design is acceptable,
e. Reactor building pressure The high reactor building pressure trip setting limit (4 sig) pro-vides positive assurance that a reactor trip will occur in the un-likely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

f. Shutdown bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing cer-tain segments of the reactor protection system. The reactor protec-tion system segments which can be bypassed are shown in Table 2.3-1.
Two conditions are imposed when the bypass is used:
1. By administrative control the nuclear overpower trip set point must be reduced to a value < 5.0 percent of rated power during reactor shutdown.

2. A high reactor coolant system pressure trip set point of 1720 psig is automatically imposed.
The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed. This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The overpower trip set point of i 5.0 percent prevents any significant reactor power from being produced when performing the physics tests. Sufficient natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating. Refereces (1) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.2 (3) FSAR, Section 14.1.2.7 (4) FSAR, Section 14.1.2.8 (5) FSAR, Section 14.1.2.6 (6) Technical Specification Change Request No. 31, January 16, 1976, and Technical Specification Change Request No. 84, June 23, 1978. (7) " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant," Volumes I & II, Babcock and Wilcox, May 7, 1979. (8) "ECCS Analysis of B&W's 1"7-FA Lowered Loop NSS," BAW-10103, Rev. 2, Babcock and Wilcox, April 1976. 2-8
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SUPPLEMENT 1, PART 2 OUESTION
38. With regard to a recent event at Oconee Unit 3 in which certain indications in the control room became unavailable, discuss the vulnerability of TMI-l to a similar malfunction. Also, consider modifications which would reduce the potential for this type of es_nt.
RESPONSE
The electrical distribution system to the ICS at TMI-l has been reviewed and it has been concluded that the ICS is vulnerable to an event of the type that took place recently at Oconee-3. This would occur upon loss of inverter lA and failure of the automatic transfer switch. A modification is being made which will ensure that the operator can take prompt action to minimize the consequences of such an event. }bnual transfer contactors, operable from the control room will be provided to transfer the ICS supply bus (Distribution Panel ATA) from the inverter bus (Distribution Panel VBA) to the Regulated AC Supply (Distribution Panel TRA). The contactors will be interlocked to prevent paralleling of the supplies. An alarm will be provided to indicate that voltage to Distribution Panel ATA has been lost. This modification will be installed before restart. The following longer term actions are being pursued:
1. Investigation of the feasibility of modifying the ICS to allow some of ICS power feeds to be supplied from a separate source.

2. Further review of the electrical distribution system to determine whether the reliability of the ICS/NNI power supplies can be inproved.
i8609 237 Am. 11
SUPPLEMENT 2 THREE MILE ISLAND NUCLEAR STATION UNIT 1 OPERATIONAL QUALITY ASSURANCE PIAN 1869 238 Am. 11
i
?M. M f
i x V5 , F .k.L .
-1 .
r u w w: w f Generation Group O P E RA-' ON A!_ l, QJAL Y ASS JRANCE PLAN
, IOR .
THREE MILE ISLAND i NUCLEAR STATION d 1869 239 L - - . -
STATEMENT OF POLICY AND AUTHORITY It is the policy of the Metropolitan Edison Company to operate the Three Mile Island Nuclear Station so as to ensure the safety and health of the puolic and the person-nel on site. It is also the policy of the Metropolitan Edison Com-pany to comply with the requirements of the Code of Federal Regulations, the NRC Operating Licenses and the applicable codes, guides and standards with respect to operation, inservice inspection, refueling, maintenance, procurement, repair and modification of the Station. The Senior Vice President of Metropolitan Edison Com-pany has the overall responsibility for establishing the policies, goals and objectives of the Quality Assurance Program. He utilizes the support and services of the GPU Service Corporation in implementing the requirements of the Quality Assurance Program. The Senior Vice President of Metropolitan Edison Company is also the Vice President -Generation of the GPU Service Corporation. This dual role gives him the authority to manage and control the activities of the TMI Generation Group. The TMI Genera-tion Group was formed to strengthen the management of and to provide greatly increased technical resources to the Three Mile Island Nuclear Station. The Senior Vice President Metropolitan Edison /Vice President GPU Service Corporation is the head of the TMI Genaratior Group. In this position, he reports directly to the President of GPU Service Corporation (wno is also President of General Public Utilities Corp.) and President of Metropolitan Edison. This reporting structure provides a direct line of authority from the Chief Operating Officer of these three companies to the activities at the Three Mile Island Nuclear Station. A primary objective of the group is to ensure safe operations by means which include strict ad-herence to NRC Regulations, Technical Specifications and Plant Procedures. The Director-Reliability Engineering who reports directly to the Senior VP Met-Ed/VP GPUSC provides, by way of the Quality Assurance Department, the staf f necessary to develop and maintain the Quality Assurance Program con-sistent with the applicaole Federal and State requirements and to verify the implementation of the Program. Rev. 8 1869 240
The Manager-Quality Assurance, who reports airectly to the Director-Reliability Engineering, has the overall authority and organizational freedom to identify quality assurance or management control problems and provide rec-ommended solutions. This authority and responsibility includes stop work authority in activities associated with operations, maintenance, repair, modification, refueling and manufacturing at or for the Three Mile Island Nuclear Station. With regard to the stoppage of work including the recommendation that an operating nuclear unit be shut down, the Manager-Quality Assurance has direct access to the Senior VP Met-Ed/VP GPUSC and shall use this path when differences of opinion within the Generation Group regard-ing quality arise. The ef fectiveness of any Quality Assurance Program is dependent upon the individuals who implement the program. Accordingly, all personnel of the General Puolic Utility System and their contractors must comply with tr.e appli-cable requirements of this Quality Assurance Program. All members of management must give full support to maintain-ing an effective quality program as defined in this Plan. The Quality Assurance Program, as described in this Plan, is approved for implementation at Three Mile Island Nuclear Generation Station. Date Presioent, Met-de (Acting) Date Presioent, GPU Service Corp. Rev. 8 1869 241
TABLE OF CONTENTS Statement of Policy and Authority Introduction 1.0 Organization 2.0 Quality Acsurance Program 3.0 Control of Documents and Recoras 3.1 Instruction, Procedures and Drawings 3.2 Document Control 3.3 Quality Assurance Records 4.0 Design Control 5.0 Procurement and Material Control 5.1 Control of Procurement 5.2 Identification and Control of Materials Parts and Components 6.0 Control of Station Activities 6.1 Policy 6.2 Requirements 6.2.1.1 Control of Inspection 6.2.1.2 Control o f Special Processes 6.2.1.3 Test Control 6.2.1.4 Control of Measuring and Test Equipment 6.2.1.5 Handling, Storage and Shipping 6.2.1.6 Inspection, Test and Operating Status 6.2.1.7 Fire Protection 6.2.1.8 Plant Security 6.2.1.9 Housekeeping and Cleanliness 1869 a2st28
6.2.1.10 Equipment Control 6.2.1.11 Control of Construction, Maintenance (Preventive / Corrective), and Modifications 6.1.1.12 Procedural Requirements 6.2.1.13 Control of Surveillance Testing and Inspection . 6.2.1.14 Radiation Control 6.3 Responsibilities 7.0 Control of Radioactive Wastes 8.0 Control of Corrective Actions and Nonconformances 9.0 Audits 243 1869 Rev. 8
OPERATIONAL QA PLAN Introduction This Quality Assurance Plan is formatted in such a manner to provide all users with a functionally workaole document. It is structured to describe how the Quality Assurance Program is to be functionally implementeo with due regard to the safety and health of the public and the personnel onsite. The plan contains a descriptiun of the organizations responsiole for the implementation of the Quality Assurance Program (Section 1) and an overall descriptian of the Program (Section 2). The remaining sections are structured in a functional manner. The requirements for administrative control which are generic and apply to all subsequent sections are as follows: control of documents and records are contained in Section 3.0; control of design is contained in Section 4.0; control of materials and services through 'h3 control of procurement activities is contained in Sectian 5.0. Sections 6.0 and 7.0 contain the program require.ments for those direct and supportive activities, associated with the operation and safety of the plant; construction and/or modifications associated with corrective maintenance, plant improvement, and/or repair; and the processing and transportation of radioactive wastes. Specific require-ments such as tnose including control of measurement and test equipment, inspection, special processes, test con-trol, and status of inspections, tests and operations are included therein. Sections 8.0 and 9.0 again apply to all functions covered by the scope of this Quality Assurance Program. Section 8.0 addresses the subjects of identification and disposi-tion of nonconformances associated with all aspects of the program. In adoition, this section contains the manage-ment controls provided for evaluating, collectively all nonconformances and determining what corrective actions should be taken to preclude their recurrence. Section 9.0 contains the requirements and administrative controls applicaole to safety reviews and audits. Appendices A, B, and C contain additional Quality Program requirements associated with the functional areas discussed in the plan. 1869 244 Rev. 8
1.0 Organization 1.1 Policy The responsibility for the safe ar.d economical operation of the TMI Nuclear Statico rests with the TMI Ccneration Group. As identified in the Statement of Policy and Authority, the TMI Generation Group utilizes the support and services of the GPUSC in implementing the Operational Quality Assurance Program. The organizations having responsibility for the operation, maintenance, refueling, inservice inspection, modifications and repair of TMI Units 1 and 2 include the THI Genecation Group, under the direction of the Senior Vice President of Met-Ed/Vice President GPUSC, and Materials Management under the direction of the Vice President, GPUSC Materials Management. It is the policy of the Metropolitan Edison Company (Met-Ed) to meet the quality assurance requirements of Nuclear Regulatory Commission as presented in 10 CFR 50, Appendix 8, and other applicaole regulatory guides, codes, and standards pertinent to the operation of the Three Mile Island Nuclear Station. The Pro-gram, which is described in the fo_ lowing sections, shall be implemented throughout the operation phase in documented approved poll-cies, procedures, instructions which comply with this Plan and the design specified in the license application. 1.2 Organization The structure of the organizations responsible for the operation, maintenance, modification, repair, inservice inspection and re fueling o f the TMI Nuclear Station is illustrated in Figure 1. The overall organization chart is provided to illustrate the interfaces between the various departments and to identify those normally located on-site and o f f-site. 1.2.1 President In the Statement of Policy and Authority, the Presidents of Met-Ed and GPUSC have identified that the safe operation of TMI is the respon-sibility of all persons performing activities which affect quality and that management will Rev. 8 1869 245
give full support to the proper and complete implementation of the OQ. Program. Lines of authority and responsibilities havt' been established for maintaining ond implementing the Program; for provioing independent verifi-cation of the activity; and for appraising management of the ef fectiveness of the Program. The responsibilities for establishment, imple-mentation and measurement of the ef fectiveness of the Program are assigned to the TMI Genera-tion Group with support from Materials Manage-ment. In providing support to the TMI Generation Group, Materials Manage?ent will comply with the requirements of thls Opera-tional Quality Assurance Plan as defined further in Subsection 1.2.7. 1.2.2 Senior Vice President Met-Ed/Vice President GPUSC (Figure 1) The TMI Generation Group, under the direction of a Senior Vice President Met-Ed/Vice Presi-dent GPUSC, is responsiole for estaolishing the policies, goals and objectives of the OQA Program and for providing the on-site and off-site staffs necessary to i m p l e,n e n t the Program and provide the verificaticn necessary to assure the effectiveness of the Program. The responsibilities are carried out through four (4) Directors and one (1) Manager. The Senior Vice President Met-Ed/Vice Presi-dent GPUSC has the responsibility for direct-ing and assuring that the management controls and the quality assurance program necessary for the safe operation of THI are established and effectively executed. To this end, they include providing the management oersonnel, the staff support, and the appropriata invest-ment of time and financial resources to enable the designated individuals to properly execute tneir responsibilities. The Senior Vice President Met-Ed/Vice Presi-dent GPUSC is also responsible for assessing the effectiveness of the program and for assuring that decisions af fecting nuclear safety are made at the proper level of respon-sibility ano with the necessary technical advice ano review. This responsioility shall be met, as a minimum, oy:
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a. Assuring that an independer.t managment review of the effectiveness of the OQA Program is conducted annually.
The results of this review snall be documented in a report.
b. Receiving anc reviewing summaries of reports prepared by the organizations performing independent sudits and safety review. These organizations include the Quality Assurance Depart-ment, the Nuclear Safety Evaluation Department, the Generation Review Committee (GRC), and the Plant Operations Review Committee (PORC).
To the extent necessary to assure the health and safety of the public and the employees and contractors working at TMI, the Senior Vice President Met-Ed/VIce President GPUSC shall have the authority, the organizational freedom and the responsibility to order the shutdown of one or both of the operating units. 1.2.3 Director - TMI Unit 1 The Director - TMI Unit 1 is responsible for the overall safety of the TMI Nuclear Station, for ensuring that the applicable procedures for the management control and quality assurance program activities are implemented in the conduct of operations, preventative and corrective maintenance, replacment, modifica-tion, re fueling , engineering support, in-service inspection, radiation protection and control of radioactive wastes, training, plant security, and for ensuring that those activi-ties are performed in accordance with the provisions and limitations set forth in the licenses and permits of the juriscictional agencies of Federal, State, and local govern-ments. He is also responsible for ensuring that the operations organization is adequately staffed and that the personnel are adequately trained and qualified to perform their assigned tasks. Additionally, the Vice Presi-cent-Nuclear Operations is responsiole for activities such as support services and logis-tics and for the planning and scheculing of I-3 Rev. 8 1869 247
plant operations such as start-up and test, refueling and planned outages, and produc-tivity of the generating station. The Director - TMI Unit 1 gives his fullest support to the quality assurance requirements set forth in this Operational Quality Assurance Plan, assuring compliance to the fullest degree by his staff. 1.2.3.1 Manager-Plant Engineering The Manager-Plant Engineering is responsible for maintaining technical liaison and coordi-nation between operating shift personnel and the technical support engineering staff. This is accomplished by providing on-sh ft engi-neers to the Operations staff for direct tech-nical coverage of the plant reactor perfor-mance and associated safety systems in order to improve the safety of unit operations and maintenance performance. In addition, the Manager-Plant Engineering is responsible for in-plant engineering support in the nuclear, mechanical, instrumentation and control, and electrical engineering areas. The Manager-Plant Engineering is also responsible for plant chemistry, fire protection and engineering input for procurement of items and services important to safety. 1.2.3.2 Manager-TMI Unit 1 The Manager - TMI Unit 1 is responsible for the day-to-day operation of Unit 1. He will have a Shift Foreman directing the operations of each shift through the Control Room Operators and Auxiliary Operators; a maintenance force under the direction of a Supervisor-Maintenance, covering the areas of electrical, mechanical and instrument control maintenanct and surveillance, will also report to the Manager. Additionally, tha Manager - TMI Unit 1 is responsible for the coordination of start-up and test evaluations. 1.2.3.3 Manager-Radiological Controls The Manager-Radiological Controls is respon-sible for the personnel, procedures and admi-nistrative controls of the radiation protec-tion programs. He provides the administrative 8 I-4 j g4g } 4gev.
and technical guidance applicable to opera-tionc in the areas of radiation protection, radioactive waste, respiratory protection, health physics engineering including ALARA programs, and dosimetry control. The Manager-Radiological Controls is respon-sible for providing and maintainin] up-to-date procedures controlling the activitias of the department, providing training of all Unit personnel in the basic rules of radiation protection, providing adequate staffs of trained personnel to perform the duties of radiation protection, implementing the "as low as reasonably achievable" policy and making it a formal part of the Radiation Protection Program, and assisting in the development, training and implementation of the Station Emergency Plan. 1.2.3.4 Manager-Training The Manager-Training is responsible for over-all administrative control and coordination of all training activities. The purpose of the training program is to develop and maintain an organization fully qualified to ensure the continued safe and efficient operation and maintenance of the Units. In this position, the Manager-Training is responsible for the training of personnel requiring an NRC License and the training of personnel not reauiring an NRC License. The former includes an accele-rated operator retraining program and the continuous program of training and examination of operators necessary to maintain their NRC Operator's License; the latter includes management and technical training in the areas of Radiation Protection, Industrial Safety, Plant Security, Quality Assurance, Fire Protection, "31ntenance and specialized areas as may be loentified by management. The Manager-Training is responsible for the coordination and scheduling of the training, assuring that the training is given by indivi-duals or organizations qualified in the speci-fic subjects, and maintaining records of the training provided and the attendance. I-5 Rev. 8 1869 249
1.2.3.5 Manager-Administration and Services The Manager-Administratica and Services is responsible for coordination of facility func-
'lons such as office management, facilities, personnel, station security, and the Station Document Center. Relative to the activities applicable to this Quality Assurance Plan, these responsibilities include establishing, supervising and operating the Station Document Center; providing and maintaining up-to-date procedures for controlling the distribation of documents, and the collectio4, indexing and storage of records; providing the staff necessary to fulfill these responsibilities include establishing, supervising and operat-ing the Station Document Center; providing and maintaining up-to-date procedures for con-trolling the distribution of documents, and the collection, indexing and storage of records; providing the staff necessary to fulfill these responsibilities, and ensuring that staff is adequately trained and qualified to perform their assigned tasks.
1.2.4 Director-Technical Functions The Director-Technical Functions reports directly to the Senior Vice President Met-Ed with responsibility for the ditalled develop-ment, direction and overall coordination of all engineering activities. He is rasponsible to assure compliance and implementation of tt.2 Quality Assurance Program requiremants appli-cable to technical support activities, Tech-nical support includes various disciplines such as mechanical, civil, electrical and instrumentation, nuclear, and plant opera-tions. He is responsibile to develop and control the Quality Classification List (QCL). Additionally, he is responsible for nuclear fuel management, process computer, control and safety analysis, and plant operational analysis. The Director-Technical Functions end his staff give full support to the TMI Operational Assurance Plan set forth herein, tnereby assuring that all work performed under their cognizance will conform to and support the requirements as applicable to their activities. I-6 Rev. 8 1869 250
1.2.4.1 Manager-Engineering and Design The Manager-Engineering and Design is respon-sible for providing technical support for the operations of the TMI Nuclear Station. He is responsible to assure compliance with the implementation of the Quality Assurance Program requirements applicable to engineering and design activities. He will assure the maintenance of technical capaollity in the various disciplines, such as general mechani-cal, civil, electrical and instrumentation, and engineering mechanics. The department will review, and where appropriate, approve the work of Architect / Engineering Organiza-tions, and will perform basic engineering and design for modifications. The department has capabilities in the following areas:
a. Engineering Mechanics

b. t'echanical Systems

c. Mechanical Components

d. Electrical Power & Instrumentation

e. Design & Drafting Additionally, he is responsible for the iden-tification and classification of materials and activities important to safe.ty, and the devel-opment and control of the Quality Classifica-tion List.
1.2.4.2 Manager-Sv-tems Engineering The Manager-Systems Engineering is responsible for technical support in the areas of nuclear fuel management, process computer, control and safety analysis, and plant operational analy-sis. His department provides capal-ilities in the following functional areas:
a. Nuclear Analysis and Fuels D. Process Computers

c. Control and Safety Analysis

d. Plant Analysis I-7 8 i 869 Obv.
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He is responsible to assure compliance with and implementation of the Quality Assurance Program requirements applicable to Systems Engineering activities. 1.2.4.3 Project Engineering Manager The Project Engineering Manager is responsible for the coordination, staffing and directing of engineering projects that are assigned by the Director-Technical Functions. These activities will vary, depending on the scope and purpose of the assigned project. These responsibilites generally include providing the technical support necessary for a stuor, an evaluation or a modification and include the coordination of the Departments within Technical Functions with those of Plant Opera-tions. 1.2.5 Director-Environment, Health and Safety The Director-Health and Safety reports
.directly to the Senior Vice President Met-Ed and is responsible for the development, direc-tion and overall coordination of the environ-mental, regulatory, water resources, offsite radiological healtn and safety efforts at THI in compliance with the TMI Quality Assurance Program. His responsioilities include:
a. Developing and implementing Environ-ment, Health and Safety Group proce-dures covering items such as safety evaluations, and others as required to fulfill the responsibilities of this Plan.

