ML20151H242

Improvements in Training & Requalification Programs as Required by TMI Action Items I.A.2.1 & II.B.4, Technical Evaluation Rept

ML20151H242
Person / Time
Site:	Hatch
Issue date:	07/01/1982
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20151H244	List: ML20151H242
References
TASK-1.A.2.1, TASK-2.B.4, TASK-TM SAI-186-029-34, SAI-186-29-34, NUDOCS 8207080458
Download: ML20151H242 (14)

Text

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SAI-186-029-34 TECHNICAL EVALUATION REPORT IMPROVEMENTS IN REACTOR OPERATOR AND SENIOR REACTOR OPERATOR TRAINING AND REQUALIFICATION PROGRAMS for the Hatch Nuclear Power Plant Units 1 and 2 (Docket 50-321 and 50-366)

July 1, 1982 Prepared By:

Science Applications, Inc.

. 1710 Goodridge Drive McLean, Virginia 22102 Prepared for:

U.S. Nuclear Regulatory Conunission Washington, D.C. 20555 Contract NRC-03-82-096 pi x).

' Science Apphcations.Inc.

TABLE OF CONTENTS ,

1 Section Page I. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . 1 II. SCOPE /30 CONTENT OF THE EVALUATION . . . . . . . . . 1 A. I.A.2.1: Immediate Upgrading of R0 and SR0 Training and Qualifications. . . . . . 1

6. II.C.4: Training for Mitigating Core Damage. . 7 III. LICENSEE SUBMITTALS . . . . . . . . . . . . . . . . . 7 IV. EVALUATION. . . . . . . . . . . . . . . . . . . . . . . 8 A. I.A.2.1: Imnediate Upgrading of R0 and SR0 Training and Qualifications...... 8 B. II.B.4: Training for Mitigating Core Damage. . 10 V. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . 11 VI. REFERENCES. . . . . . . . . . . . . . . . . . . . . . 12 I

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I. INTRODUCTION j

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Science Applications, Inc. (SAI), as technica1 assistance l j

contractor to the U.S. Nuclear Regulatory Commission, has evaluated the l response by Georgia Power Company for Hatch Nuclear Power Plant, Units 1 and 2 (Docket 50-321 and 50-366) to certain requirements contained in post-TMI Action Items I.A.2.1, Immediate Upgrading of Rear. tor Operator and Senior Reactor Operator Training and Qualification, and II.B.4 Training for Miti-gating Cors Damage. These requirements were set forth in NUREG-0660 (Ref-erence 1) and were subsequently clarified in NU3EG-0737 (Reference 2).*

The purpose of tne evaluation was to determine whether the licensee's operator training and requalification programs satisfy the requirements. The evaluation pertains to the following Technical Assignment Control System numbers:

TAC Nos.

I.A.2.1 II.B.4 Unit 1 44166 ' 44516 Unit 2 44167 44517 As delineated below, the evaluation covers only some aspects of item I.A.2.1.4.

The detailed evaluation of the licensee's submittals is presented in Section IV; the conclusions are in Section V.

II. SCOPE AND CONTENT OF THE EVALUATION A. I.A.2.1: Immediate Uograding of R0 and Sr.0 Training and Qualifications The clarification of TMI Action Item I.A.2.1 in NUREG-0737 incor-porates a letter and four enclosures, dated March 28, 1980, from Harold R.

Denton, Director, Office of Nuclear Reactor Regulation, USNRC, to all power reactor applicants and licensees, concerning qualifications of reactor operators (hereaf ter referred to as Denton's letter). This letter and enclosures imposes a number of training requiremerits on power reactor .s licensees. Tnis evaluation specifically addressed a subset of the require-melts stated in Enclosure 1 of Denton's letter, namely: Item A.2.c, which relates to operator training requirements; item A.2.e, .which concerns inst: uctor requalification; and Section C, which addresses operator requali-fication. Some of these requirements are elaborat6d in Enclosures 2, 3, and

• Enclosure 1 of NUREG-0737 and ARC's Technical Assistance Control System ,

distinguish four sub-actions within I.A.2.1 and two sub-actions within I I .8.4. These subdivisions are not carried forward to the actual presentation of the requirements in Enclosure 3 of NUREG-0737. If they had been, the items of concern here would be contained in I.A.2.1.4 and II.B.4.1.