b. Concurring with important to safety Design Criteria Documents from the standpoint of having addressed all applicable regulatory requirements and licensing commitment.

c. Exercising project control of amendment requests to the Safety and Environ-mental Technical Specifications and the FSAR in accordance with 10 CFR.

d. Developing biological monitoring pro-grams and special studies in the Environmental Technical Specifications
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to quantify the impact of Station operation on the environment.
e. Performing an environmen'tal evaluation of proposed modifications, including publishing of environmental reports.

f. Maintaining liaison between the TMI Generation Group and NRC's Project Management regarding licensing and environmental issues which are appli-cable to operating facilities.
1.2.6 Director-Reliability Engineering The Director-Reliability Engineering nas the overall authority and direct responsibility for all Quality Assurance Department activi-ties as defined in this plan. These activi-ties include, but are not limited to:
a. Development, distribution, and main-tenance of tne Quality Assurance Plan,

b. Assessing program implementation and evaluating its effectiveness,

c. Identification of quality problems,

d. Initiation and recommendations of cor-rective actions for quality related problems.
Additionally, he is responsible for providing Quality Assurance and Quality Control support services such as laboratory analysis, safety review (Nuclear Safety Evaluation), audits and reliabiliaty information systems. The Director-Reliability Engineering etilitizes the following management staf f members in carrying out his responsibilities: Manager-Quality Assurance Manager-Nuclear Safety Evaluation Manager-System Laboratory 1.2.6.1 Manager-Quality Assurance (Figure 2) The Manager-Quality Assurance Department (QAD) has the functional authority, independence and I-9 ) bbh R25.38
responsibility to verify the ef fective imple-mentation of the administrative controls and compliance to the Quality Assurance Program during the operational phase of TM! Nuclear Station. The Manager of QAD reports directly to the Director-Reliability Engineering. Additionally, he has direct unencumbered ac-cess to the Senior Vice President of Met-Ed, the Vice President-Materials Management, and the Director-TMI Unit 1 with regard to quality activities. This reporting relationship has been esta-blished to provide the quality assurance organization with suf ficient independence from the influence of costs and schedules to be able to effectively assure conformance of operational Quality Assurance Program require-ments. Figure 2 identifies the Quality Assurance Depertment organizational elements which function under the Quality Assurance Program. The Manager-QAD has no duties or responsibilities unrelated to Quality Assurance that would prevent his full atten-tion to Quality Assurance matters, and he has authority:
a. To evaluate the manner in which all activities, both onsite ano offsite are conducted, with respect to quality, by means of review, survey, audit, sur-veillance, monitoring, and inspection,

b. To perform evaluations on a planned and periodic basis to verify that the Quality Assurance Program is being effectively implemented.

c. To identify quality problems, and to initiate, recommend or prov!de solu-tions through designated channels to verify implementation of resolutions.

d. To stop work or further processing, delivery, or installation of noncon-forming material, to stop work on non-conforming activities, to initiate unit shutdown recommenoations and to ootain unit shutdown with appropriate upper management concurrence as described in applicable Quality Assurance procedures.
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The specific responsibilities of the Manager-QAD, include the following:
a. Provide for the review and acceptance of the Quality Assurance Program of contractors providing services affecting quality and of vendors sup-plying materials, parts, or components covered by the scope of this Quality Assurance Program,

b. Provide for review and acceptance of procedures prepared by other TMI or-ganizations when these procedures con-trol or exercise an effect upon items and activities important to safety.

c. Provide direction and management of the QAD.

d. Provide a working interface and com-munication with the TMI Generation organizations, consultants, contrac-tors, vendors, and others with respect to QA matters. Additionally, in con-junction with the licensino organiza-tion, he shall provide a working interface and communications with the NRC with respect to QA matters,

e. Provide, as applicable, planned and periodic audits, monitoring, sur-veillance, and inspections of organiza-tions, contractors, and vendors performing work functions important to safety,

f. Establish and assure the ccotinuous implementation of an indoctrination and training program for QA and QC per- -
sonnel and assure that a quality assurance indoctrination is provided to appropriate personnel outside the Quality Assurance organization.
g. Issue periodic reports to the Director-Reliability Engineering, and the Director-TMI Unit 1 on the status of quality activities, and oring to thier attention immediately any significant quality-related problem or deficiency.
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h. Provide for quality assurance review and acceptance of design and engi-neering documents, as delineated in the detailed procedures.

i. Provide for quality assurance review and acceptance of procurement documents generated for the acquisition of materials and services within the scope of the program.

j. Provide for and maintain quality assurance records generated by QAD until such time as they are turned over to document control for storage.
The Manager-Quality Assurance shall have, as a minimum, a baccalaureate degree in Engineering or Science, with at least five years of ex-perience in nuclear power plant operations and supporting activities. 1.2.6.2(a) Quality Assurance Design and Procurement Manager The Quality Assurance Oesign and Procurement Manager is responsible for establishing quality programs and inspection requirements in support of design and procurement activi-ties in compliance with the TMI Quality
.ssurance Program. These activities include, but are not limited to:
a. Review and approve contractar and ven-dor quality programs for those supply-ing services or items important to safety.

b. Reporting quality trends to his super-visor and to the cognizant purchasing or contract manager,

c. Review and accept design control proce-dures prepared by other TMI organiza-tions when these procedures control or exercise an effect upon systems, com-ponents, or activities important to safety.
1869 256 I-12 Rev. 8
1.2.6.3(b) Quality Assurance Manufacturing Assurance Manager The QA Manufacturing Assurance ~ Manager is responsible to perform the necessary postaward quality-related activities in compliance with the THI Quality Assurance Program required to assure that vendor products are designed, manufactured, tested and/or inspected in accordance with the procurement specifica-tions. These activities include post-award surveys and surveillances, and source inspec-tions. He is responsible for the coordination with the QA Modifications / Operations Section to assure that Manufacturing Assurance discre-pancies are available to the receiving inspec-tors and the cognizant purchasing or contracts manager. Additionally, he is responsible for providing the Design / Procurement Assurance Section with the results of Manufacturing Assurance activities and recommendations relative to the acceptaoility of a vendor. 1.2.6.4(c) Quality Assurance Modifications / Operations Manager The Quality Assurance Modifications / Operations Manager is responsible for monitoring the implementation and effectiveness of the Quality Assurance Program onsite. These ac-tivities include the establishiment of ade-quate site monitoring and inspection programs necessary to verify conformance to Quality Assurance Program requirements. In addition, he is responsible to review site procedures from a QA/QC standpoint and to provide nondes-tructive examination support for TMI. He reports directly to the Manger of Quality Assurance and, he periodically reports on the implementation and effectiveness of the Operational Quality Assurance Program to the Director - TMI Unit 1. He has the authority and organizational freedom to identify quality assurance problems, provide or recommend solutions, and verify implementation of solutions. He has the authority to stop work on all important to safety activities associated with the onsite TMI Nuclear Station Op? rational QA program. He is responsible to notify appropriate TMI Station management and I-13 Pev. 8 1869 257
the Manager of Quality Assurance immediately of any condition that warrants operational shutdown of a nuclear unit as defined in ap-propriate QA0 procedures. The' Quality Assurance Modifications / Operations Manager is assisted in carrying out his responsibilities by an Operations Quality Assurance Supervisor, a Quality Control Manager and their associated staffs located onsite. 1.2.6.5(d) Quality Assurance Methods /Ocerations/ Audits Manager The Quality Assurance Methods / Operations /Au-dits Manager is responsible for maintaining the Quality Assurance Plan and all those pro-cedures applicable to the activities of the QAD. He is responsible, therefore, for coor-dinating the activities associated with the requirements of the Quality Assurance Plan, including interpretations. In addition, he is responsible for coordinating Quality Assurance indoctrination and training to employees and contractors conducting independent evaluations and assessments of the Program's implementa-tions by performance of internal audits. The Quality Assurance Methods / Operations /Au-dits Manager maintains a full-time staff of quality assurance engineers and qualified quality auditors at both the corporate and site offices. The audit activities and the results of the audits are provided tc the audited organization and to the Safety Review Groups who provide the management 1ssessments of the significance of the audit findings and the effectiveness of the Quality Assurance Program. 1.2.6.6(e) Materials Technology Manager The Materials Technology Manager directs and supervises the offsite engineering organiza-tions which have the responsibility for acti-vities in the estaolishment of requirements for welding, inservice inspection, materials, and materials evaluations. Materials Techno-logy provides NDE and ISI program flow analy-sis and reporting, technical requiraments for repair and repair program and related I-14 Rev. 8
corrective action recommendations ta Engineer-ing. Additionally, he has a staff Tunction to support manufacturing and the evaluation of system materials technology problems. He is directly responsible for the implementation and compliance with the Quality Assurance Program requirements applicable to his areas of responsibility. The specific disciplines included in the Materials Techno?.ogy sections are:
a. Nondestructive Examination

b. Inservice Inspection

c. Material; Engineering

d. Welding Engineering

e. Metallurgical Analysis 1.2.6.7.1 Minimum Qualifications of Quality Assurance Personnel The qualification requirements ano excerience levels for key Quality Assurance personnel are such as to assure competence commensurate with the responsibilities of the position. Quality managers and supervisory personnel are re-quired to have a degree in Engineering (or equivalent) and experience in a position having responsibility for the performance of quality activities. The degree requirement may be waived for personnel with exceptional qualifications and a minimum of seven (7) years related experience.
1.2.6.8 Manager-System Laboratory The Manager-System Laboratory is responsible for the administration and operation of the Environmental and Operational Chemistry Analyses Section of the System Laboratory in compliance with the TMI Quality Assurance Program. This section provides the central-ized laboratory analyses services for TMI. The specific responsibilities of the Manager-System Laboratory include the following: - 1869 259 I-15 Rev. 8
a. Perform analysis of water and waste-water samples submitted by the generating station,

b. Prepare calibration standards for the labs at the generating stations,

c. Monitor the analysis capabilities of the labs at the generating stations through aucits and independent analysis of samples,

d. Assist station personnel in unusual operations such as chemical cleaning,

e. Provide consultation on equipment startup and performance, as requested,

f. Perform chemical analysis of fuels, lubricants, insulating fluids and ion exchange resins,

g. Ensure that the laboratory is ade-quately staffed and that the laboratory personnel are adequately trained and qualified to perform their assigned tasks,

h. Ensure that the System Laboratory meets the applicaole requirements of the Quality Assurance Program,

i. Decelop and implement laooratory proce-dures covering the control of the laboratory activities and the records documenting the results of the analysis.
J. Provide support to Materials Technology regarding material specimen oreparation and testing. 1.2.6.9 Manager-Nuclear Safety Evaluation The Manager-Nuclear Safety Evaluation is responsible for the development, direction and supervision of the Nuclear Safety Evaluation Department in compliance with the TMI Quality Assurance Program. The function of this group is to review the broad range of activities, practices and conditions which may have an adverse effect on quality, to assess thgf9 94n safety significance of the conditions IMJ to C v V I-16 Rev. 8
make recommendations to the appropriate levels of management for corrective action to pre-clude repetition. The activities of the Nuclear Safety Evaluation Department will include the receipt and review of al] docu-ments and reports identifying conditions ad-verse to quality (Audit Reports, Noncoa-formance Reports, Surveillance / Inspection Reports, Reportable Occurrences, NRC Inspec-tions, etc.); analyzing the conditions reported, both individually and collectively; identifying the safety significance of the conditions and reporting, at intervals not to exceed six (6) months, the results of the evaluations to the Senior Vice President of Met-Ed and the Director-TMI Unit 1. Addi-tionally, the Nuclear Safety Evaluation Department working with Systems Engineering will evaluate the operational experience of other nuclear power stations to improve plart operational status and derive benefit from other stations experience. 1.2.7 Vice President-Materials Management The Vice President-Materials Management, who reports to the President-GPUSC, via the Exe-cutive VP GPUSC, is responsible for assuring that the technical and quality requirements, as established by the Generation Group, are incorporated into orocurement documents. He is directly responsiole for the compliance and implementation of the TMI Quality Assurance Program with regard to procurement activi-ties. His responsibilities will be adminis-tered through the following: Director-Materials Management Systems Manager-Contracts, TMI Site 1.2.7.1 Director-Materials Management Systems The Director-Materials Management Systems is responsible for the development and management of a procurement system in compliance with the TMI Quality Assurance Program including:
a. the establishment and implementation of a procurement control process, I-17 Rev. 8 1869 261

b. the incorporation of quality assurance program requirements, as ideatified by ll4 engineering documents, into procurement documents,

c. the coordination of quality assurance activities in the procurement process.
He reports administratively to the GPUSC Vice President - Materials Management, and func-tionally to the Senior Vice President - Met-Ed for the priorities accorded to the TMI Generation Group's requirements. His responsibilities will be adminisered through the following: Manager - Field Warehousing, TMI Manager - Field Procurement, TMI 1.2.7.1.1 Manager-Field Warehousing, TMI The Manager-Field Warehousing, TMI is respon-sible .for maintaining an inventory, initiating requisitions for inventory reorder, receiving both direct turnover and inventory items, maintaining adequate storage space and facilities, and issuance of material from storage. 1.2.7.1.2 Manager-Field Procurement, TMI The Manager-Field Procur' cent, TMI is respon-sible for all TMI purchasing and expediting activities including the following:
a. the implementation of an approved procurement control process,

b. the receipt, review, recording and tracking of purchase requisitions,

c. the incorporation of engineering re-quirements into purchase orders,

d. compliance and implementation of the TMI Quality Assurance Program with regard to his areas of responsibility.

e. the preparation and document control of all p, -"ase orders including those for contri I-18 Rev. 8 1869 262
1.2.7.2 Manager-Contracts, TMI Site The Manager-Contracts, THI Site.is responsible for all site-related contracts with respect to the bidding, bid evaluation, award and ad-ministration of construction, maintenance, water processing, decontamination, and waste removal contracts, and those for various tech-nical and other consulting services. He is additionally responsible for the evaluation, validation, dismissal or negotiation, where warranted, of vendor extras, delays and other claims and proposals. He is directly responsible for the compliance and implemen-tation of the TMI Quality Assurance Program with regard to his areas of responsibility. 1.2.8 Manager-Management Services The Manager-Management Services, who reports directly to the Senior Vice President Met-Ed, is responsiale for the administrative control of the Generation Group in compliance with the TMI Quality Assurance Program. In this capa-city, he provides support activities relating to the Quality Assurance Program. The speci-fic responsibilities of the Manager-;4anagement Services include the following:
a. Delegating the preL; ration of the Generation Group Administrative Proce-dures, coordinating their review, approving procedures and ensuring their implementation in accordance with Appendix B of this Plan,

b. Establishing, supervising and operating the TMI Generation Group Occument Centers in accordance with Generation Group procedures.

c. Establishing an effective interface between the Corporate and Station Docu-ment Centers.

d. Supervising and directing the activi-ties for maintenance and control of the off-site TMI Generation Group records.