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4 of Denton's letter. The training requirements under evaluation are sum-marized in Figure 1. The elaborations of these requirements in Enclosures 2, 3 and 4 of Denton's letter are shown respectively in Figures 2, 3 and 4.

As noted in Figure 1, Enclosures 2 and 3 indicate minimum require-

.aents concerning course content in their respective areas. In addition, the Operator Licensing Branch in NRC has taken the position (Reference 3) that the training in mitigating core damage and related subjects should consist i .

of at least 80 contact hours

• in both the initial training and the requali-fication programs. The NRC considers thermodynamics, fluid flow and heat transfer to be related subjects, so the 80-hour requirement applies to the combined subject areas of Enclosures 2 and 3. The 80 contact hour criterion is not intended to be applied rigidly; rather, its purpose is to provide greater assurance of adequate course content when the licensee's training courses are not described in detail.

Since the licensces generally have their own unique course out-lines, adequacy of response to these requirements necessarily depends only on whether it is at a level of detail comparable to that specified in the enclosures (and consistent with the 80 contact hour requirement) and whether it can reasonably be conclu,ded from the licensee's description of his train-ing material that the items in the enclosures are covered.

The Institute of Nuclear Power Operations (INP0) has developed its own guidelines for training in the subject areas of Enclosures 2 and 3.

These guidelines, given in References 4 and 5, were developed in response to the same requirements and are more than adequate, i.e., training programs based specifically on the complete INP0 documents are expected to satisfy all the requirements pertaining to training material which are addressed in this evaluation.

The licensee's resp ase concerning increased emphasis on tran-sients is considered by SAI to be acceptable if it makes explicit reference to increased emphasis on transients and gives some indication of .he nature of the increase, or, if it addresses both normal and abnormal transients L (without necessarily indicating an increase in emphasis) and the requalifi-

! cation program satisfies the requirements for control manipulations, Enclo-l sure 1, Item C.3. The latter requirement calls for all the manipulations

' listed in Enclosure 4 (Figure 4 in this report) to Tiiperforined, at the frequency indicated, unless they are specifically not applicable to the licensee's type of reector(s). Some of these manipulations may be performed on a simulator. Personnel with senior licenses may be credited with these activities if they direct or evaluate control manipulations as they are performed by others. Although these manipulations are acceptable for meet-ing the reactivity control manipulations required by Appendix A paragraph 3.a of 10 CFR 55, the requirements of Enclosure 4 are more demanding.

, Enclosure 4 requires about 32 specific manipulations over a two-year cycle l while 10 CFR 55 Appendix A requires only 10 manipulations over a two-year cycle.

• A contact hour is a one-hour period in which the course instructor is l

present or available for instructing or assisting students; lectures, i seminars, discussions, problem-solving sessions, and examinations are considered contact periods. This definition is taken from Reference 4.

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w Figure 1. Training Requirements from TMI Action Item I.A.2.1*

Program Elemer:t h4C Raoutrements**

Enclosure 1. Item A.2.c(1) '

I .Trairing programs st:all be modified, as necessary. :: :-evide training in Peat traMfer, fluid flow and thermodynamics. (Dc':s.re 2 :*ovides guicettnes for l; the sinima content 6f 5.ch training.)

ope 2AT10NS Enclosure 1. Item A.2.c(2) pea 50'04E-. Training programs shall be modified, as necessaef 10 scovide training in the use of installed plant systees te control or -*"ine sa accident in anien tne Tu!W y care severely damaged. (Enclssure 3 provices ;. el'aes for the minimur l cor "c of such training.)

' Enc 1 sure 1'. Item A.2.c.(3)

Training prograi s shall be modified, as necessary to provide increased s'or. asis

= or, reactor and plant transients.

! En sure 1. Item A.2.e h ItiST M TOR l lertructors small be enrolled in appropriate recualification progrars to usure

" triey are cognizact of current operating nistory, problems. ace crarges to pro.

agg;r;ca;;c., j gedures and administrative litritations. \. .

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' I Content of the 11cansed operator recualification prcgrarps shall ee modified to triclude instruction in heat transfer, fluid flow, thermodynamics, and *itiga.

tion of a.cidents involving a degraded core. (Enclosur es 2 and 3 provide guide.

'I, lines for thgyintmum content of sucn tratning.) y s Encidsu r e 1. Itan C.2 ,

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The CPUeria for recuiring a licensed indivihal to participate 'In accele.*sla[

REQUALIFICATION requalification shall be modified to be conshtent witn the n?w.patsing grtde'

• for issuap:s of a license: 80'. ove's11 and ,?0*. eacn category.

Enclosure 1. Item C.3 Programs should be ewified to reautre the catrol manipulatioet listed in Enclossre 4 Norma' control mentoulations, hen es s'ent:ar enctor startuos.

must be performed. Control manipulations during ebstmal or emergency opera.

I tions =st be walked inrough with, and evaluernd th a member of 'the training staff at a minimum. An appropriate simulator #Jy be used to satisfy the requirements for control manipulations.

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• The recutrements shown are a subset of those contained tr Item I.A.2.1. <

"eeferences to Enclosures are to Centon's letter of Maren 28. 1960. unien is contained in the clari4'.

satsu of item r.a.2.i in ws 0m. t. , ,

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Figure 2. Enclosure 2 from Denton't u tter ~

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TRAININGINMEATTRAh3FER.FLUIDFLOWANDThf,RMdQvhAMIC5 f l

1. 8asic Peoperties of Fluids and Matter.

This section should cover a basic introduction to matter and its properties. This section should include such Concepts as temperature measurements and effects, density and its effects, speCif tC weight, buoyancy, viscosity and other properties of fluids. A working knowledge of steam tables should also be included. Energy movement should be discussed including such fundamentals as heat enchange, specific heat. latent heat of vaporisation and sensible heat.

2. Fluid Statics.

This section should Cover the pressure, temperature and volume effects on fluids. Example of these Da-ametric Changes should be illustrated by the instructs and related Calculations should be performed of t*e st6 dents and discussed in the training sessions. Causas and effects of pressure and temperature Causes and Maqes in the various Components and systems should be discussed in the training sessions.

eHetts of pressure and temperature changes in the various Components and systems should be $1scussed features. Tne I

as soplicable to the f acility with partiCular emphasis on safety significant I C9eracteristics of forte and pressure, pressure in liquids at rest. principles of hydraulics.

! sataration pressure and temperature and subcooling should 4!so be included.

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l 3. n .is Dynam Cs.

i Ns section should Cover the flow of fluids and such Concepts as Iernoulli's principle. ent*1y in l ming fluids, flou measure theory and devites and oressure losses due to fr1Ctton and orificing.

I Ot9e* Concepts and terms 10 Se discussed in this section are NPSM Carry over, carry under, kinetic f*e*gy.

Road. loss relation $Nps and two phase flow fundamentals. PractiGal appitCations relating to l lae reactor Coolant system and steam gene *ators'shoWId SIso be included.

I 4. Meat Traasfee by Conduction. Convection and eddtation.

Tnis section should cover the fundameatals of heat transf er by Conductions. This section sno.ld Heat include discussions on such Concepts and terms as specif f t nest. Rest flus and atomic action.

trarsfer Characteristics of fuel rods and heat exchangers should be included in this section.

This section should cover the fundamentals of heat transfer by Convection. Natural and forced Circula.

Ine ConveClion Current tion should be discussed as applicable to the various systems at (Pe f 4Clltty.

l d in this section. = tat

' patterns trsnSportcreated and fluidby espanding fluids in a Confined area should be inc udereductions or Stoppage should be flow

[ [. " gas formation during normal and accident Conditions.

This section should Cover the fundamentals of heat transfer by thermal radiation in tne form r of radiant energy. The electromagnet 1C energy emitted by a body as a result of its temperatushould Comparisons e should be be sa:e discussed and illustrated by the use of equations and sample calculations.

of a blatt body absorber and a unite body emitter.

5. Chance of Phase - Sotlina. '

their innerent Characteristics and I'n This section should inClvde destetations of the state of matter, Calculations should be performed involving thermodyn4mit proMeties such as enthalpy and entropy.The types of boiling should be discussed as applitaole to

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steam quality and void fraction properttes.