e. Coordinating personnel mansgement development and professional training.
I-19 Rev. 8 1869 263
The Manager-Management Services accomplishes these responsibilities via the direct super-vision of the activities of the Supervisor-Accounting and Budgets, and Supervisor-Corporate Records Management. The Manager-Management. Services and his staff give full support to the TMI Quality Assurance Program set forth herein, thereby assuring that all work performed under their activity will conform to and support the requirements as applicable to their activity. I-20 Rev. 8 1869 264
f i President FIGURE I
.,t_g3 DtI CENEF ATION CRO'JP Ceneration Crcup Sr. VP Met-Ed -___7-____q l I Director Director Manager Director Director Manager Director Director Reliability Materials Mgmt Contracts Technical Managenent ttI TMI Environment. Engineering Systems TMI Site Functions Services Unit 1 Unit 2 Health &
91fety I e J L______ p_-- i I Manager Manager Manager Project Manager g g Admin. & Unit I Quality Field Engineering g Procurene it
! Manager Services Assurance l TMI a l Q l .. I
& ( CRC l Manager g ( , Manager Engineering Manager Plant Manager Trainirg l l System Manager Field
& Design Engineering Laboratory arehousing l TMI N
Os i U1 Manager Manager I_J Manager Systems Radiologica' _ I flu elea r Engineering Controls PORC S3I*tY
/ Evaluation Located on Site Funct ional Reportir.d
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s ue s t c A cnr aae f rg d ry et uua gi nsn al A asa na QMAM au s r MQ n o o s y i i sg t v l o a r al r r e i oe e p rng pA u eh a OQS t cn aea s MTM p O
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2.0 Quality Assurance Program 2.1 Policy 2.1.1 General The TMI Operational Quality Assurance Program has been established to provide overall qual-ity assurance of operations activities within the scope of the program. Adherence to the requirements of the THI Operational Quality Assurance Program is mandatory for all TMI organizations and for all contractors or ven-dors providing items or services covered under the scope. 2.1.2 Scope The scope of the TMI Operational Quality As-surance Program includes all items and activ-ities considered to be "important to safety." \ This term is intended to be broader than
" safety-related" and encompasses structures, systems, and components (including nuclear.
fuel and radwaste) which have been designated as Safety-Related, Safety Class, IEEE Class IE, Seismic Category 1 Fire Protection. The scope of the Program will include all items required by the following:
a. Title 10, Code of Federal Regulations, Part 50, Appendix A " General Design Criteria for Nuclear Power Plants"

o. Title 10, Cooe of Federal Regulations, Part 50, Appendix B " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants"

c. Title 10, Code of Federal Regulations, Part 71, Appendix E " Quality Assurance for Shlpping Packages for Radioactive Material"

d. United States Nuclear Regulatory Com-mission Regulatory Guide 1.33 " Quality Assurance Program Requirements (Opera-tion)"

e. United States Nuclear Regulatory Com-mission Regulatory Guide 1.143 " Design Rev. 8 1869 267
Guidance for Radioactive Waste Manage-ment Systems, Structures and Components Installed in Light Water Cooled Nuclear Power Plants"
f. United States Nuclear Regulatory Com- s mission Regulatory Guide 1.120 " Fire Protection Guidelines for Nuclear Power Plants" The Program also includes certain non-safety related items when designated by engineering.
Activities which are important to safety shall include, but not be limited to, those activ-ities covered by Appendix A of Regulatory Guide 1.33 and ANSI N18.7. In addition, the requirements of other Regula-tory Guides applicable to operations, mainte-nance, modification, repair, and refueling of a nuclear power plant were considered. Met-Ed position with regard to these Regulatory Guides is included as Appendix C. The deter- i mination that an item or activity is, or is not, important to safety is a design decision governed by approved engineering procedures. Items and activities determined to be impor-tant to safety are defined as those items on the Quality Classification List (QCL) and those activities covered by procedures which have oeen designated during the review cycle as "important to safety." The QCL may also be utilized to record information regarding the level of Quality Assurance Program deemed appropriate for a particular item or service. This information is not required, but may be utilized ,] facilitate procurement and imple-mentation. For new design efforts, such as plant modifications and new construction, the s classification determination is recorded on design criteria documents. New items will oe included in the next appropriate revisions to QCL. 2.1.3 Quality Assurance Plan This Quality Assurance Plan is the primary document wnich provides a description of the program. The program is authorized by the II-2 Rev. 8 s
President of Met-Ed and GPUSC to assure that the appropriate levels of management, as designated herein, are directed to implement the program. The Plan is contrdlled to assure that only the latest approved revision is implemented. The Plan is implemented by approved detailed procedures and instructions. The purpose of this Plan is to establish the principles which, when implemented, will pro-vide that level of quality assurance which is appropriate to each activity affecting quality. It is recognized that the degree of management control or quality assur6nce to De applied varies with different systems and activities, and the degree of applicability of any specific item in this Program will differ from system to system and activity to activity. The degres to which the requirements of this Plan and its implementing procedures are ap-plied will be cased upon the following:
a. The importance of a mElfunct. ion or failure of the item to safety;

b. The design and fabrication complexity g or uniqueness of the item;

c. The need for coecial controls and sur-veillance or mon:toring of processes, equipment and operutional activities.

d. The degree to wnich functional compli-ance can ce demonstrated by inspection or test; and

e. The quality history and degree of stanoarization of the item or activity.
The TMI Generation Group is committed to a comprehensive Quality Assurance Program con-sisting of a three level approach to assure satisfactory and complete implementation of the program commensurate with its requirements g for safety and performance. The Program's foremost considerations are the protection of the general public's health and safety. The three level approach is defined below: Level 1 - activities at this level include independent inspections, checks and tests. II-3 Rev. 8 1869 269
This level of activity may be performed by the Operations Department by surveillance tests, calibration of instruments, radia-tion surveys, analyses of samples, etc., the Quality Control Section by receipt inspection or inspections of modification or corrective msintenance activities, or by contractors as part of their scope of work. In all cases, the activity is per-formed by individuals knowledgeable of the activity being performed and qualified to perform the work. Checklists or data sheets are also used for. documenting the results of the activity and for providing a permanent plant record of the performance of the activity. In all cases where the first level activities involve inspection for purposes of acceptance and/or verification of modifications to safety systems, the activity will be perturmed by personnel who are independent of those per-forming the work. Level II - the activities at this level are primarily th1se of surveillance or moni- , toring and are performed as deemed neces-sary by the (A Modifications / Operations, QA Design and Pr)curement or QA Manufacturing Assurance SP',tions. The level of sur-veillance/ monitoring applied is consistent with the importance of the item to safety. For activities, whereby Quality Control is performing first level inspection, no second level activity will be required. At this level procedures and instructions are established and surveillance records will be completed and maintained. Such surveillance / monitoring normally includes observation of quality ccatrol tests and inspections, obser-vation of significant operations, review of records, verifications of test reports, and direct inspection on a spot check basis. The , organizations performing this activity have the levels of authority, the lines of internal and external communication for management direction, and the properly trained personnel for implementation of these activities. II-4 Rev. 8 1869 270
*' Level III - the purpose of this level of activity is to assure through a comprehen-sive program of review and auditing that the first and second levels'of the program '
are properly functioning. The purpose of - this level is also to establish that all other organizations including Operations, Maintenance, Engineering Materials Management, etc. are prop,erly satisfying all the requirements of the Operational Quality Assurance Program. - At this level procedures and instructions are established including the use of comprehensive checklists third level for documentation activity in accordance of thewith audit re-or quirements o f ANSI N45. 2.12. personnel are included that satisfy Quali fied audit - quirements of ANSI N45.2.23. the re- ' nical experts from areas with administrative Additional tech-reporting outside the function that is being audited will be include Leader deems necessary.dThe as the Audit organization Team
;_n performing this activity has sufficient
authority and lines of internal and external communications management for obtaining the necessary direction. - Appendix A is included to provide a comparison of the sections of the Plan with the require-ments of 10CFR50, Appendix B, 10CFR71, Appendix E, ANSI N18.7, and ANSI N45.2. 2.1.4 4 Quality Assurance Procram Review The TMI ness and Quality Assurance Program ef fective- . ~ implementation uated by independent review is periodically eval-to TMI Generation Group management. groups reporting s groups provide safety review These ational methods. These groups eacq have oper- reports and tech-nical expertise r.ecessary to support their areas of concern. The independent review . .' committees and operational review groups in-clude the Generation Review Plant Operations Review Committee andCommittee, the the Nuclear addition, Sa fety Evaluation Department. In ', the Quality Assurance Department ~ conducts activities which provide management with additional effectiveness and information pertaining to . implementation. 3 II-5 Rev. 8 - 1869 271 -
2.1.5 Training The TMI Quality Assurance Program includes , requirements for formal training programs for personnel performing or verifying activities important to safety. 2.2 Requirements 2.2.1 Quality Assurance Plan The Operational Quality Assurance Plan and any significant revisions shall be approved by the , following: Sr. VP Met-Ed/VP GPUSC VP - Materials Management Director - TMI Unit 1
\
Director - Reliability Engineering Manager - Quality Assurance The Plan includes a Statement of Policy which is signed by the Presidents - GPUSC and Metropolitan ~ Edison. The Statement of Policy provides authorization and evidence of manage-ment commitment to the Quality Assurance Program. Plan revisions which represent significant changes or personnel reassignments of a sub-stantive nature shall be submitted to the Nuclear Regulatory Commission for approval. The Manager-Quality Assurance is responsible for notifying the NRC of all changes to the , Plan within 30 days of the change and for obtaining the required approvals prior to issuance. Plan revisions not considered by the Manager - Quality Assurance to be significant can be issued with approval of the Manager - Quality Assurance and the Director - TMI Unit 1. Copies of Quality Assurance Plan may be dis-tributed as " Controlled" or " Uncontrolled" copies in accordance with the requirements established in Section 3. II-6 Rev. 8 1869 272
2.2.2 Classification The TMI Operational Quality Assurance Program applies to all items on the QCL and activities designated as "important to safety." The QCL will be periodically updated to include new plant modifications or construction, or any changes in classification. The list will oe treated as a controlled document. The QCL will normally list systems and compo-nents, but not parts. For procurement of spare or replacement parts, classification will be on a case by case basis. The deter-minations will not necessarily be added to the QCL. An approved engineering evaluation shall be documented and maintained as a quality assurance record. This does not apply to those items which were originally specified as commerical quality. 2.2.3 Regulatory Commitments A listing is maintained of commitments to regulatory requirements. Each new or revised USNRC Regulatory Guide will be evaluated for applicability and acceptability to TMI. The TMI Generation Group position on each is docu-mented stating the method and degree of com-pliance or the justification for lack of compliance. 2.2.4 Safety Review Safety review groups have been established with primary responsibilities for review of
operational phase activities. In addition to performing regulatory required reviews, these groups provide management with visability ano recommendations for improved plant safety.
2.2.4.1 Generation Review Committee (GRC): The GRC is an off-site organization reporting to the Director-Technical Functions. This g:oup is responsible to provide independent sofety review of operations, nuclear engi-n ering, chemistry and radiochemistry, matal-1 rgy, nondestructive examination, instrumen-tition and control, radiological safety, ~ II-7 Rev. 8 1869 273
mechanical and electrical engineering, rad-waste, administrative controls, quality assurance and other appropriate fields associated with the unique characteristics of TMI. The GRC is responsible for reviewing the following specific subjects:
a. Written safety evaluations of changes in the facility as described in the Safety Analysis Report, cnangas in procedures as described in the Safety Analysis Report and tests or experi-ments not described in the Safety Analysis Report which are completed without prior NRC approval under the provisions of 10 CFR 50.59(a)(1). This review is to verify that such changes, tests or experiments did not involve a change in tiie technical specifications or an unreviewed safety question as defined in 10 CFP 50.59 (a)(2).

b. Proposed changes in procedures, pro-posed changes in t7e facility, or pro-posed tests or experiments, any of which involves a change in the techni-cal specifications or an unreviewed safety question as defined in 10 CFR 50.59(c). Matters of this kind shall be referred to the GRC by PORC follow-ing its review, or by other functional organizational units within the TMI Generation Group, prior to implementa-tion.

c. Changes in the technical specifications or license amendments relating to nuclear safety prior to submittal to the Commission for approval and prior to implementation, except in those cases where the change is identical to a previously reviewed, proposed change.

d. Violations, deviations and reportaole events which require reporting to the NRC in writing within 24 hours, such as:

1. Violations of applicable codes, regulations, orders, technical specifications, license require-ments, internal procedures or II-8 Rev. 8 1869 274
instructions having safety significance.
2. Significant operating abnor-malities or deviations from nor-mal or expected performance of plant structures, systems, or components important to safety.

3. Reportable events, which require reporting to the NRC in writing within 24 hours, as defined in the plant technical specifi-cations.
Review of events covered under this subsection shall include the results of any investigations made and the recom-mendations resulting from such investi-gations to prevent or reduce the proba-
~bility of recurrence of the event.
e. Any other matter involving safe opera-tion of the nuclear power plant which an idependent reviewer deems appro-priate for consideration, or which is referred to the independent reviewers by the onsite operating organization or by other functional organizGtional units within the owner organization.
2.2.4.2 Plant Operations Review Committee (PORC): The PORC is an on-site operations review organization functionally reporting to the Director-TMI Unit 1. This group screens sub-jects of potential concern by reviewing and/or performing preliminary investigations of sub-jects requiring independent review. PORC shall provide, as part of the normal duties of plant supervisory personnel, timely and continuing monitoring of operating activ-ities to assist tne Director - TMI Unit 1 in keeping abreast of general plant conditions and to verify that the cay-to-day operating activities are conducted safely and in accor-dance with applicaule administrative controls. These continuing monitoring activities are considered to be an integral part of tne rou-tine supervisory function and are important to II-9 Rev. 8 1869 275
the safety of plant operation. The Director - TMI Unit 1 in carrying out his responsibility for overall safety of plant operations, shall be responsible for timely referral of appro-priate matters to management and independant reviewers. 2.2.4.3 Nuclear Safety Evaluation Department (NSED): NSED is an independent organization reporting to the Director-Reliability Engineering. It perform evaluations and investigations as assigned by Director-Reliability Engineering. NSED evaluates infcrmation from external sources for applic bility to TMI. They are responsible for evaluations of hardware and sof tware systems which af fect the safe reli-able operation of the plant. NSED personnel support the activities of GRC and PORC and contribute as requested. They interface with the QA0 audit section to assure complete cov-erage and utilization of the audit program. 2.2.4.4 Quality Assurance Department: The normal audit program conducted by the Quality Assurance Department and described in Section 9.0 also provides management with assessment of program status and effectiveness. 2.2.5 Indoctrination and Training Indoctrination and training programs are es-tablished for both on-site and off-site per-sonnel performing important to safety activi-ties by the organizational units responsible for the activities. These programs are imple-mented by appropriate training plans and pro-cedures which assure that:
a. Personnel responsible for performing important to safety activities are instructed as to the purpose, scope, and implementation of manuals, proce-dures, and instructions;

b. Personnel performing important to safety activities are trained and qual-ified in the principles and techniques of the activity being performed; II-10 Rev. 8 1869 276

c. Proficiency of personnel performing important to safety activities is main-tained by retraining, re-examining, or recertifying;

d. The scope, method and objective of the training is documented;

e. Records of training sessions are pre-pared and maintained, including identi-fication of the content, the attendees, and the date the training was conducted.
2.3 Resoonsibilities 2.3.1 Sr. Vice President - Met-Ed The Senior Vice President - Met-Ed shall reg-ularly assess the scope, statur, adequacy and compliance of the Quality Assurance Program to the requirements of 10 CFR 50, Appendix B. This assessment shall be the combined result of:
a. Frequent contact with Quality Assurance Program status through attendance at meetings, and review of periodic status reports on the effectiveness and imple-mentation of the Quality Assurance Program,

b. Performance at least once a year of a preplanned and documented assessment of the ef fectiveness of the Quality As-surance Program to assure that the program meets regulatory requirements, and the policies and directives of TMI. This assessment may be performed utilizing the safety review groups, an independent consultant, or his own staff. Any corrective action which may be deemed necessary as a result of these assessments shall be formally ioentified and tracked through resolu-tion.
2.3.2 Director - Reliabilty Engineering The Director-Reliability Engineering has over-all responsibility for establishment of the Operational Quality Assurance Program. He II-ll Rev. 8 1869 277
also has overall responsibiliity for estab-lishment and management of the Nuclear Safety Audit Department and the Quality Assurance Department, including Methods / Operation / Audit Section. He shall provide periodic status reports to the Sr. Vice President Met-Ed on the Quality Assurance Program. 2.3.3 Manager - Quality Assurance The Manager-Quality Assurance, has the direct responsibility for verifying the effective implementation of the Quality Assurance Pro-gram. He shall establish and implement a formally documented and procedurally control-led program to evaluate and report to the Director-Relicbility Engineering on the ade-quacy and continued effectiveness of the over-all TMI Operational Quality Assurarice Pro-gram. Reports of audits performed by the Quality Assurance Department or their agents, and quality trend analyses based on nonconfor-mance and deficiency repcrts will provide the basis for this evaluation. Corrective action shall be implemented by responsible management as deemed appropriate when analyses reveal adverse quality trends. These actions may involve specific actions to provide compliance with the Quality Assurance Program, and may include follow-up system attribute audits and even revision to the TVI Operational Quality Assurance Program. Implementation and closecut of corrective actions shall be effectively monitored by the Manager-Quality Assurance to assure timely correction and compliance. The Manager-Quality Assurance is responsible for the contents of Quality Assurance Plan and for ensuring that the Quality Assurance Plan is modified and updated as standards, regula-tion, requirements and experience dictate. Proposed revisions to the Plan may be sug-gested by TMI Generation Group personnel by submitting the request, in writing, to the Manager-Quality Assurance for review and action, as applicable. The Manager-Quality Assurance is responsible for the monitoring, surveillance and auditing of Qu ality Assurance Program implementation. II-12 Rev. 8 1869 278
He is also responsible to provide the required training and qualification of Quality Assurance Department personnel. 2.3.4 Manager - Engineering & Design The Manager of Engineering & Design is respon-sible for development and maintenance of the QCL. He solicits input and coordinates with affected organizations to assure a uniform approach to classification of items and activ-ities important to safety. 2.3.5 Generation Group Directors and Managers Management personnel in each department are responsible for Quality Assurance Program implementation. They are further responsible for development of procedures, for scope of involvement, for activities important to safety, and for training and indoctrination of personnel. 2.3.6 External Organizations Quality Assurance Programs and implementing procedures for suppliers or contractors pro-viding materials and services for the TMI ~ Nuclear Station which are covered under the scope of this Quality Assurance Program shall be subject, when required, to review and acceptance by the Quality Assurance Department prior to the commencement of any important to safety activity. Procurement documents shall require, and the Quality Assurance Department shall assure, through their review and audit, that supplier and contractor Quality Assurance programs comply with the commitments of this document. 2.4 Resolution of Disputes Resolution of disputes involving quality, arising from a difference of opinion between QA/QC personnel and other organization (engi-neering, procurement, manufacturing, construc-tion, operation, maintenance, etc.) personnel shall, if possible, be accomplished at the level such disputes occur. If tnis is not possible the difference of opinion shall be escalated through supervisory /managemelt II-13 Rev. 8 1869 279
levels until resolution is achieved. The Manager-Quality Assurance shall be the arbi-trator on differences of opinion involving conformance of items, components and systems to specified requirements and interpretation of the Quality Assurance Program. The Sr. VP Met-Ed/VP GPUSC and/or Vice President - Material Management shall be the final arbitrator. II-14 Rev. 8 1869 280
3.0 Control of Documents and Records 3.1 Instructions, Procedures, and Drawinas, 3 3.1.1 Policy The TMI Quality Assurance Program requires that activities important to safety be pre-scribed by documented procedures, instruc-tions, and/or drawings and that these quali-ty-affecting activities be accomplished through the implementation of these documents. 3.1.2 Requirements TMI procedures, instructions, and/or drawings which prescribe the performance of activities important to safety shall:
a. Include quantitative (such as dimen-sions, tolerances, and operating T limits) and qualitative (such as work-manship samples) acceptance criteria sufficient for determining that impor-tant activities have been satisfac-tority accomplished.