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s *Me f acility during normal evolutions and accident Conditions.

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$. 9eut and Flow lastaoility.

\ , Tnis settien should Cover descriptions and mechanisms for Calculating such teras as Critical flus.

Critical power, DNS ratio Jed hot Channel ' actors. This section should also include instructions foe

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' $ ample Calculations should

,(' \' preve9tify and monitoring forL

• lad or fueNamage and flow instabilities.oe illustrated by th g the training selstons. Methods and p*dtidures for using the plant Computer to determine cuantitative w

• values of various f actors durir.g plant detration and plant heat balance determinations should also be i dovered in this section.

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7.yeactor west Transfer Limits.

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This section should include a discussion of heat transfer limits by esacining f uel rod and reac design and limitations.

reCcomended methods to ensure tht limits are not approached or esCeeded. This section should cover 8

discussions of peaking factors. fadial and asf al power distributions and Changes of these f actors due s" "k to the influence of other variables such as moderator tegerature, senon and Control rod posttion.

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Figure 3. Enclosure 3 from Denttn's Letter TGAINING CRITERIA FOR MITIGAfinG Cast DAMAGE

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A. Incore Instepentation

1. Use of fined or novanie incere detectors to deterune estant of core damage and geometry changes.

2. Use of thermocouples in deteretning een temperatures; setnods for estended range readtags; methods f or direct readings s'. ter91nal junctigns.

3. Metnods for calling up (printing) inCore data from int plant Comouter.

t 8. Escore Mucitar Instrumentation (NIS)

I Use af NIS for determination of void formation; void location tests for NI$ response as a function 1.

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of core temoeratu res and denstty enanges.

I a C. Vital Instrumentation r

1. Instr mentation u response 18 an scetdent environme91; f atlur e sequence (ttee to f atlu e, met 9cd of f ailure); indication Fellantltty (actual vs indicated Ievel).

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2. Alteenative metnods for measu r ing flows, pressures, levels, and ted9f#$ lures.

I a. Cetermination of pressu r tter level if all level transmitters f ail.

N. 04teemination of letdo.n f rom och a clogged filter (Iow (1o.).

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c. Determination of other Reactor Coolant System parametees tf the primary metnod of measu essat has fatted.

I I D. 8r wa v C w stry

1. Espected cnestst*y results atta severe core damase; csseewe*<es of transferring small cuanttttes of lieute outside contatronent; tesortance of uste; ives tty, systees.

2. Espected isotopte areaadown fur core damage; for claJ damage.

' 3. Corrosion effects of entended taspersion in primary eatee; time to f ailure.

E. eadiation monttoetaa r

1. Response of Process and Area Monitors to tevere damages; benavior of detectors =nen satu ated; method for detecting radiation readings by direct measurement at detector output (ovecranges detector); espected accuracy of detectors at diff erent locations; use of detectors to deteemine entent of core damage.

2. Metnods of deteretning dose rate inside containment from sessurements taaen outside containment.

F. Gas Geaeretton i

1. Metnods of Mp nge eration during en accident; otnee sources of gas (Ie. ce); tecnnteues for venting or disposal of non*CondenstDIes.

in containment or Reactor Coolant System.

2. "2 flansnantltty and emplosive limit; sources of 02 5

e Figure 4. Control Manipulations Listed in Enclosure 4.

CONTROL M4m!PuLAT!0m1

1. Piant v reactor startues to inciude a ran,e snat reutivity feedesca from nuciear neat aarition is noticenie sad neatus rate is estmiisned.

2. Plant snutdown.

+3. Manual control of steam generators and/or feed =ater durtng startys and sautdoen.

4 Boration and or etlution dur ing so ee cceration.

• 5. Any significant (greater than 13' ;ower changes in manual rod control or reCtrculation flo .

6, Any reactor ooser enange of 13 ;r geester enere load c.iange is per*0rmed attn load limit contest or enere flua, temperatu re, or s;ees c:ntrol is on manual (for M7*.4).