b. Require approval of appropriate person-nel prior to the initiation of the quality-affecting activity.

c. Describe the sequence of action to De accomplished,

d. Define the responalbilities and autho-rities of personnel performing the activity, s

e. Describe interfaces with other company elements or other organizations.

f. Require indoctrination of user per-sonnel prior to implementation,

g. Be distributed in a controlled manner to preclude the use of oosolete docu-ments.

h. Be distributed with su f fic ient con-trolled copies to assure availability to responsible personnel.
1869 28i s Rev. 8
3.1.3 Responsibilities 3.1.3.1 Department Managers The Manager of each department performing activities important to safety is responsible for the preparation, approval and implementa-tion of procedures, instructions and/or drawings. He is responsible to assure that provisions are made for interface controls for internal and external lines of communications among participating organizations and tech-nical disciplines. Additionally, he is re-sponsible to insure that the procedures reference the documents used in their prepara-tion and the extent to which the procedures meet the requirements of the references. 3.1.3.2 Quality Assurance Department The QAD shall review and approve those admini-strative policies, procedures, instructions and/or drawings which delineate the methods of complying with the requirements of this manual. Contractor Quality Assurance program documents specified in the applicable procurement docu- . ments shall be reviewed and accepted by QAD. Compliance with detailed procedures and in-structions snall be audited at specified fre-quencies. Vendor Quality Assurance Plans / Manuals, special process procedures, and inspection and test procedures shall be reviewed and approved by QAD prior to releasing the vendor to imple-ment such documents. Contractor Quality As-surance Plans / Manuals, work plans, selected drawings, instructions and procedures shall be reviewed and approved by QAD prior to releas-ing the contractor to start work. Adequacy shall be verifled by audit and inspection programs. 3.1.3.3 TMI Unit Managernent The Management of each TMI Unit is responsiole for assuring that instructions, drawings, and procedures associated with the administr;tive controls, operation, fuel handling, inservice inspection, calibration, maintenance, modifi-cation, repair and operational testing of III-2 Rev. 8 1869 282
structures, systems, and components important to safety are prepared, reviewed, approved and _ in accordance with approved written procedures which conform to the requireme~nts of the TMI Quality Assurance Program. All activities important to safety accomplished by the plant staff shall be performed in accordance with approved procedures, instructions, or drawings. 3.1.3.4 Delegated Authorities Those activities important to safety which are performed by contractors, agents, contractors, or vendors shall be delineated by documented, approved, ar.d controlled procedures, instruc-tions or drawings. 3.2 Document Control 3.2.1 Policy Measures shal) De established and documented to control the issuance of documents, such as program documents, design documents, instruc-tions, procedures, and drawings, including changes thereto, which prescribe activities important to safety. These measures shall assure that Jocuments, including changes, are reviewed for adequacy and approved for release by authorizec personnel and are distributed to, and used at, the location where the pres-cribed activity is performed. 3.2.2 Requirements Written document control procedures shall be established to provide for control of the following documents as a minimum:
a. As-built Drawings

b. Quality Assurance Plans / Manuals, and Instructions

c. Operating Procedures & Instructions

d. Maintenance Procedures & Instructions

e. Design Documents (e.g., calculations, drawings, specifications, analyses) including documents related to computer codes.
III-3 Rev. 8 1869 283
f. Manufacturing, Construction and Installation Drawings

g. Manufacturing, Construc' tion Modifica-tion, Installation, Test, and Inspec-tion Procedures and Instructions

h. Procurement Occuments

1. FSAR and Related Design Criteria Documents

j. Nonconformance Reports

k. Design Change Documants

1. Test Specifications

m. Operating and Special Orders

n. Equipment & Material Control Procedures

o. Refueling Procedures

p. QCL

q. Topical Reports All procedures established for document con-trol shall meet the following requirements:

a. Review, approval and issuance criteria for documents and their revisions shall be specified to assure adequate techni-cal and quality requirements are met prior to issue.

b. The individuals or elements responsible for reviewing, approving and issuing documents specified.
and their revision snall be
c. Changes must be documented, approved and included in the appropriate revi-sion document prior to being imple-mented.

d. Revisions shall be reviewed and ap-proved by the same organizations tnat performed the original review and ap-proval or by other qualified, respon-sible, and designated organizations.
III-4 Rev. 8 1869 284
e. Document distribution must be suffi-cient to assure that the documencs are readily available at convenient loca-tions to plant personne'l prior to com-mencement of work.

f. Appropriate document transmittal and maintenance measures shall be incor-porated in document control systems to prevent inadvertent use of voided, superseded or obsolete documents.
Holders of controlled documents are responsible for maintaining their assigned copies in a current status. Documents distributed for information only will not be considered a con-trolled copy, and, as such, will not ce used in performing an activity impor-tant to safety since they will not De maintain >d current. Exceptions to tnis requirement must be approved, in writing, by QAD.
g. A master list or equivalent will be established and maintained to identify the current revision numoer of instruc-tions, procedures, specifications, drawings, and procurement documents.
This list will be distributed to pre-determined responsible personnel to preclude the use of superceded docu-ments.
h. Status indication measures shall be established for all controlled docu-ments. Status lists or logs shall be maintained and be made available to project personnel to identify the cur-rent revision levels of controlled documents.

1. Maintenance, modification and inspec-tion procedures shall De reviewed by the responsible Quality Assurance or-ganization to determine:

1. The need for inspection, the identification of inspection personnel and the documentation of inspection results.
III-5 Rev. 8 1869 285
2. That necessary inspection re-quirements, methods, and accep-tance criteria have been iden-tified.
3.2.3 Responsibility 3.2.3.1 Manager - Management Services Responsible to approve the TMI Generation Group procedures for document control. 3.2.3.2 Manager - Administration and Services Responsible for implementation of the document control system for all instructions, proce-dures, drawings and other controlled documents prepared for TMI in administration, operation, testing, maintenance, and modification of structures, systems and components important to safetv. 3.2.3.3 Manager - Quality Assurance Responsible for the review and approval of document control procedures for quality as-surance requirements and document control measures; to evaluate the document control system effectiveness through review and audit. 3.2.3.4 Department Manaaers Responsible to ensure that documents are available when required; to properly review and approve documents such as procedures, instructions, specifications, drawings, etc. to ensure that changes to documents are re-viewed and approved by the same organization that performed the original review and ap-proval of the document; to ensure that ap-proved changes are promptly transmitted for incorporation into documents; to ensure that obsolete or superseded documents are elimi-nated from the system. 3.2.3.5 Delegated Authorities Vendor, contractor, and agent QA programs shall be reviewed to assure compliance with the requirements of this section. III-6 Rev. 8 i869 286
3.3 Quality Assurance Records 3.3.1 Policy Quality Assurance records for items and ac-tivities covered under the scope of the TMI QA Program shall be identified, reviewed, re-tained, and retrievable. These requirements are imposed on all organizations performing activities important to safety. Quality As-surance record systems shall be described and controlled by approved written procedures and instructions, adequately implemented, and verified by QAD through inspections and audits. 3.3.2 Requirements The procedures established for the generation, collection, storage, maintenance, and retriev-al of the TMI Quality Assurance records shall meet the following minimum requirements:
a. The applicable design specification, procurement documents, test procedures, operational procedures and other docu-ments shall specify the records to De generated, supplied and maint ined by or for the owner. These records shall include results of reviews, inspec-tions, tests, audits, and material analysis; monitoring of work perfor-mance; qualification of personnel, procedures, and equipment; and other documentation such as calculations design verifications, drawings, speci-fications, procurement documents, cali-bration procedures and reports; noncon-formance reports; and corrective action reports,

b. Sufficient records and documentation shall be maintained to provide evidence of the quality of items or activities important to safety. Inspection and test records shall contain the fol-lowing where applicable:

1. A description of the type of observation.
III-7 Rev. 8
2. The date and results of the inspection or test.

3. Information relatsd to conditions adverse to quality.

4. Inspector or data recorder iden-tification.

5. Evidence as to the acceptability or the results.

6. Action taken to resolve any dis-crepancies noted.

c. Documented and approved measures shall be established for complying with the applicable requirements of codes, standards, and procurement documents regarding record transmittal, reten-tion, and maintenance subsequent to completion of work.

d. Record storage facilities shall be .
established and utilized to prevent destruction of quality records by fire, flooding, theft and deterioration by environmental conditions such as tem-perature or humidity in compliance with the applicable standaros, codes and regulatory guides endorsed in Section 2 of the TMI Quality Assurance Plan. 3.3.3 Responsibilities 3.3.3.1 Manager - Quality Assurance
a. Responsible for reviewing and approving major participating organizations pro-cedures for the maintenance of Quality Assurance records; establishing a pro-gram for the identification, storage, retrieval, and maintenance of Quality Assurance records generated by QAD, until they are turned over for storage, and performing planned and periodic audits to verify adequacy and implemen-tation of Quality Assurance records requirements by both internal TMI or-ganizations and external suppliers.
III-8 Rev. 8 1869 288
3.3.3.2 Manager - Administration and Services Responsible for the collection, main-a. tenance, and storage of records at the plant site in accordance with approved written procedures which conform to the requirements and policy of this section.
b. Responsible for providing procedures which ensure the maintenance of records sufficient to furnish objective evi-dence that activities affecting quality are in compliance with the applicable standards, codes and regulatory guides endorsed in Section 2 of the TMI Quality Assurance Plan.
3.3.3.3 Manager - Management Services
a. Responsible for the collection, main-tenance, and storage of records at the home office in accordance with approved written procedures which conform to the requirements and policy of this section.

b. Responsible for providing procedures which ensure the maintenance of records sufficient to furnish oojective evi-dence that activities affecting quality are in compliance with the applicable standards, codes and regulatory guides endorsed in Section 2 of the TMI Quality Assurance Plan.
3.3.3.4 Delegateo Authorities Records generated by site contractors shall be controlled according to contractor procedures until such time as they are turned over to the QAD for review, acceptance, and transmittal to the permanent records file. Purchased equip-ment records shall be retaineo by the vendor until the equipment is released for shipment. When required by the procurement documents, contractors and vendors shall establish pro-cedures to control Quality Assurance records. Implementation of these procedures shall be assured by performance of source surveillance by QAD and through audits performed by QAD. III-9 jggg }89 Rev. 8
Records to De suomitted with the shipment or retained by the vendor will De specifically identified in procurement cocuments. These records will be reviewed as necessary by QA0 to provide the required degree of confidence in the adequacy of compliance of the vendor with the requirements of this section. 111-10 Rev. 8
4.0 Design Control 4.1 Policy Measures shall be established and documented to assure that the applicable specified design requirements, such as design bases, regulatory requirements, codes and standards are correct-ly translated into specifications, drawinne; procedures or instructions. These measures shall include provisions to assure thet appro-priate quality standards are specified and included or referenced in design documents for design of systems and structures; external design of systems and structures; anc assess-ment of damage. 4.2 Requirements 4.2.1 Design control measures require that:
a. The organizational structure be de-fined, and authority and responsibility of personnel involved in preparing, reviewing, approving and verifying design documents be delineated.

b. The FSAR design bases, FSAR safety analysis, design regulations, codes and standards and Plant Technical Specifi-cations be adhered to in design work, except where the changes will be the subject of an operating license amend-ment application.

c. The materials, parts and processes selected oy design are reviewed to assure that they are suitable for the intended application, including compat-icility of materials, accessioility for inservice inspection, maintenance and repair, associated computer programs, and quality standards. The review will also evaluate suitability with regard to human factors wnich may effect safe operation.

d. Internal and external design interface controls, procedures, and lines of communication among participating design organizations and across techni-cal disciplines are established and 1869 291
described for the review, approval, release, distribution, and revision of documents involving design interfaces.
e. Errors and deficiencies in approved design documents, includin i design methods (such as computer codes) that could adversely affect items and activ-ities important to safety shall be documented, and action shall be taken to assure that these errors or defi-ciencies are corrected.

f. Deviations in specified quality stand-ards shall be identified and procedu as shall be established to assure their control.

g. Review of standard "off the shelf" commercial materials, parts, and equip-ment for suitability of application with structures, systems, and compo-nents important to safety shall be conducted prior to selection.

h. Design verification methods (design review, alternate calculations or qualification testing) shall De estab-lished. Guidelines shall be established for determining the appropriate methods.

1. Design verification procedures.shall be established which assere the following:
- The verifier is qualified and is not directly responsible for the design. - Verification shall oe ucmplete prior to relying upon the compo-nent, system, or structure to perform its function during plant operations.
Procedural control is established for design documents that reflect the commitments of the SAR. Design documents suoject to pro-cedural control include, but are not limited to, specifications, calculations, computer programs, IV-2 Rev. 8 1869 292
system descriptions, and draw-ings, including flow diagrams, piping and instrument systems for major facilities, site arrange-ments, and equipment locations.
- The responsibilities of the veri-fler, the areas and features to ,
be verified, the pertinent con-siderations to be verified, and the extent of documentation shall be identified in procedures. J. When verifications may be accomplished by test: Prototype, component or feature testing shall be performed as early as possible prior to in-sta11ation of plant equipment, or prior to the point when the in-stallation would become irrevers-ible. Verification by test shall be performed under conditions that , simulate the most adverse design conditions as determined by analysis.
k. Procedures shall be established to assure that computer codes are verified prior to use.

1. Design and specification changes, in-cluding field changes, will be subject to design control measures commensurate with those applied to the original design. Design changes shall be re-viewed and approved by the Organization responsiole for the original design or by another organization with comparaDie expertise designated to review and approve changes. -

m. Measures shall be provided to assure that responsiole plant personnel are made aware of design changes and/or modifications, which may af fect the performance of their auties.
IV-3 Rev. 8 1869 293
4.3 Responsibilities 4.3.1 Plant Engineering The Plant Engineering Department is responsi-ble for providing technical support to opera-tions and maintenance personnel. This is accomplished by providing on-shift engineers to the operations staff for direct technical coverage of the plant systems performance. In addition, Plant Engineering is responsible for in-plant fuel management and accountability, control rod programs, core calculations, pro-viding technical assistance to unit management and the preparation and maintenance of proce-dures related to the activities of the depart-ment. Plant Engineering is also resoonsible for engineering activities related to routine maintenance and minor plant modifications. 4.3.2 Project Engineering Manager The Project Engineering Manager is responsible for coordination, staffing and directing of engineering tasks which are outside of the normal scope of activities of Plant Engineer-ing. To fulfill these responsibilities he will:
a. Maintain a listing of all identified design tasks and the person (s) or or-ganization assigned. For each outside design task, a Cognizant Engineer will be identified.
D. Maintain schedule and status informa-tion for each task.
c. Coordinate the efforts of the System Engineering Department and the Engi-neering and Design Department. Person-nel from these departments will oe utilized to perform the assigned tasks.

d. Control and coordinate the activities of A/E's providing direct engineering service.
IV-4 Rev. 8 1869 294
e. Coordinate with the engineering manage-ment personnel o f contractors with design responsibility. ,

f. Review and approve baseline design
documents such as design criteria, flow diagrams, system descriptions, arrange-ment drawings, one-line diagrams and logic diagrams, as appropriate.
g. Review and approve final design docu-ments including specifications and drawings (as required).
Note: This desion review does not re-place or eliminate the need for design verification by the organ-ization who performed the design. 4.3.3 Systems Engineering Department The Systems Engineering Department is respon-sible for providing conceptual and analytic engineering service to other engineering groups as required. They are directly respon-sible for technical administration of nuclear fuel-related engineering activities. The Systems Engineerinq Department is respon-sible for implementatian of the design control program for their own activities which are important to safety. 4.3.4 Engineering and Design Department The Engineering and Design Department provides detailed mechanical and electrical engineering as well as design service to support TMI engi-neering activities. They are responsible for classification of items and activities impor-tant to safety and for preparation and main-tenance of the Quality Classification List (QCL). The department is also responsible for implementation of the design control program for their own activities which are important to safety. 4.3.5 Other Design Organizations All design organizations performing design activities for TMI shall have quality programs IV-5 Rev. 8 1869 295
which include design control provisions equi-valent to those provided in the TM1 Quality Assurance Program. 4.3.6 Quality Assurance Department The Manager - Quality Assurance is responsible for providing Quality Assurance review and concurrence with design and engineering docu-ments relating to items and activities impor-tant to safety to assure that appropriate quality requirements have been included. In addition, Quality Assurance will perform plan-ned and periodic audits of responsible design organizatione to verify implementation of design control measures. IV-6 Rev. 8 1869 296
5.0 Procurement and Material Control 5.1 Control of Procurement 5.1.1 Policy Procurement of material, equipment and ser-vices which are considered important to safety shall be performed in accordance with written policies, procedures and instructions. These x shall establish methods for preparation, re-view approval, and control of procurement documents and shall provide measures to comply with applicable regulatory requirements. Appropriate measures shall be established to evaluate procurement sources, monitor the activities of consultants, vendors and con-tractors, and confirm that purchased items conform to procurement document requirements. The programs of all participants shall be in accordance with the applicable requirements of the TMI Quality Assurance Program. The general and specific requirements for the quality assurance program of all vendors and contractors, including r:mir subvendors and subcontractors supplying material, equipment, or services which are considered important to , sa fety , are delineated by procurement docu-ments. The procurement documents impose qual-ity program requirements that are commensurate with the degree of complexity, the uniqueness, and the importance to safety of the items and services being performed. Quality Assurance measures shall apply to the procurement parts, of materials including spare replacement parts, of f-the-shelf items and consumables. Procurement of spare or replacement parts for structures, systems, and components shall be subject Assurance program controls andtotocurrent Quality codes, stan-dards, and technical requirements equal to, or cetter than, original technical requirements, or in accordance with an approved engineering document. - 5.1.2 Requirements 1869 297 "*"'*
5.1.2.1 Procurement Documents s The sequence of actions for the preparation, review, approval and control o'f procurement documents shall be delineated in detailed procedures. These procedures shall delineate requirements to asssure that procurement docu-ments:
a. Specify applicable quality assurance requirements.

b. Require applicable quality program requirements to be passed on to sub-vendors and subcontractors.

c. Specify or refere.nce design bases tech-nical requirements, including appli-cable regultory requirements, material, and component identification require-ments, drawings, specifications, codes and standards, test and inspection requirements, and special process instructions,

d. Identify the documentation to be pre-pared, maintained, and suomitted for review, approval and record information as applicable.

e. Include an identification of those systems and activities important to safety.

f. Identify those records which vendors or '
contractors shall retain, maintain, and control; and those which vendors or contractors shall deliver prior to use or installation of the item.
g. Include right of access to vendors or contractors and their subtier vendor and Contractor facilities and records for source inspection and/or audit.

h. For spare or replacement parts, contain requirements at least equivalent to those used for the original procure-ment. The original procurement docu-ment may be used as the technical re-quirements for purchase of spare or replacement parts.
V-2 Rev. 8 1869 298
1. Include the provision that suppliers shall refrain from implementing pro-cedures which require owner approval prior to ootaining such' approval.
3 Measures shall be established for the review, approval, and release of procurement documents and subsequent revisions. The reviews shall assure the inclusion of the applicable tech-nical, quality, and administrative require-ments in procurement documents prior to their use. Requisitions for professional service agreements 'for services covered by the scope of this Quality Assurance Program shall be reviewed by the QA0 to assure inclusion of quality requirements. QAD personnel shall review and concur with the adequacy of quality requirements to determine that they are correctly stated, inspectable and controllable; that there are adequate acceptace criteria; and that procurement docu- ' ments have been processed in accordance with established requirements. Review of procurement documents shall be docu-mented to provide objective evidence of their m approval prior to their release. 5.1.A.2 Qualification and Selection of Suppliers The TMI Quality Assurance Program requires documented evaluations of prospective sup-pliers which demonstrate qualifications based upon one or more of the following criteria:
a. Review of performance histories which provide records of suppliers previous capability to provide similar products or services. '

b. Review of the supplier's capability to comply with the criteria of 10 CFR 50, Appendix B, applicable to the items or services to be supplied.

c. A pre-award survey of supplier's facil-ities and Qualit! Assurance program to determine his capability to supply the items or services that meet the design V-3 Rev. 8 1869 299
and quality requirements of the speci-fication. Procedures shall be established to accomplish the evaluation and selection of suppliers of equipment, material or services. Contracts or purchase orders for material, equipment or services covered by the scope of the Quality Assurance Program shall be awarded only to vendors or contractors who have been qualified by the QAD as having a Quality Assurance pro-gram commensurate with the equipment or ser-vices to be provided. When a supplier quality program is required, it shall be reviewed and approved prior to initiation of the activity affected by their program. For certain ser-vices, the supplier may be required, by pro-curement documents, to work under the direct control of the TMI Quality Assurance Program. In these instances, the supplier will not be required to have a separate quality assurance program, but will be required to work within the applicable requirements of this Quality Assurance Program and will require the ap-proval of the Quality Assurance Department. 5.1.2.3 Manufacturing Assurance Measures shall be established to provide con-trol of manufacturing activities of vendors. These methods shall be described in detailed written procedures. The extent to which these specific controls will be applied to vendors will be described in individual vendor inspec-tion plans. A vendor inspection plans will be prepared for each major contract within the scope of the TMI Quality Assurance Program. The attributes of the manufacturing assurance program shall include:
a. Provisions for the review, approval and status tracking of the vendor's draw-ings, Quality Assurance manual and selected manufacturing and quality procedures prior to fabrication. Ven-dors may not implement procedures until written notice of approval is received, if applicable.

b. Establishet vendor inspection plans that delineate, as required the hold 1869 300
and/or witness points in the manufac-turing process for specified review, inspection, verification and test.
c. Methods for resolution of nonconfor-mance where the vendor's suggested disposition is "Use-as-is" or " Repair".
Such nonr.onformances require approval by the responsible engineer.
d. Planned and systematic audit and sur-veillance of vendor quality activities.
Scope of coverage and frequency shall be determined by the criticality of the furnished items and the evaluated re-sults of vendor qualifications, in-cluding p're-award surveys and quality procedure reviews. Revisions to sur-veillance plans shall be made as war-ranted by vendor performance.
e. Control of vendor document package including review for completeness and acceptability. Inadequate records shall be sufficient cause to reject the items furnished due to their indeter-minate quality status.

f. Assessments of vendor control of quality shall be made at a frequency and depth commensurate with the import-ance, complexity and quantity of the items furnished. These assessments shall utilize the qualitative and quan-titative information provided by vendor noncompliance documents; surveillance, inspection and audit reports; and re-ceiving inspection and test records,

g. Receiving inspection procedures assure that:

1. The material, component, or equipment is clearly identified and that the identification and quantity correspond to the infor-mation on the shipping documents and quality records.

2. The item's handling and shipping, requirements have been met by the V-5 Rev. 8 1869 301
vendor and maintained by the carrier.
3. The item's quality record package or compliance certificate is complete, and adequate.

4. Items delivered, which are not in compliance with requirements are documented in accordance with the ranconformance procedure, tagged, seyregated (if possible), and prevented from being inadvertent-ly issued for installation or use.

5. Items accepted and released are identified as to their inspection status prior to forwarding them to a controlled storage area or releasing them for installation or further work.
5.1.3 Responsibilities 5.1.3.1 Materials Management Materials Management is responsible for com-plying with the requirements of this Plan and for the administration and operation of pro-curement and warehousing associated with the operation of the TMI Nuclear Station. In this regard, they are responsible for assuring that the technical and quality requirements, as established oy the Generation Group, are in-corporated into procurement document; without revision. Furthermore, Materials Management is responsible for assuring that the contrac-tural, legal and commercial requirements are incorporated into the procurement documents in a manner which will not alter the technical or quality requirements. The Manager - Field Warehousing, TMI is responsible for the operation and maintenance of the company warehouses and storerooms at the TMI Nuclear Station.
.l.3.2 Manager - Quality Assurance The Manager - Quality Assurance is responsible for assuring that QAD procedures for the con-trol of purchased equipment, material, and services are established, approved, implement-ed and effective. He is also responsible for V-6 Rev. 8 1869 302
the approval of all TMI procedures necessary for the control of purchased equipment, mate-rial, and services within the scope of the TMI Quality Assurance Program. He'is responsible for approval of suppliers' Quality Assurance Program to the extent required in the procure-ment documents. He is also responsible for review on acceptance of supplier document record packages. He is responsible for estab-lishing and implementing an adequate program of source inspection, surveillance and receipt inspection to assure supplier compliance with contract requirements. 5.1.3.3 Responsible Engineer A responsible engineer is that engineer as-signed responsibility for the design and/or procurement of each structure, system, or item. He shall review, approve and control procedures, drawings and other quality-related documents submitted by the supplier of the specified equipment. He shall maintain a status reference of all documents requiring approval and distribute such information as required. 5.2 Indentification and Control of Materials, Parts and Components, 5.2.1 Policy Measures shall be established to provide for the identification and control of materials, parts and components important to safety. These measures shall assure that incorrect or nonconforming items are identified and con-trolled in order to prevent their inadvertant installation or use at TMI. Where required by design documents, the system established shall provide traceability of components from the receipt of material through fabrication and testing. Verification shall include review of objective evidence of inspections ano tests which demonstrate that product identification and control is m81ntained at various stages of manufacture, installation, or erection. Iden-tification requirements shall be specified in the applicable design and procurement documents. V-7 Rev. 8 1869 303
5.2.2 Requirements
a. Identification requirements shall be included in specifications and drawings.

b. Material, parts, and components, in-cluding partially fabricated subassem-blies or subdivided materials shall be identified to preclude the use of in-correct or defective items.

c. Materials ano parts important to safety shall be identified so that they can be traced to the appropriate documenta-tion, including, but not limited to:

1. Specifications

2. Drawings

3. Procurement Documents

4. Physical and Che nical Test Reports S. Nonconformance Reports

6. Inspection Reports and Checklists

7. Storage Maintenance Instructions

8. NDE Reports

9. Vendor Certificates o f Compliance d.
The location and method of id3ntifica-tion shall De specified so as not to af fect the form, fit, function or quality of the item being identified,
e. Correct identification of materials, parts and components shall be verified prior to release for faorication, ship-ping, installation, and testing. .
f. Where physical identification is either impractical or insufficient, physical separation, procedural control, or other approved means may be employed.
g. A receipt inspection at tne site ware-house verifies that identification for V-8 Rev. 8 1869 304
received items is complete and accom-panied by appropriate documentation. 5.2.3 Resoonsibility 5.2.3.1 Responsible Engineer
a. Responsible for ensuring that procure-ment documents contain appropriate requirements for the identification and control of materials, parts, or compo-nents.
5.2.3.2. Manager - Quality Assurance
a. Responsible for Quality Assurance review and concurrence of procedures for maintaining identification in accordance with the requirements of this section.

b. Responsible for verification of identi-fication during receipt inspection.

c. Responsible for monitoring and conduct-ing inspections, surveillances and audits to verify conformance to the requirements of this section.
5.2.3.3 Manager - Site Warehousing, TMI
a. Responsible for maintaining identifica-tion and control of materials, parts or components received and stored at TMI in accordance with written procedures.
O
~'
1869 , 305 "*"'*
6.0 Control of of Station Activtties 6.1 Policy Station activities considered important to safety shall be conducted in accordance with the requirements of this Plan. These activi-ties include design changes, procurement, fabrication, handling, snipping, storage, cleaning, erecting, installation, inspection, tes. ting, operation, maintainance, repair, refueling and modification. 6.2 Requirements The Quality Assurance requirements for station activities are contained in this Plan and include compliance with applicable USNRC Regu-latory Guides and ANSI Standards indicated in Appendix C. These requirements shall be im-plemented in appropriate TMI procedures governing station activities. The require-ments of the Plan apply to all individuals or organizations performing functions which af-fect the quality of structures, systems, com-ponents, or activities important to safety. 6.2.1 Details The following subsections discuss typical activities which are representative of the broad scope of administrative controls and quality assurance requirements that are appli-cable to station activities. The organiza-tional structures and functional responsibi-lities governing station activities shall be structured so that attainment of Quality As-surance Plan objectives is accomplished by those who have been assigned or delegated responsibility for performing the work and verification of conformance to estaolished requirements is accomplished by qualified personnel who do not have direct responsi-bility for performing or directly supervising the work. Quality Assurance Department acti-vities sucn as inspection, monitoring, sur-veillance, reviews and audits are performed to independently verify conformance to this plan, applicable station administration controls, and applicable regulatory and licensing com-mitments. These independent verifications are applied to station activities on a graded Rev. 8 1869 306
approach and to the extent necessary to pro-vide adequate confidence that structures, systems, components, and personnel perform satisfactorily to maintain the' safety of the station. Station work functions such as rou-tine and abnormal operations, maintenance, repair or rework, in-service inspections, technical specification compliance, fuel handling, radwaste handling, radiation protec-tion, chemical analysis, housekeeping and cleanliness, fire protection, security, train-ing, environmental requirements, health physics, and other activities considered im-portant to safety which are discussed in the Quality Assurance Plan are controlled to an extent consistent with their importance to safety. 6.2.1.1 Control of Inspection A program for inspection of activities af fect-ing quality shall be established and executed by, or for, the organization performing the activity to verify conformance to the docu-mented instructions, procedures, and drawings for accomplishing the activity. Design speci-fications, drawings, procedures, or instruc-tions shall include the necessary inspection requirements. These requirements include acceptance criteria and reference to codes, standards, and regulatory documents. These requirements shall be furtner translated into inspection procedures, instructions, or check-lists which shall contain, as required, the following:
a. Identification of characteristics and activities to be inspected.

b. Inspection methods.

c. Idet i fication o f organization respon-sible for performing the inspection,

d. Acceptance and rejection criteria,

e. Identification of applicaule revisions or required procedures, drawings and specifications.
VI-2 Rev. 8 1869 301
f. Documentation of inspection results including identificat',on of the in-spector.

g. Listing of necessary measuring and test equipment including their accuracy requirements.
Inspectors (including NDE personnel) shall be qualified in accordance with applicable codes, standards and TMI training programs and their qualification and, certification shall be kept current and documented. Individuals performing inspections shall be other than those who performed or directly supervised the activity being inspected and shall not report directly to the immediate supervisors who are responsible for the work activity being inspected. If the individuals performing inspections are not part of the responsiole Quality Assurance organization, the inspection procedures and personnel quali-fication criteria shall be reviewed and con-curred with by the responsible Quality Assurance organization prior to the initiation nf the inspection activity. Inspection of activities as defined in ANSI N45.2.10 may be conducted by second line supervisory personnel or by other qualified personnel not assigned first line supervisory responsibility for the conduct of work. These inspections, i.e., those performed by individuals not assigned first line supervisory responsibility, are not intended to dilute or replace the clear re-sponsibility of first line supervisors for the quality of work performed under their seper-vision. .When inspections associated with normal operations of the plant (such as rou-g
- tine maintenance, surveillance ano tests) are performed by individuals other than those who performed or directly supervised tne work, but are within the same group, the following con-trols shall be met:
a. The quality of the work can be demon-strated througn a functional test when the activity involves breaching a pres-sure retaining item.
VI-3 Rev. 8 1869 308
b. The qualification criteria for inspection personnel are reviewed and found acceptable by the Quality Assurance organization prior to initiating the inspection.
Work authorization documents relating to work considered important to safety shall be reviewed by Quality Assurance Department per-sonnel to determine the need for: a) inspeu-tion, b) identification of inspection organization, c) identification of inspection witness and hold points, d) documenting in-spection results. When hold points have been establisned, either contractually by procurement or internally by plant procedures, work may not proceed until either inspection is performed or waived oy the responsible Quality Assurance organization. Inspection of modifications, repairs, and replacements shall De by the same method and to the same criteria as the original inspec-tion or by an approved, documented, engi-neering and QA alternate. Where direct inspection is not practicable, control of s processing, equipment and personnel .ia C he based on statistically valid sampling plans. Inspection personnel shall be provijed with suitable equipment and tools, which are cali-brated as necessary, and controlled to assure that accuracy rcquirements are sati;fied and that inspections are complete. Inspection data ead results snall te evaluated by designated personnel to assure that the inspection objectives have been met and that items requiring action or follow-up are iden-tified and documented. Records shall be kept in sufficient detail to provide adequate confirmation of a 1 inspection program. 6.2.1.2 Control of Soecial Processes Measures shall be estaolished and documented to assure that special processes are accom-plished under controlled conditions in accor-dance with applicable codes, standards, appli-cations criteria, and other special require-ments including the use of qualified personnel VI-A Rev. 8 i869 309
and procedures. Special processes are those that require interim in process controls in addition to final insp.ction to assure quality including, but not liriteo to, such processes as welding, heat treating, chemical cleaning, and nondestructive examination. Procedures for special processes shall be estholished to meet the requirements or applicable codes and standards, where applicable, or to meet the requirements of special process specifications which may be produced for TMI. These proce-dures shall provide for recording evidence of acceptable accomplishment of special pro-cesses. Procedures and 17structions for the control of special processes shall be reviewed and approved by qualified personnel. Proce-dures, equipment, and personnel performing special processes shall be qualified in accor-dance with applicable codes, standards, and specifications. Organizational responsibili-ties shall be delineated for the qualification of special processes, equipment and per-sonnel. Qualification records of personnel equi.pment and procedures associated with special processes snall be established, main-tained and kept current. For special pro-cesses not covered by the existing codes or ' standards, or when item quality requirements exceed the requirements of establisned codes or standards, the necessary qualifications of personnel, procedures and equipment shall be defined. 6.2.1.3 Test Control A documented test program shall be established to assure that all testing requireo to demon-strate that the structure, system or component considered important to safety will perform satisfactorily in service. The tests shall be performed in accordance with written, ap-proved, and controlled test proceoures which incorporate or reference the requirements and accaptance standards contained in the appli-cable design documents. The extent of testing shall be based on the complexity of the modi-fication, replacement, or repair. Testing, including proof tests prior to installation and preoperational tests, necessary to demon-strate that structures, systems and components will perform satisfactorily in service, shall be accomplished in accordance with written VI-5 Rev. 8 1869 310
approved procedures. These procedures shall be based on requirements and acceptance limits contained in applicable design and procurement documents. These test proceoures or instruc-tions shall provide for the following as required:
a. A description of the test objective.

b. Instructions for performing the test, including caution or safety notes in sufficient detail to avoid operator interpretation.

c. Test prerequisites such as calibrated instrumentation, adequate test equip-ment and instrumentation including accuracy requirements, completeness of item to be tested, suitaole and con-trolled environmental conditions, and trained qualified and licensed or certified personnel.

d. Provisions for data collection and storage,

e. Acceptance and rejection criteria as specified in design anu procurement documents, f.
Methods of documenting or recording test data and results, in sufficient detail to prevent misinterpretation.
g. Provisions for assuring that test pre-requisites have reen met,

h. Mandatory nold or witness points for inspection oy TMI Quality Assurance and/or other designated personnel,

i. Provisions for control o f jumpers, lifted leads ano jurisdictional or safety tags.

j. Provisions for returning a system to normal configuration upon completion of the test.
Test results shall be documenteo, evaluated, and their acceptaDility determined Dy a re-sponsible individual or group. VI-6 Rev. 8
The test program shall cover all reqilired tests including:
1. Tests during the preoperational period to demonstrate that plant performance is in accordance with 'tne design intent.

2. Tests during the initial operational phase to demonstrate the performance of systems that could not be tested prior to operation to confirm that plant behavior conforms to design criteria.

3. Tests during the operational phase to provide assurance that failures or substandard performance do not remain undetected and that tne required rella-bility of systems important to safety is maintained.