• ). Loss of coolant including:

1. significant PWR steas geae tu ieses

2. Instde an1 outside primacy :*tsweat ,

3. targe end small, inctuotag 'ess este detereinstion 1 4 saturated teactor Coolant response (P=E). l

8. Loss of instepent air (if simulated plant specific).

9. Loss of electrical power (and/or degraded power sources).

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'U. Loss of cor* coolant flow /natu*al circulation.

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11. Loss of condenser vacuum.

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12. Loss of service water if required for safety.

13. Loss of snutdown ' cooling.

14. Loss of component cooling system ce cooling to an individual ctroonent.

15. Loss of normal femater or normal feed =ater systee f atture.

• 16. Loss of all feed ster (normal and emergency).

17 Loss of protective system channel.

!8. Mispositioned control rod or rods (or rod drops).

l l 19. Inacility to drive control rods.

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20. Condittoas requiring use of emergency boration or stand 6p 1141d control system.

21. Fuel cladding f ailure or nign activity in reactor coolant or offges.

l t 22. Termine or generator trip.

l 2L walfunction of automatic control systee(s) iMen affect reactivity.

24 malfur<tton of reactor coolant pressure /volure control system.

25. Reactor trip.

26. Mata steam line 1preek (inside or outside containment).

! 27. nuclear instrumentstion fatture(s).

• 5 tarred items te te performed ahnuelty, all others tienntally.

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B. II.B.4: Training for Mitigating Core Dacage Item II.B.4 in NUREG-0737 requires that " shift technical advisors and operating personnel from the plant manager through the operations chain to the licensed operators" receive training on the use of installed systems to control or mitigate accidents in which the core is severely damaged.

Enclosure 3 of Denton's letter provides guidance on tne content of this training. " Plant Manager" is here taken to mean the highesi ranking manager at the plant site.

For licensed personnel, this training would be redundant in that it is also required, by I.A.2.1, in the < -erator requalification program.

.However, II.B.4 applies also to operaticas personnel who are not licensed and are not candidates for licenses. This may include one or more of the highest levels of management at the plant. These non-licensed personnel are nat explicitly required to have training in heat transfer, fluid flow and thermodynamics and are therefore not obligated for the full 80 contact hours of training in mitigating core damage and related subjects.

Some non-operating personnel, notably managers and technicians in instrumentation and control, health physics and chemistry departments, are ,

supposed to receive those portions of the training which are costmensurate with their responsibilities. Since this imposes no additional demands on gorogram itself, we do not address it in this evaluation. It wnold.bE

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appropriate for resident inspeci; ors i.u vm ify i.h at non-uperating personnel receive the proper training.

The required implementation dates for all items have passed. '

Hence, this evaluation did not address the dates of implementation.

Moreover, the evaluation does not cover training program modifications that might have been made for other reasons subsequent to the response to Denton's letter.

l III. LICENSEE SUBMITTALS The licensee (Georgia Power Co.) has submitted to NRC a number of items (letters and various attachments) which explain their training and requalification programs. There submittals, made in response to Denton's letter, form the information base for this evaluation. For Hatch 1 and 2, there were 2 submittals with attachments, for a total of 4 items, which are listed below.

1. Letter from M. Manry, Plant Manager, E.I.

Hatch, Georgia Power Co., to P.F. Collir.s, Chief of Operator Licensing Branch, NRC.

August 12, 1980. (1 pg, with enclosure:

7 item 2). NRC Acc No: 6009050291. (re:

l Response to NRC letter dated March 29,1980).

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1 j 2. Excerpt from the new training outline to be used for future license candidates for plant

. E.I. Hatch. Untitled, vndated. (6 pp, attached to item 1). (re: Course Outline). .

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! 3. Letter frcm J.T. Beckham,Jr., Vice President, Georgia Power Co., to the Director of Nuclear Reactor Regulation, NRC. May 6, 1982. (2 pp, with enclosure: item 4). (re: Transmittal of response to NRC RAI dated March 31,1982).

NRC Acc No: 8205110481.

4. " Response to RAI Regarding Upgraded SR0 & R0 Training & Training for Mitigating Core Damage - NUREG-0737. Items I.A.2.1 & II.B.4",

Georgia Power Co. Undated. (6 pp, attached to item 3). -

. Item 1 is the original submittal letter. Item 2 is a new training outline of several courses to be used for future license candidates.