4. Tests during activities associated with plant maintenance during the opera-tional phase and to oemonstrate satis-factory performance following plant maintenance or procedural changes.
Tests performed following plant repairs or replacements shall be conducted in accordance with the original design and testing require-ments or engineering approved, documented alternatives. Testing shall be sufficient to confirm that the changes reasonably produce expected results and that the change does not reduce safety of operations. 6.2.1.4 Control of Measuring and Test Equipment Measures shall be established to assure that tools, gauges, instrumcats, and other measuring and testing devices used in activi-ties affecting the function or quality of structures, systems, and components covered unde- the scope of the TMI Quality Assurance Program be properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within specified limits. Additional measures shall be established to ensure the range, type and accuracy of test equipment conforms to the specified testing requirements. VI-7 Rev. 8
]g( }}}
Requirements for each control program shall include inspection and verification of ac-curacy upon receipt of equipment, identifica-tion of all gauges and instruuents, cali-bration and scheduled recall for calibration and traceability to an accepted Standard. Procedures shall be established to implement the following requirements:
a. To establish the calibration technique and frequency maintenance, and control of all measuring and test eqaipment which are used in the measurement, inspection, and monitoring of compo-nents, systems, and structures covered under the scope of the TMI Quality Assurance Program (instruments, tools, gauges, fixtures, reference and trans-fer standards, and nondestructive examination equipment).

b. The identification of measuring and test equipment traceable to the cali-bration test data.

c. Installed operations measuring and test equipment requiring calibration shall be labelled, tagged or otherwise con-trolled in accordance with written, approved procedures to assure that approved calibration intervals are not exceeded. Portable measuring and test equipment may be similarly controlled; but shall, as a minimum, be clearly labelled to indicate the date on which the current calibration expires. Port-able measuring and test equipment that has exceeded the approved calibration interval shall not be used for measure-ments or tests.

d. Establish calibration frequency for measuring and test equipment based on required accuracy, purpose, degree of usage, stability characteristics, and/-
or any other condition which may affect the measurement. A calibration recall system shall be implemented to assure recalibration within the required period for each piece of measuring and test equipment covered under the scope of this program. VI-8 jggg }}} Rev. 8
e. Methods for determining the validity of previous inspections performed when the measuring and test equipment is found to be out of calibration. Inspections or tests are repeated on items deter-mined to be suspect. Such determina-tion is to be documented in suitable form. If any calibration, testing or measuring device is consistently found to be out of Calibration, it shall De repaired or replaced.

f. Calibration shall be against standards that have an accuracy of at least four times the required accuracy of the equipment being calibrated. When this is not possible, standards shall have an accuracy that assures the equipment being calibrated will be within required tolerance and that the basis of acceptance is documented and authorized by the supervisor of the cal *5 rating organization,

g. A status of all measuring and test equipment under the calibration program is to be maintained.
m
n. Utilization of reference and transfer standards traceable to nationally re-cognized standards. Where national standards do not exist, provisions shall be established to document the basis for the calibration.

1. NDE equipment, sucn as ultrasonic equipment, shall oe controlled and calibrateo in accordance with the ASME code governing its use.
6.2.1.5 Handling, Storage and Shipping Measures shall be establisned ano documented to control handling, storage. and shipping, including cleaning, packaging, and preserva-tion of items important to safety in accor-dance with established instructions, proce-dures, and drawings to prevent damage, deterioration or loss. VI-9 Rev. 8 i869 314 .
Organizations performing special handling, preservation, storage, cleaning, packaging, and shipping activities shall.do so in accor-dance with predetermined work and inspection procedures or instructions utilizing suitably trained individuals. Procedures shall be established to control the cleaning, handling, storage, packaging, and shipping of materials, components, systems in accordance with design and procurement re-quirements to preclude damage loss or deter-ioration by environmental conditions such as temperature or humidity. These procedures shall include an assessment of, but not limited to, the following:
a. Packaging and preservation procedures to provide assurance of adequate pro-tection against corrosion, contamina-tion, physical damage or any effect which would lower the quality of the items or cause them to deteriorate during shipping, handling or storage.
Special protective environments, special coverings, inert gas atmos-phere, allowable moisture content, and temperature level shall be specified as required and their existence verified and documented.
b. Cleaning procedures to provide assurance that necessary cleaning operations are carried out prior to packaging, storage or installation.
The level of cleanliness required, and verification and documentation require-ments shall be specified in the proce-dures.
~
c. Detailed handling procedures to be provided for all items that require special handling. Special handling tools and equipment shall be provided and controlled to ensure safe and ade-quate handling. These tools and equip-ment shall be maintained, inspected and tested in accordance with written pro-Cedures at established intervals to ensure their reliability and availa-bility for use.
VI-10 Rev. 8 g g 3jg
d. Storage procedures to provide for methods of storage and the control of items in storage which will minimize the possibility of damage or deteriora-tion during storage. Periodic inspec-tions of storage areas snall be per-formed and documenteo to verify com-pliance with storage procedures. Re-lease of items for installation shall also be procedurally controlled.

e. Procedures to be provided to assure that proper marking and labeling of items and containers is accomplished to provide identification and necessary instructions during packaging, shipment and storage.

f. Procedures for documenting and report-ing noncompliance and nonconformance to handling, and shipping requirements.

g. Provisions for the storage of chemi-cals, reagents, lubricants and other consumable materials which will be used in conjunction with systems which are important to safety.

h. Provisions for " Limited Life" require-ments (including " Shelf Life" and "Ser-vice Life" for applicable materials.
6.2.1.6 Inspection, Test, and Operating Status Measures shall be established and documented to ensure that the required inspections and tests are performed and that the acceptability of items with regard to inspection and tests performed is known throughout manu f ac turing , installation, and operation. Status of items covered by the scope of the TMI Quality As-surance Plan shall oe controlled in accordance with approved procedures. These procedures shall include the use of appropriate tags, markings, lists, logs, diagrams, or other suitable means, to assure that required in-spections and tests are satis factorily com-pleted to prevent inadvertent Dypassing of required inspections and tests and to prevent inadverent operation. VI-ll Rev. 8
The requirements for an acceptable inspection, %q test and operating status program for struc-b tures, systems, and components.throughout fabrication, installation and test include:
a. Design and quality documents which address the requirements for the iden-tification of inspection, test, and operating status of structures, systems and components,

b. Procedures which include controls for the application and removal of inspec-tion and welding stamps, and other status indicators such as tags, markings, labels, and stamps.

c. Bypassing or altering the sequence of required inspections, tests or other critical operations procedurally con-trolled by Engineering procedures with concurrence by the appropriate quality organization. Where necessary to pre-clude inadver.ent bypassing of required inspections and tests, the procedures shall provide for the identification of items which have passed such inspec-tions and test.

d. In cases where documentary evidence is not available to confirm that an item has passed reauired inspections and tests, that item shall be considered nonconforming until such evidence be-comes available. Affected systems shall also be considered to be in-operable and reliance snall not be placed on such systems to fulfill their intended sa fety functions.

e. Procedures to be provided to require identification of t?.e operating status of systems, components, controls, or support equipment in order to prevent inadvertent or unauthorized operation.
These procedures shall require control measures such as locking or tagging to secure and identify equipment in a 1869 317 VI-12 Rev. 8
controlled status. Incependent verifi-cation shall be required, where appro-priate, to ensure that necessary measures, such as tagging equipment, have been implemented correctly,
f. Temporary modifications snall be con-trolled by approved procedures which include a requirement for independent verification. A log shal] be main-tained of the current status cf such temporary modification.

g. Nonconforming services and inoperative or malfunctioning structures, system, components or materials shall be iden-tified, documented and controlled in accordance with the requirements of this Plan.
6.2.1.7 Fire Protection The primary objective o f a Fire Protection Program is to minimize Doth the probability
and consequences of postulated fires. Fire Protection starts with design and must be carried through all phases o f construction and operation. The re fo re , Quality Assurance Program requirements in accordance with Regu-latory Guide 1.120 and this Plan shall be established to assure the reliaoility of the THI fire protection systems. Quality measures shall be established to ensure tnat the guide-lines for design, measurement, installation, testing and administrative controls for the fire protection systems are satisfied.
6.2.1.8 Plant Security Procedures shall be developed utilizing the guidelines of ANSI N18.17-1973 to supplement features and physical barriers designed to control access to the plant and, as appro-priate, to vital areas within the plant. Information concerning specific design fea-tures and administrative provisions of the plant security programs shall De confidential and thus accorded limited districution. Quality measures snall be establisned to en-sure that the guidelines for design, measure-ment, installation, testing ano administrative VI-13 1869 318 Rev. 8
controls for the plant security systems are satisfied. s 6.2.1.9 Housekeeoinq and Cleanliness Housekeeping practices on a regularly scheduled basis shall be utilized recognizing the requirements for the control of radiation zones and the control of work activities, conditions and environments that can affect the quality of important parts of the nuclear plant. Housekeeping encompasses all activi-ties related to the control of cleanliness of facilities, materials, equipment fire preven-tion and protection including disposal of comoustible material and debris ano control of accesses to areas, protection of equipm'.at, radioactive contamination control and storage of solid radioactive waste. Housekeeping practices shall assure that or'y proper materials, equipment processes, and proceoures are utilized and that the quality of the item is not degraded as a result of nousekeeping practices or techniques. During maintenance activities, certain portions of safety-related systems may be subject to potential contamina-tion with foreign materials. To prevent such contamination, control measures, including measures for access control, snall be esta-blished. Additionally, immediately prior to closure, an inspection shall De conducted and documented to ensure cleanliness. Special housekeeping considerations shall be made for maintenance of radioactively contaminated systems for components. 6.2.1.10 Equioment Control Permission to release equipment or systems for maintenance shall be granteo by oesignated NRC SRO licensed operations personnel. Procedures shall De provided for control o f equipment, as necessary, to maintain personnel ano reactor safety, to avoid unauthorized operation of equipment, and to assure that operatit..al equipment is a ready status. Tnese procedures shall require:
a. Control measures such as locking or tagging or secure and identiry equip-ment in a controlled status.
i869 319 VI-14 Rev. 8
b. Independent verifications, where appro-priate, to ensure that necessary measures, such as tagging equipment,
' has been implemented correctly.
c. Control measures for temporary modifi-cations, such as temporary by-pass lines, electrical jumpers, lifted elec-trical leads, and temporary trip point settings. Included shall be a require-ment for independent verification. (A log shall be maintained of the current status of temporary modifications.)

d. Control of inspection and test status on individual items by the use of markings such as stamps, tags, labels, routing cards or other suitable means.

e. When equipment is ready to De returned to service, operating personnel shall place the equipment in operation and verify and document its functional acceptability.
6.2.1.11 Control of Construction, Maintenance (Preventive / Corrective) ano Mocifications
- Construction, maintenance or modifications which has the potential to affect the func-tioning of structures, systems or components important to safety shall be performed in a manner to ensure quality at least equivalent to that specified in original design bases and requirements, materials specifications and inspection requirements. A suitable level of confidence in structures, systems or compo-nents on which maintenance or modifications have been performed shall be attained by ap-propriate inspection and performance testing.
Construction, maintenance or modification of equipment shall be preplanned and performed in accordance with written procedures, documented instructions or drawings appropriate to the circumstances which conform to applicaole codes, standards, specifications, and criteria. Skills normally possessed by quali-fied maintenance personnel may not require detailed step-cy-step delineations in a written procedure but are subject to general admi' rative procedural controls that govern. or define the following areas: 1869 320 VI-15 Rev. 8
a. Methods for ootaining permission and s clearance for operation personnel to work and for logging such work,

b. Factors to be taken into account, including the necessity of maintaining occupational radiation exposure as low as is reasonably achievable (ALARA).

c. Method for identification of what procedural coverage is necessary for the maintenance, construction and modi-fication activity.

d. Considerations for system / equipment cleaniness control.

e. Method for identification of post _ main-tenance, construction or modification, testing, including system / equipment functional capability to meet opera-tional requirements in all respects.

f. Method for ensuring that maintenance, contruction or modification activities, performed either on-site or off-site, are properly reviewed.
The following type of activities are among those that may not require detailed step-by-step written procedures:
a. Gasket replacement

b. Trouble shooting electrical circuits

c. Changing chart or drive speed gears or slide wires on recorder.
Means for assuring quality of maintenance, modifications or construction activities (for example, inspections, measurements, tests, welding, heat treatment, cleaning, nondestruc-tive examination and worker qualifications in accordance with applicable coaes and standards) and measures to document the per-formance thereof shall De established. Measures shall be establisheo and documented to identify the inspection and test status of items to be used in maintenance modification, and construction activities. the Normally69 18 32 i VI-16 Rev. 8
point of control for such items should be the plant storage area. A corrective maintenance program shall De developed to maintain structures, systems and components important to safety at the quality required for them to perform their intended functions. Corrective maintenance shall be performed in a timely manner to insure that important to safety items are adequately main-tained in the original, design, functional status. A preventative maintenance program including procedures as appropriate for structures, systems, and components important to safety shall be established which prescribes the frequency and type of maintenance to be per-formed. In all cases, maintenance shall De scheduled and planned so as not to compromise the safety of the plant. Planning shall consider the possible safety consequences of concurrent or sequential maintenance, testing or operating activities. Preventive main-tenance shall be performed in a timely manner to insure that important to safety items are adequately eaintained in the original, design, functional status. 6.2.1.12 Procedural Requirements Measures shall be estaolished to control and coordinate the approval ano issuance of docu-ments, including changes, which prescribe all activities affecting quality. Those documents which are considered important to safety require a documented Quality Assurance Depart-ment review. This review is to provide an independent verification ' hat the procedures have been prepared, reviewed and approved in accordance with established policy and program controls; they contain the necessary policy and program requirements including the inspec-tion and verification requirements where ap-plicable; and they contain clear descriptions related to tne extent of documenting results of completed actions when required. These documents include operating and special orders, operating procedures, test procedures, equipment and material control procedures, maintenance or modification procedures, and refueling procedures. Each procedure shall be 1869 322 VI-17 Rev. 8
- reviewed and approved prior to initial use. Plant procedures shall be reviewed by an in-dividual knowledgeable in the area affected by the procedure no less frequently than every two (2) years to determine if changes are necessary or desirable. 6.2.1.13 Control of Surveillance Testing and Inspection A surveillance testing and inspection program shall be established to insure that important to safety structures, systems, and components will continue to operate, keeping parameters within normal bounds, or will act to put the plant in a safe condition if they exceed normal bounds. Provisions shall be made for performing required surveillance testing and inspections, including inservice inspections. Such provi-sions shall include the estaolishment of a master surveillance schedule reflecting the status of all planned inplant surveillance tests and inspections. Frequently of sur-veillance tests and inspections may be related to the results of reliability analyses, the frequency and type of service, or age of the item or system, as appropriate. Additional control procedures shall be instituted, as necessary, to assure timely conduct of surveillance tests and inspections and appropriate documentation, reporting, and evaluation of the results. Following tne completion of testing, procedures shall be established to assure tne return of systems to an operable status. These procedures shall include provisions for the documentation of authority, conduct, responsibility, and veri-fication involved in returning the system to an operable status. Such provisions shall include the use of procedures, checklists, and independent verification as appropriate, considering the degree that system status was altered during the perfor9ance of the test. 6.2.1.14 Radiation Control Procedures shall be provided for the implemen-tation of a radiation control program. The radiation control program involves the acqui-sition of data and provision of equipment to 1849 323 VI-18 Rev. 8
perform necessary radiation surveys, measure-ments and evaluations for the assessment and control of radiation hazards associated with TMI. Procedures shall be developed and implemented for quality assurance review of records and programs to insure the adequacy of measures taken to control radiation exposure of employees and others. Additionally, quality measures for radwaste management shall be implemented in accordance with 10 CFR 71, Appendix E. 6.3 Responsibilities 6.3.1 Senior Vice President - Met-Ed The Senior Vice President - Met-Ed has overall responsibility for the management, super-vision, and control of all station activities. 6.3.2 TMI Generation Group Station Management The TMI Generation Group Station management is responsible for the implementation and com-pliance of the Quality Assurance Plan and directly responsible to insure their respective activities and responsibilities are conducted in accordance with applicable adminsitrative controls, regulatory and licensing requirements. 6.3.3 Delegated Authorities Contractors or other agents outside the TMI Generation Group who are assigned or delegated responsibilities ano/or activities governed by this Plan shall comply with the applicable requirements of the Plan. 1869 324 VI-19 Rev. 8
7.0 Control of Radioactive Waste 7.1 Policy Measures shall be establisned ano documented to assure that the applicable requirements of the Code of the Federal Regula' ions, Title 10, Part 71 and Title 49, Parts 100 through 199 applicable to the packaging and transporting of radioactive wastes are satisfied. Appendix E to 10 CFR 71 identifies the quality as-surance criteria applicable to the control of radioactive waste. The applicabl.S portions of this Plan that relate to tne criteria in Appendix E to 10 CFR 71 describe to i large extent the admini-strative controls and quality requirements to be ap lied in the control of Radioactive Waste. Typically, Sections 6.2.1.1 thru 6.2.1.6 and 6.2.1.9 apply to Control of Radio-active Waste. 7.2 Reouirements 7.2.1 Procedures snall be developed and implemented to cover the following: s a. Processing of radioactive wastes in-c.'.cding the collection, handling and preparation for shipment of radioactive 11gulds and solids. These procedures shall be consistent with the ALARA program and shall clearly identify the administrative controls and organiza-tional responsibilities.
b. Training and qualification of personnel operating radioactive waste processing equipment, health physics monitoring, packaging and shipping and other opera-tions deemed appropriate by management,

c. The activities associated with the packaging of radioactive wastes to include the proper selection of the receptacles to be used for containing the waste materials, the selection of the shipping containers (structures used to contain and support the receptacle and its contents) Health Physics inspections of the packaging prior to release, proper markings on
) 8 h.9Re}2)
the outside of the package and the preparation of shipping papers and certificates.
d. Movement of radioactive materials witn-in and outside the protected area to assure personnel protection at all times,

e. The shipment of radioactive material from the Station to be in accordance with the regulations of the U.S.
Department of Transportation for the transportation of hazardous materials (49 CFR) and of the NRC (10 CFR 71).
f. The packaging used for transporting of radioactive wastes, whether purchased from an outside supplier or designed by GPUSC, shall meet the applicable re-quirements of 10 CFR 71 and 49 CFR.
7.2.2 The carriers to be used for transporting of radioactive wastes shall be selected on the basis of their experience, knowledge of DOT regulations, control and maintenance of their equipment and the selection and control of their drivers. The carrier is required to have or shall be supplied documented proce-dures covering acceptance of materials from a shipper, certification requirements, placard-ing, stowage control, reporting of incidents and security. 7.2.3 Radwaste operations shall be controlled to minimize personnal exposures or environmental contamination consistent with ALARA. 7.2.4 Operations procedures shall be reviewed by QA0 to establish any necessary witness or hold points or activities to be monitored. 7.3 Responsibilities 7.3.1 The Manager - TMI Unit 1 and Manager - Radio-logical Controls shall develop and implement procedures for processing activities and move-ment of radioactive materials. 7.3.2 The Operations Department shall be respons1 Die f~ the processing and packaging of liquid wm .tes and for the packaging of o(1 wa g VII-2 Rev. 8
in preparation for shipment. Additionally, the Operations Department is responsible for the collection and identification of radio-active solids, such as rags, papers, boots, gloves, etc., and having them moved to the Radwaste facility for packaging. 7.3.3 The Radiological Controls Department is responsible for monitoring all activities associated with the processing and handling of radioactive wastes and for providing advice on radiological matters relating to processing, packaging and shipping. 7.3.4 The Operations Department is responsible for the selection of the proper packaging for the specific contents tu be. shipped, taking into consideration the radiation levels, contamina-tion limits and shipping requirements. Health Physics inspects the packaging for radiation level and, if acceptable, the Operations Department marks the outside of the package with the appropriate markings, completes the shipping papers and certificates, attaches the security seal and advises the carrier that the shipment is ready. ' 7.3.5 Plant Engineering is responsible for reviewing and accepting the designs of packaging pur-chased from an outside supplier. If packaging is to be designed by GPUSC, the design, fabri-cation and licensing of the packaging shall be the responsiollity of the Director - Technical Functions. 7.3.6 Each manager for this functional area related to the control of radioactive wastes, shall establish the requireme- - for personnel qualification and instiu a training and in-s doctrination to satisfy t.ese requirements. Training requirements shall be consistent witn the importance and complexity of the activity performed. 7.3.7 Quality Assurance Modifications /00erations Manager is responsible for review and con-currence with procedures describing control of radioactive waste. He is also responsiDle to monitor and/or inspect radioactive waste pro-cessing operations to the extent to verify VII-3 ]P4Q o T
Rev2.7'8
they are preformed in accordance with esta-blished procedures, applicable administrative controls and regulatory requirements. The Operations Department shall review and accept carriers' documented procedures as specif1.ed by procurement documents covering acceptance of radioactive waste materials for shipment. 1869 328 VII-A Rev. 8
8.0 Control of Corrective Actions and Nonconformances 8.1 ~ Policy Nonconforming materials, parts, components, services or activities witnin the scope of the QA Plan shall be controlled to prevent their inadvertent utilization. As a result, mea-sures shall be established which ensure that conditions adverse to quality, such as fail-ures, malfunctions, deficiencies, deviations, defective material and equipment, and noncon-formances be promptly identified and correct-ed. The cause of the conoition adverse to quality shall ce determined and aopropriate action taken to preclude repetition. The identification, cause, and actions taken to correct conditions adverse to quality shall be documented and reported to the appropriate levels of management. Significant conditions sitnin the intent of 10CFR 30.55(e) or 10 CFR 21 shall be reported to appropriate management levels within the affected organization for review and evalua-tion. 8.2 Requirements Procedures shall be estaolished which detail and implement the following carrective action system measures:
a. Conditions adverse to quality shall be evaluated to determine the need for corrective action.

o. Corrective action documentation shall include identification, cause, and actions taken to correct and to pre-clude the similar recurrence for condi-tions adverse to quality.

c. Follow-up activities shall oe Conducted to verify implementation of corrective actions and to close out corrective action in a timely manner.

d. Significant deficiencies, noncon-formances and de fec ts , reoortable under 10 CFR 50.55(e) or 10 CFR 21 shall be 1B^9 329 Rev. 8
reported to appropriate management levels for evaluation and possible reporting to the Nuclear Regulatory Commission.
e. Control of nonconforming materials, parts, components, services, or activ-ities. These procedures shall address and detail measures to implement the following requirements:

1. Measures for the identification, documentation, segregation *, and dispositions of nonconforming materials, parts or components.

2. Disposition of nonconformances shall be made by the organization that established the governing requirements or by other quali-fled individuals or committees authorized by the TMI Generation Group.

3. Nonconformance reports shall be used to identify materials, parts, components, and activities s which are not in compliance with the requirements o f specifica-tions, codes, drawings, and de-tailed installation or manufac-turing program requirements.
This shall ' include use o f noncon-
~
formance reports on items whose status is indeterminate due to the lack of documentation. Non-conformance recorts on items shall contains the following minimum information: (a) Identification of the non-conforming item and date of inspection. (D) Identification of the initi-ator o f tne nonconformance report. (c) Cescription of the noncon-formance. 1869 330 VIII-2 Rev. 8
(d) Disposition of the noncon-formance (repair, rework, use as is, or scrap). (e) Inspection requirements. (f) Required approval signatures of the disposition and the verification. (g) Evidence of review for re-porting per 10 CFR 50.55(e) or 10 CFR 21.
4. Reworked, repaired, and replace-ment items shall be reinspected and tested in accordance with the original inspection and test requirements or acceptable alter-natives as determined by Engi-neering and Quality Assurance.
All inspection, testing, rework, and repairs snall oe by approved procedures and the results docu-
, mented.
5. Identification of nonconforming items by appropriate means (tags, labels, etc.) and segregallun, if practical, until disposition of the nonconforming item has been determined.

6. Prior to the initiation of a preoperational test on a safety-related item all nonconformances snall be evaluated for signifi-cance or impact on further test-ing or operation.

7. Nonconformance reports shall be periodically analyzed to show quality trends. Sucn analysis will be based upon severity, numoer, frequency of noncon-formances, the causes of tne nonconformances, and the timell-ness ano adequacy of the report-ing and resolution of nonconfor-mances. Significant results shall be reported to management for review and assessment.
VIII-3 )8Ikbv.3bl
8.3 Responsibilities 8.3.1 The Manager - Quality Assurance is responsible for the review and concurrence of all proce-dures for reporting and controlling of noncon-formances for compliance with the requirements of the Operational Quality Assurance Plan. 8.3.2 Operations Management is responsible for en-suring that nonconformances are reported and corrected for plant personnel acti,ities in-volving operation, maintenance, repair re-placement, addition, modification, health physics, environmental monitoring, fuel han-dling, and inservice inspection. Plant items such as failures, malfunctions, deficiencies, deviations and defective materials, parts or components are handled in a manner consistent with their importance to sa fety and reviewed in accordance with appropriate procedures and the applicable Technical Specifications. 8.3.3 Each Manager is responsiole for the disposi-tion and corrective action of nonconformances identified as within the scope of his respon-sibilities. In the specific case of mate-rials, parts, components, or systems which s have not been installed or accepted as opera-tional at the Station, the responsible Manager and the Manager - Quality Assurance approves the resolution of nonconformances. 1869 332 VIII-4 Rev. 8
9.0 Audits 9.1 Policy A comprehensive system of planned and docu-mented audits shall be established and exe-cuted:
a. To ensure that Quality Assurance re-quirements are adequate, effective and implemented.

b. To ensure than nonconformance and Quality Assurance deficiencies ara identified and corrected.

c. To verify compliance with the TMI Quality Assurance Program.
In addition, this audit program shall provide data for a continuing evaluation of the ef fec-tiveness of the TMI Quality Assurance Program. 9.2 Reauirements A comprenensive system of audits snall be establishea for oot'1 internal and external functions which af fect structures, systems, components, operations and activities covered by the scope of the TMI Quality Assurance Program. Planned and scheduled audits shall verify compliance with the following:
a. TMI Quality Assurance Program

b. 10 CFR 50, Appendix 8

c. Regulatory Guides, ANSI, and other codes and standards ac enoorsed in the TMI Quality Assurance Program.

d. Operating procedures

e. Plant technical specifications

f. Administrative procedures 1869 333 Rev. 8

g. Otner procedures and instructions af-fecting quality

h. Procurement documents 9.2.1 Auoit Program Audits shall be performed in accordance with pre-estaolished written procedures and check-lists, and shall be conducted by trained and qualified personnel having no oirect responsi-bilities in tne areas being aucited. The audit program shall include

a. Audit schedules

b. Procedures for preparation, performance and reporting of audits

c. Analysis of aucit data anc reporting results to appropriate levels of management

d. Follow-up action to be taken based upon individual and collective audit reports

e. Qualification of auditors

f. Delineation of the autnority, r e s po n s i--
oility, and organizational independence of those responsiole for the audit program. Audits snall be regularly scheduled based upon the status and safety importance of activities being performed and shall oe initiated in a timel f manner to assure tne effectiveness during design, procurement, manu f ac turing, construction, installation, inspection, test-ing ano as required by the technical specifi-cations for TMI. In addition, audits may be scheduled and performed as required by manage-ment or the safety review groups for special evaluations. Implementation of corrective action snall be verified in a timely manner. Unschedulec audits may be concucted at any time on any aspect of this Quality Assurance Plan. Botn TMI Generation Group and organizations providing goods and/or services are subject to the audit requirements of this Programtg(,g }}f IX-2 Rev. 8
Audits will be performed by the Quality As-sorance Methods / Operations / Audit group. Each audit team shall oe led by a qualified Audit Team Leader. Audit team members shall be utilized as required and will oe classifi.ed as either auditors or technical specialists, depending on their function on the audit team. Procurement documants shall include audit access requirements in insure vendor com-pliance to the audit program. Audited organi-zations shall cooperate with the auditing organization, providing whatever assistance is nece;sary in the performance of the audit. The audited organization shall take corrective action for findings and resolve coservations in a timely manner. 9.2.2 Audit Frequency Audit frequencies shall be cased upon the status and s a f e t~y importance of activities, degree o f previous experience, consistency of overall coverage, unique testing / operating activities, and follow-up on previous audit findings. 9.2.3 Documentation Audit results shall be documented in a written report to the audited organization. The Quality Assurance organization conducting the audit is responsible for conducting follow-up actions including re-audit of de ficient areas, as required, to assure correction of the defi-ciencies. 9.2.4 Trainina Auoits shall be performed oy personnel who are trained and qualified to the requirements defined in ANSI N45.2.23. Tnese requirements provide the means to assure that audits are performed in a thorougn and professional man-ner. Documented training programs shall oe organized to provice auditors witn tne neces-sary training and knowledge of regulatory requirements, codes, standards, procedures, etc. aoolicable to the activities being audited. 1PA9 335 IX-3 Rev. 8
9.3 Responsibilities 9.3.1 Senior Vice President - Met-Ed Responsible for the performance of an indepen-dent review of the TMI Quality Assurance Pro-gram and related activities. 9.3.2 Manager - Quality Assurance
a. The Manager-Quality Assurance is re-sponsible for establishing and imple-menting the overall Quality Assurance audit program. He assures that all applicable areas are audited and that the auditing organization meets the requirements of this Plan. He evalu-ated the effectiveness of the overall audit program, analyzes tne reports and related information for quality trends and appraises the TMI Generation Group Management and the Director-Reliability Engineering of significant findings of the program. The Manager-Quality As-surance further ensures that an overall Quality Assurance Audit Program Sche-dule is established and implemented,

b. The Manager-Quality Assurance has the author'ty and organizational freedom to schedu.<' and perform audits and to identify quality or management control problems and provide recommended solu-tions.
1869 336 IX-4 Rev. 8
APPENDICE! APPENDIX A Comparison Chart o f Quality Assurance Plan Requirements with those of various parts of the Code of Federal Regulations and Nuclear Industry Standards APPENDIX B Minimum Document Control Responsibility for =
"Important to Safety" Documents Quality Assurance Program APPENDIX C NRC Regulatory Guide Commitments and Exceptions O .
1869 337
APWEk .A W COMPARIS0tl CHART OF' QUALITY ASSURANCE PLAN REQUIREMENTS WITH THOSE OF VARIOUS PARTS OF THE CODE OF FFDERAL REGULATIOdS AND NUCLEAR INDUSTRY STANDARDS 10 CFR 50, App. B ANSI N45.2 10 CFR 71, App. E ANSI N18.7 - 1976 Criterion QA Plan Paragraph QA Plan _ Criterion QA Plan Paragraph QA Plan Paragraph QA Pla I 1.0 2.0 2.0 1 1.0 3.1 1.0 5.2.11 8.0 II 2.0 3.0 1.0 2 2.0 3.2 1.0 5.2.12 3.3 III 4.0 4.0 h.0 3 h.0 3.3 1.0 5.2.13 50 IV 5.1 5.0 5.1 L 5.1 3.4 1.0 5.2.14 8.0 V 3.1 6.0 3.1 5 3.1 3.5 2.0 5.2.15 3.0 VI 3.2 7.0 3.2 6 3.2 4.1 2.0/9.0 5.2.16 6.2.1.h VII 51 8.0 5.1 7 5.1 4.2 2.0/9.0 5.2.17 6.2.1.1 VIII 5.2 9.0 5.2 8 5.2 4.3 2.0 5.2.18 6.2.1.2 IX 6.2.1.2 10.0 6.2.1.2 9 6.2.1.2 h.4 2.0 5.2.19 6.2.1.3 x 6.2.1.1 11.0 6.2.1.1 7.0 4.5 9.0 5.3 6.2.1.12 XI 6.2.1.3 ]2.0 6.2.1.3 10 6.2.1.3 5.1 2.0 XII 6.2.1.4 13.0 6.2.1.4 7.0 5.2.1 2.0 XIII 6.2.1.5 14.0 6.2.1.5 11 6.2.1.3 5.2.2 3.1 XIV 6.2.1.6 15.0 6.2.1.6 7.0 5.2.3 3.1 XV 8.0 16.0 8.0 12 6.2.1.4 5.2.4 3.1 xv1 8.0 17.0 8.0 7.0 5.2.5 3.1 XVII , 3.3 18.0 ' 3.3 13 6.2.1 5 5.2.6 6.2.1.10 XVIII 9.0 19.0 9.0 7.0 5.2.7 6.2.1.11 14 6.2.1.6 5 2.8 6.2.1.13 7.0 5.2.9 6.2.1.8 15 8.0 5.2.10 6 2.1.10 16 8.0 17 3.3 18 9.0
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APPENDIX B (Notes) Doc ume nt Prepare $ By Reviewed by
Approved By/ Concurrence Issuel By (1,2,3)
C.5 TMI Station TMI Statica TMI Station Sections 'THI Station Section Manager / Supervisor ThG Station Section Sections QA Mod / Ops Manager QA Hod / Ops Manager Administration
.istructions C.6 Tt1I Special T1H Station PORC aUnit Manager THI Station Test Procedures Organizations CRC NRC Administration (Per 10 CFB 50.59) Unit Manager QA Mod / Ops Manager C.7 TMI Radiatior Radiation Badiation Protection
Radiation Protection Manager / ThG Station P utection Prctection Engineering Supervisor Administration Proc ed ur es D.1 Procurement Off Site Applicable Section Managers " Project Engineering Manager Project Engineering Hequisition Organizations QA Design / Procurement Manager QA Design / Procurement Manager Manager D.2 Procurement Site Organizations Applicable Section Managers

Manager - Plant Engineering Or13 1 nating Requisition QA Design / Procurement Supvr. QA Design / Procurement Supvr. Organizations E.1 Engir,eering ThG Ge::eration Applicable Section Managers " Project Engineering Manager TMl Generation (Notes)
Cttur.ge I:cmarandums Crvep QA Design / Procurement Manager QA Design / Procurement Manager Group (k,5,6) Engineering Applicable Section Manager Engineering E.2 Endineering TMI Plant Eng. Applicable Delertment Managers Applicable Department Managers TMI Plant Eng.*(Notes) Change Memorandums QA Mod / Ops Manager *TMI Plant Engineering (k,5,6) QA Design / Procurement Supvr. {0 Ts W u
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APPENDIX C QUALITY ASSURANCE PROGRAM NRC REGULATORY GUIDE COMMITMENTS AND EXCEPTIONS Engineering, in establishing specific requirements for design will use regulatory guide positions controlled oy Engineering in a project criteria document. Examples of positions taken relative to regulatory guides are listed. Those identified by an asterisk cover regulatory guides which are specifically quality related or impacteo and are therefore controlled by this manual. The TMI Quality Assurance Program complies witn Section C of the NRC Regulatory Guides indicated below. Exceptions to NRC Regulatory Guide position are detailed in Part 2 of this Appendix. This Appendix addresses additional Reg. Guioes not listed in Rev. 7 of the Operational Quality Assurance Plan. Compliance with these added Reg. Guides will apply to modi-fications, additions ano activities performed after issue of Rev. 8 and does not imply backfitting and/or retroactive compliance. It is also to be recognized that existing plant conditions, may prevent or preclude the satisfaction of all requirements of a specific design related regulatory guide. The deviation will De documented and, along with the justification, will be approved by the Manager of Design and Engineering. C-1 Rev. 8 s 1869 342 __ __; _ = - . . - -
-- ~ - ~ ~~ -
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( a < AFPL. DIX C, PART I JAI.~JABY, 1933 C? ? 'T'm *:' TO qu AIIT ASTURA!CF W FAL"?BY PUIFS FOR *::RFE MILE ICI A?;D DECEFE LF
!< r ? C- A';SI STD. COW LIAN2E REMAPES
1. C > / l i , !< e v . ;-R Persur.nel Selecticn and Training N18.1 1971 Madified Ces;,1y with "Regulatcry Positicr "

1. . , c/77, Etv. 2 Quality Assu.rar.ce Fregram BIquiremer.ts Nk5 2 1977 Full Comply with "Pegulatory Positicn" (Design ar.d Ccnstruction) 1.30 b/11/7. QA Requiremer.ts for the Installation, Inspection NL5.2.L 1972 Full Comply with " Regulatory Position" and Testing'cf Instrumentation and Electrical I. quit te r.t