Because it is an outline,it is only indicative of the actual trainingcg!Id N

_. A2equalification proggra Items 3 and 4 are in a combined submittal .in re'sMnse~tV a request for additional information prepared by SAI, dated February 24, 1982, and transmitted by NRC in a letter dated March 31, 1982.

Item 4 provided 2 course outlines which included the number of training hours involved. It should be noted that these 2 course outlinas are a part of the outlines identified in item 2.

IV. EVALUATION SAI's evaluation of the training programs at Georgia Power Company's Hatch Nuclear Power Plant, Units 1 and 2, is presented below.

Section A addresses TMI Action Item I.A.2.1 and presents the assessment

organized in the manner of Figure 1. Section B addresses TMI Action Item l

II.B.4.

(- A. I.A.2.1: Immediate Upgrading of Reactor Operator and Senior

! Reactor Operator Training and Qualification

! Enclosure 1. Item A.2.c(1_1 l

l The basic requirements are that the training programs given to reactor operator and senior reactor operator candidates cover the subjects of heat transfer, fluid flow and thermodynamics at the level uf detail

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specified in Enclosure 2 of Denton's letter.

Georgia Power Company's original submittal provided a training outline (submittal ites 2) which addresses item A.2.c(1). The first part of l, this training outline consists of seven major course outlines (mathematics,

( classical physics, chemistry, thermodynamics, fluids mechanics and heat

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transfer). Although the items identified in these major course outlines do not have one-for-one correspondence with the item: of Denton's Enclosure 2, SAI believes that the Hatch 1 and 2 training program meets this requirement because (1) there is significant detail in the training outline which 8

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. l matches Enclosure 2, and (2) in the response (submittal item 4) to NRC's ~

request for additional information (Reference 6), Georgja. Power Company stated that their level of instruction on these areas was comparable with Enclosure 2 of Denton's letter.

Enclosure 1. Item A.2.c(2)

The requirements are that the training programs for reactor and f senior reacter operator candidates cover the subject of accident mitigation at the level of detail specified in Enclosure 3 of Denton's letter (see Figure 3 of this report).

' Georgia Power Company's original submittal provided a training  ;

outline (submittal item 2) which addresses item A.2.c(2). The second part of this training outline has a course titled " Severe Core Damage Accident" l which does not explicitly identify all the items included in Denton's Enclo-  !

,sure 3. However, SAI has examined this course outline and believes that it does cover all the items in Enclosure 3 except possibly for the area titled

" Primary Chemistry." We believe that this item is covered under the topic of " Reactor Water Cleanup." Georgia Power Company also stated in submittal

item 4 that their level of instruction was comparable with Enclosure 3 of Denton's letter. Therefore, SAI believes this requirement is met at Hatch 1 and 2.

In the response to NRC's request for additional informatiion, Georgia Power Company'provided a training outline (submittal item 4) which addresses a training course of 120 contact hours related to training for mitigating core damage. In view of NRC's requirement for 80 contact hours, we take this as further evidence that the training program at Hatch 1 and 2 satisfies NRC's requirements regarding course content and level of detail.

Enclosure 1. Item A.2.c(3)

The requirement is that there be an increased emphasis in the training program on dealing with reactor transients.

Submittal item 2 identifies a course titled " Accident and Tran-sient Analysis" which includes 6 tooics, " Reactor Saf ety Criteria,"

" Inherent Reactor Protective Systems," " Plant Protection Systems," " Accident Analysis," " Reactor Safety Experience," and " Plant Response." Since this shows a significant emphasis on plant transients, SAI concludes that this requirement is met at Hatch I and 2.

Enclosure 1. Item A.2.e The requirement is that instructors for reactor operator training r

programs be enrolled in appropriate requalification programs to assure they i are cognizant 6f current operating history, problems and changes to I procedures and administrative limitations.

The original submittal letter (submittal item 1) stated that Georgia Power Company requires instructors to maintain active licenses under their current policy. This implies that instructors will go through a l

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periodic requalification program. In addition, the response to NRC's i - request for additional information stated that their instructors do enroll l

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4 1 in appropriate requalification programs to assure they are cognizant of current operating history, problems, and changes to procedures and adminis-trative limitation. Based on these f acts SAI concludes that this require-ment .is met at the Hatch 1 and 2 plant.