1.33 2/7c, Fev. 2 Quality Assurance Prcgram Pequirements N18.7 1976 Madified see alternate method attached.
(C;erstien) 1.37 3/16/73 QA Require ents fcr Cleaning of Fluid Systems NL5.2.1 1973 Modified see alternete method attact.ed. ar.d Associatei Cc=ponents of Water Cocled liaclear Power Plants 1.jb >/77, Bev. 2 QA Bequirements for Packegir.g. Chirpirg, NL5.2.2 1972 Madified see alternate, method attached. Ecceivirg, Ctcrage ar.d lianaling of ! tens f;r Water Co:. led Nuclear Fower Plants 1.J9 'f /77, hev. 2 }icusekeepir.g Fequire=ents for Water Cooled NL5.2.3 1973 Full Corply with " Regulatory Position" buclear Pcwer Plar.ts
1. % t/73 C3 QA Fequirements for Prctective Coatings 101.L 1972 Modified See alternate method attached.
'A Applied to Water Cocles I.uclear Pcwer Plants G
l.53 7/79 Qualifications of Nuclear Power Plant Inspection, Nk5.2.6 1978 Modified see alternate method attacted.
f rato;.ca !< ev. 1 Exesinaticn and Testing Personnel '1.t.l c/ 7t. Rev. 2 D Quality Assurance Pequirements for the Design Nk5.2.11 1971 Modified See alternate method attached. U cf .uclear Pcwer Plants
l.7k 2/7b Quality Assurance Terms arid Definitions NL5.2.10 1973 Pull Comply with " Regulatory Pcsition" C-2
i ' AFrumiX C, PA3T I JA*rJ GY , 19%
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Nk5.2.9 lg7L M ;-d i n e d See alterr. ate method attac!.ed. ICj,c, bev. 2 Cclie: tion Sterage end "aintenance of
'l.Cd
{ ..melor Fa er f lar.t Q;ality Assurance Peccrds i QA Repirecents for Installation Inspection Nk5.2.5 197k Modified See alternate rettod attached. l .91 4 / it , bev. 1 ar.d Testing cf Etructural Cc:. crete & S teel curir.g huelcar E uwer Plur.t ir.structicn Nk5 2.8 1975 Modified See alterr. ate method attached. 1.116 S/77, bev. U- B QA Rewirerents fcr Installaticn, Jr.spection a:.1 ~estir.g cf Mechanical Equipment and Systems
/ Nk5 2.131976 Full Comply with "Begulatory Position" i
I 'l .12 3 7/ 77, hev . 1 QA Bequirements for ..cr. trol of Precure=ent of 1 Iters and Services for Naclear Power Plants Nk5.2.12 1977 Modified See corcnents attached.
*l.13.k 1/7) Fequirements for Auditing of Quality Assurance Frc re_e.s for :.aclear I ;wer Plants Mclified See alternate method attached.
1.06 2 / 7 t' , bev. ; QA Classificatior.s and Ctandards for Vater stre c and Ead u ctive Waste Ccntair.ing C; ; :.eus cf :;a:1 ear Fc.er I l a:.t s
.r 1
Modified 1.d1 I./7c, bev. 3 Ccr. trol cf Ferrite C.: .en, in Stainless Steel Weld }'etal I.EEE-317 1976 Modified See clarificatien attached. 1.63 8/70, Bev. 2 Electric Tenetration Assen.tlies in Contair.nent l Structure fcr Light Water Cooled Nuclear Power . l j CD Plar.ts t :D u 4 4 1 l l
APPENDIX C, PART 2 NRC Regulatory Guide 1.30, August. 1972 Quality,f.ssurance Reauirements for the Installation, Inspection and Testing of Instrumentation ano Electric Equipment Met-Ed shall comply with the Regulatory Position estab-lished in this Regulatory Guide in that QA program-matic/ administrative requirements included therein shall apply to maintenance and modification activities even though such requirements were not in effect originally. Technical requirements associated with maintenance and modifications shall be the original requirements or better (e.g., code requirements, material properties, design mar-gins, manufacturing processes, and inspection requirements). NCR Regulatory Guide 1.33, Rev. 2, Feoruary 1978 Quality Assurance Program Requirements (Operation) The TMI QA Program complies with the regulatory position of this guide with the following clarifications:
1. Paragraph C.4.a is interpreted to mean audits will be made once each 6 months to verify the nonconfor-mances and corrective action program is properly implemented and documented, particularly as related to actions taken to correct deficiencies that affect items important to safety.

2. Paragraph 5.2.8 of ANSI N18.7 - 1976 titled-"Sur-veillance Testing and Inspection" In lieu of a " master surveillance" schedule, a surveillance testing schedule shall be established reflecting the status of all inplant surveillance tests and inspections.

3. Paragraph 5.2.15 of ANSI N18.7 - 1976 titled
" Review, Approval and Control of Procedures" The third sentence of the third paragraph is inter-pretea to mean applicable procedures shall be reviewed following a reportable incident such as an accident, an unexpected transient, significant operator error, or equipment malfunction.
4. Paragraph 5.2.17 of ANSI N18.7 - 1976 titled
" Inspections" 18A9 345 C-4 Rev. 8
APPENDIX C, PART 2 Not all inspections will require a separate inspec-tion report. Inspection requirements may be inte-grated into appropriate procedures or other documents with the procedures or documents with the procedure or document serving as the recoro; how-ever, records of inspections will be identified and retrievable. NRC Regulatory Guide 1.3.7, March 16, 1973 Quality Assurance Requirements for Cleaning Fluids Systems and Associated Components of Water Cooled Nuclear Power Plants The TMI Quality Assurance Program complies with the regula-tory position of this guide with the following classifica-tions:
1. The second sentence of paragraph C.3 should be amended to read:
"The water quality for final flushes of fluid sys-tems and associated components shall be at least equivalent to the operating systems water, except for the oxygen nitrogen content and this does not infer that chromates or other additives normally in the system water will be added to the flush water."
7. Paragraph C.4 should be amended to add:
Expendable material such as inks, temperature indi-cating crayons, labels, wrapping materials (other than polyethylene), water soluble dam materials, lubricants, NDT penetrant materials and couplants, which contact stainless steel or nickle alloy material surfaces shall not contain lead, zinc, copper, mercury or other low melting alloys or compounds as basic essential chemical consti-tuents. Prescribed maximum levels of water leach-able chloride ions, total halogens and sulfur com-pounds shall be imposed on expendable prooucts. 1869 346 C-5 Rev. 8
APPENDIX C, PART 2 NRC Regulatory Guide 1.38, Rev. 2, May 1977 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water Cooled Nuclear Power Plants The TMI Quality Assurance Program complies with the regula-tory position of this guide with the following modifica-tions:
1. Section 3.o of ANSI N45.2.2 - 1972 concerns preven-tion of halogenated materials from contacting stainless steel or nickle alloy materials. The position stated in Reg. Guide 1.37 also applies to this guide.

2. Section 3.7.1 of ANSI N45.2.2 - 1972 Cleated, sheathed boxes will be used up to 1000 lbs. rather than 500 lbs. as specified. This type of box is safe for, and has been tested for, loads up to 1000 lbs. Other material standards (i.e.,
FED Spec. PPP-B-601) allow this. Special c"111fi-cation testing sha.ll be required #or loads in excess of 1000 lbs.
3. Section 6/2/1 of ANSI N45.2.2 - 1972 For storage of level D items access will be con-trolled and limited by posting. Other positive controls such as fencing or posting of guards will be provided for higher storage levels.

4. Section 7.3 of ANSI N45.2.2 - 1972 Rerating of hoisting equipment will be considered only when necessary. Prior to performing any lift greater than the load rating, the equipment manu-facturer will be contacted for his approval and direction. The manufacturer will be requested to s'upply a document granting approval for a limited number of lifts at the new rating and any restric-tions involved such as modifications to be made to the equipment, number of lifts to be made at the new rating, and the test lift load. At all times the codes governing rerating of hoisting equipment will be observed.
1o^9 347 C-6 Rev. 8
APPENDIX C, PART 2
5. Section A.3.4.1 Appendix to ANSI 45.2.2 - 1972 The last sentence of A.3.4.l(4) and (5) should be corrected as follows:
(4) "However, preservatives for inaccessible inside surfaces of pumps, valves and pipe systems containing, reactor coolant water
-shall be the water flushable type."
(5) "The name of the preservative used shall be indicated to facilitate touch up."
6. With regard to Section A.3.5.2 of the Appendix to ANSI N45.2.2 - 1972 entitled " Tapes and Adhesives":
Tar cs will meet a sulphur limit of 0.30% by weight instead of 0.10% as specified in A.3.5.2(1)(a). This limit is reasonable based upon the chemical content of commercially available tapes will ee of a contrasting color rather than " Brightly Colored" as required by A.3.5.l(3).
7. With regard to Section A.3.7.1 of the Appendix to ANSI N45.2.2 - 1972 entitled " Fiberboard Boxes":
In lieu of A.3.7.l(3) and (4), the following will be imposed: Fiberboard boxes shall be securely closed either with a water resistant adhesive applied to the entire area of contact between the flaps, or all seams and joints shall be sealed with not less than 2-inch wide, water resistant tape. NRC Regulatory Guide 1.39, Rev. 2, September 1977 Housekeeping Requirements for Water Cooleo Nuclear Power Plants Endorses ANSI N45.2.3 - 1973 The Operational Quality Assurance Program complies with this guide with the following clarification:
1. With regard to Sections 2.1 and 3.2 of ANSI N45.2.3
- 1973 entitled " Planning and Control of Facilities", respectively:
The TMI Nuclear Station will not utilize the five level zone designation system, but will utilize standard j6nitorial and work practices to maintain a level of cleanliness commensurate with company i869 348 C-7 Rev. 8
APPEND 1X C, PART 2 policy in the areas of housekeeping, plant and personnel safety, and fire protection. Cleanliness will be maintained, consistent with the work being performed, so as to prevent the entry of foreign material into safety related systems. This will include as a minimum documented cleanliness inspections which will be performed immediately prior to system closure. Control of personnel, tools, equipment, and supplies will be estaolished when major portions of the reactor system are opened for inspection, maintenance of repair. Additional housekeeping requirements will be imple-mented as required for control or radioactive conlamination. NRC Regulatory Guide 1.54, June 1973 Quality Assurance Requirements for Protective Coatings Apolied to Water Cooled Nuclear Power Plants Endorses ANSI N101.4 - 1972 The Operational Quality Assurance Program complies with this guide with the following clarification:
1. Met-Ed snall comply with the Regulatory Position established in this Regulatory Guide in that QA programmatic / administrative requirements included therein shall apply to maintenance and modification activities even though such requirements were not la effect originally. Technical requirements associated with maintenance and modifications shall be the original requirements or better (e.g., code requirements, material properties, design margins, manufacturing processes, and inspection require-ments).

2. The guidance or Regulatory Guide 1.54 shall be followed for organic protective coatings selected and evaluated in accordance with pertinent sections of ANSI N101.2 when applied to interior surfaces of the containment. The supplier's quality assurance program shall be approved prior to implementation.
Quality Assurance documentation may not be similar to records and documents listed in Sections 7.4 through 7.8 of ANSI N101.4 but will be evaluated to assure that they provide at least the same degree of. documentation as required oy this standard. 1869 349 C-8 Rev. 8
APPENDIX C, PART 2 NRC Regulatory Guide 1.58, August 1973 Qualifications of Nuclear Power Plant Inspection, Examination, and Testing Personnel Endorses ANSI N45.2.6 - 1973 The Operational Quality Asssurance Program complies with this guide with the following clarification:
1. The guidance of Regulatory Guide 1.58 shall oe followed as it pertains to the qualifications of personnel who verify conformance of work activities to quality requirements. The qualifications of plant operating personnel concerned with day-to-day operation, maintenance, and certain technical ser-vices shall conform to Regulatory Guide 1.8.

2. Not all personnel who approve inspection and test procedures will be certified as meeting the Level III capability requirements of ANSI N45.2.6 - 1973, out personnel who approve inspection and test pro-cedures will be determined by raanagement, through evaluation of their eoucation, training and ex-perience, to be fully qualified and competent to approve such procedures. The oasis for the deter-mination will be documented.
NRC Regulatory Guide 1.64, Rev. 2, June 1976 Quality Assurance Requirements for the Desion of Nuclear Pow 1r Plants Endorses ANSI N45.2.ll - 1974 Met Ed shall comply with the Regulatory Position estab-lished in this Regulatory Guide in that QA program-matic/ administrative requirements included therin shall apply to maintenance and modification activities even though such requirements were not in effect originally. Technical requirements associated with maintenance and modifications shall be the original requirements or better (e.g., code requirements, material properties, design mar-gins, manufacturing processes, and inspection requirements). The Operational Quality Assurance Program complies with this guide with the following clarification to paragraph C.2(1) of Regulatory Guide 1.64: If in an exceptional circumstance only the designer's techni'cally immediate Supervisor is thei e 69 3 50 quali?ied irdividual availaole, this C-9 Rev. 8
APPENDIX C, PART 2 review can be conducted by the Supervisor, providing that: (a) the other provisions of tne Regulatory Guide are satisfied, and (b) the justification is individually documented and approved in advance by the Supervisor's management, and (c) quality assurance audits cover frequency and effectiveness of use of Supervisors as design verifiers to guard against aouse. NRC Regulatory Guide 1.94, Rev. 1, April 1976 Quality Assurance Requirements for Installation, Inspection ano Tt.: ting of Structural Concrete ano Structural Steel curing the Construction Pnase of Nuclear Power Plants Endorses ANSI N45.2.5 - 1974 The Operational quality Assurance Program complies with this guide with the following clarification: Met-Ed shall comply with the Regulatory Position estab-lishea in this Regulatory Guide in that QA program-matic/admiristrative requirements included therin shall apply to maintenance and modification activities even though such requirements were not in e f fect origin-ally. Technical requirements associated with main-tenance and modifications shall be che original requirements or better (e.g., code requirements, material properties, design margins, manufacturing processes, and inspection requirements). NRC Reoulatory Guide 1.116, Rev. 0-R, June 1976 Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems Endorses ANSI N45.2.8 The Operational Quality Assurance Program complies with this guide.with the following clarification: Met-Ed shall comply with the Regulatory Position estab-lished in this Regulatory Guide in that QA program-matic/ administrative requirements included th3 rein shall apply to maintenance and modification activities even though such requirements were not in effect originally. Technical requirements associated with maintenance and modifications, shall be the original requirements or better (e.g., code requirements, material properties, design margins, manufacturing processes, and inspection requirements). 1869 351 C-10 Rev. 8
APPENDIX C, PART 2 NRC Regulatory Guide 1.26, Rev. 3, February 1976 Quality Group Classifications and Standard for Water, Steam and Radioactive Waste Containing Components of Nuclear Power Plants Since the original design and construction of the Three Mile Island Plants was to different classification criteria than contained in this guide; Met-Ed will comply with the regulatory position of this guide with the following clari-fications:
1. For modifications to existing plant systems and for new construction, iters will De classified according to this guide providing such action will improve the safety of the system being modified or make a significant improvement in overall plant safety. Otherwise the items will be classified the same as the original design and construction.

2. Tie-in's to existing plant systems will be made.to the same or better code, standard and technical requirements which were applicable to the system to which the tie in is to be made.
NRC Regulatory Guide 1.63, Rev. 2, July 1978 Electric Penetration Assemblies in Containment Structures for Light Water Cooled Nuclear Power Plants Met-Ed will comply with the regulatory position of this guide with the following clarification: For modifications to existing structures and to new constructions, this guide will be utilized providing its use will improve the safety of the structure being modified or make a significant improvement in overall plant safety. Otherwise, the code, standard and tech-nical requirements applicable to the original design anu construction will be utilized. NRC Regulatory Guide 1.144, January 1979 Auditing of Quality Assurance Programs for Nuclear Power Plants Met-Ed is in basic agreement with the position set forth by your staff in your draft response to the subject regulatory guide. Listed below are comments to the major points raised in your response: C-ll lb Rev. 8
APPENDIX C, PART 2
1. Section C.3.a(2)
The proposed scheduling requirement for internal audits appears to change the basis for having a rational, programmatic approach to auditing. In its place, the new regulatory guide requires manda-tory auditing of all program elements on a yearly basis. The latter would require that all elements obtain the same attention regardless of importance, past performance, or to what extent other aspects of quality assurance measuring and evaluating tech-niques are used; as an example, the extent to which surveillance and process monitoring is used.
2. Section C.3.b(1)
We agree that source inspection provides some con-trolled basis for replacing the need for external audits. It is recommended that the use of quality assurance program surveillance should also be viewed as another alternative.
3. Section C.3.b(2)
We agree with the staff's position that the new regulatory guide wording will lead to %"dit proli-feration". While the licensee is responsible for procurement control. This can be exercised through an annual evaluation of the contractor's perfor-mance using pertinent results from manufacturing surveillance, source inspection, receiving inspec-tion, and other applicable factors. The evaluation would include a recommendation as to the need for a scheduled program or problem area audit. Hence, auditing, like surveillance and inspection, should be treated as a quality assurance tool used for evaluation. Furtnermore, the recommendation to audit should include provisions for reviewing the importance and impact of the particular contrac-tor's scope and status. NdC Regulatory Guide 1.88, Rev. 2, Octooer, 1976 Collection, Storage, and Maintenance of Nuclear Power Plant Availaoility Assurance Records Met-Ed will comply with this regulatory guide with the following clarification: 1869 353 . lll C-12 Rev. 8
s APPENDIX C, PART 2
1. With regard to Section 5-6 of f.NSI N45.2.6-1974 titled Facility, Met-Ed will comply with the re-quirements of Section 5-6 of the 1979 revision in lieu of the 1974 revision.
In order to reach full compliance with this modi-fled position, Met-Ed must contruct a new facili-ty. This facility is scheduled for construction in 1980.
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1869 354 ' C-13 Rev. 8}}

ML19257D373

Contents

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1.0 INTRODUCTION

8.1 INTRODUCTION

11.1 INTRODUCTION

RESPONSE

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ML19257D373

Text

1.0 INTRODUCTION

8.1 INTRODUCTION

11.1 INTRODUCTION

RESPONSE

Navigation menu

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