Enclosure 1. Item C.1 The primary requirement is that the requalification programs have instruction in the areas of heat transfer, fluid flow, thermodynamics and accident mitigation. The level of detail required in the requalification program is that of Enclosures 2 and 3 of Denton's letter. In addition, these instructions must involve an adequate number of contact hours.

Georgia Power Company provided a training outline (submittal item

2) on these subject areas which applied both to their training and requali-fication programs (i.e., the same training outlines identified in items A.2.c(1) and A.2.c(2)). Therefore the evaluat1on cf item C.1 is essentially equivalent .to the combined evaluation of items A.2.c(1) and A.2.c(2). SAI believes that this requirement is met at the Hatch 1 and 2 plant.

Enclosure 1. Item C.2 The requirement for licensed operators to participate in the .

accelerated requalification program must be based .on passing scores of 80%

overall, 70% in each category.

The response (submittal item 4) to NRC's request for additional information (Reference 6) stated that Georgia Power's requalification pro-gram does meet this requirement. We would credit Georgia Power Company with

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meeting this requirement.

Enclosure 1. Item C.3 TMI Action Item I.A.2.1 calls for the licensed operator requalifi-cation program to include performance of control manipulations involving both normal and abnormal situations. The specific manipulations required and their performance frequency are identified in Enclosure 4 of the Denton letter (see Figure 4 of this report).

In the response to the request for additional information, Georgia Power Company stated that their requalification program does call for the control manipulation training as specified in Enclosure 4. This commitment to both the content and performance frequency of Enclosure 4 means that Georgia Power Company meets this requirement. Since the list of manipula-tions is not included explicitly, Enclosure 4 may be used for auditing purposes.

B. II.B.4 Training for Mitigating Core Damage Item II.BA requires that training for mitigating core damage, as indicated in Enclosure 3 of Denton's letter, be given to shif t technical advisors and operating personnel from the plant manager to the licensed operators. This includes both licensed and non-licensed personnel.

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The training of the licensed personnel are in accordance with the requirements of Action Item II.B.4. This requirement is met when the licensee instituted the training and requalification programs discussed for Action Item I. A.2.1.

Based on information supplied by Georgia Power Company in their response to NRC's request for information, particularly in view of the licensee's organization chart provided in the response, it appears that the requirement to provide this training for non-licensed operations personnel is satisfied at the Hatch 1 and 2 plants. Specifically, this training is given to personnel holding the following positions: senior reactor opera-tor, reactor operator, shif t technical advisors, plant manager, assistant plant manager, operations superintendent, operations supervisors, shif t supervisors, shift foremen, superistendent of plant engineering and services. However, the licensee also stated that certain individuals may not have received this training due to their recent assignment to their position, but it will be provided during upcoming retraining. Therefore, the resident inspector should verify the completion of this training for those individuals.

i V. CONCLUSIONS Based on the evaluation discussed above, SAI concludes that the training programs at Hatch Nuclear Power Plant, Unit 1 and 2, meet the requirements of NUREG-0737 items I.A.2.1 and ,II.B.4.

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V. REFERENCES

l. "NRC Action Plan Developed as a Result of the TMI-2 Accident." NUREG-0660, United States Nuclear Regulatory Commission. May 1980.

2. " Clarification of TMI Action Plan Requirements," NUREG-0737, United States Nuclear Regulatory Comission. November 1980.

3. The NRC requirement for 80 contact hours is an Operator Licensing Branch technical position. It was included with the acceptance criteria provided by NRC to SAI for use in the present evaluation. See letter, Harley Silver, Tech Division of Licensing, USNRC'nical Assistance Program Management Applications, Inc.,

Subject:

Contract No. NRC-03-82-096, Final Work Assignment 2, December 23, 1981.

4. " Guidelines for Heat Transf er, Fluid Flow and Thermodynamics Instruction," STG-02, The Institute of Nuclear Power Operations.

December 12, 1980.

5. " Guidelines for Training to Recognize and Mitigate the Consequences of Core Damage," STG-01, The Institute of Nuclear Power Operations.

January 15, 1981.

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