ML18139C126

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

ML18139C126
Person / Time
Site:	Surry
Issue date:	08/18/1982
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML18139C125	List: ML18139C126
References
CON-NRC-03-82-096, CON-NRC-3-82-96, RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.2.1, TASK-2.B.4, TASK-TM SAI-186-029-40, SAI-186-29-40, NUDOCS 8211220117
Download: ML18139C126 (14)

Text

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SAI-186-029-40 TECHNICAL EVALUATION REPORT IMPROVEMENTS IN TRAINING AND REQUALIFICATION PROGRAMS AS REQUIRED BY TM! ACTION ITEMS I.A.2.1 AND II.B.4 for the Surry Power Station, Units 1 and 2 (Dockets 50-280 and 50-281)

August 18, 1982 Prepared By:

Science Applications, Inc.

I!

1710 Goodridge Drive  !;

McLean, Virginia 22102 t,

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• Prepared for:

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract NRC-03-82-096 8211220117 821105 .

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Science Applications, Inc.

• e TABLE OF CONTENTS Section I. INTRODUCTION *** -1 I I. SCOPE AND CONTENT OF THE EVALUATION ** 1 A. I.A.2 .1: Immediate Upgrading of RO and SRO Training and Qualifications * * * *

• 1 B. II.B.4: Training for Mitigating Core Damage.

• 7 III. LICENSEE SUBMITTALS 7 IV. EVALUATION.

• 8 A. I.A.2.1: Immediate Upgrading of RO and S-RO Training and Qualifications . . * . . 8 B. II.B.4: Training for Mitigating Core Damage. 10 V. CONCLUSIONS 11 VI. REFERENCES.

• 12

• I. INTRODUCTION e

Science Applications, Inc. (SAI), as technical assistance contrac-tor to the U.S. Nuclear Regulatory Commission, has evaluated the response by Virginia Electric and Pov,er Company for the Surry Power Station, Units 1 and 2 (Dockets 50-280 and 50-281) to certain requirements contained in post-TMI Action Items I.A.2.1, Immediate Upgrading of Reactor Operator and Senior Reactor Op.erator Training and Qualifications, and II.B.4, Training for Mitigating Core Damage. These requirements were set forth in NUREG-0660 (Reference 1) and were subsequently clarified in NUREG-0737 (Reference 2).*

The purpose of the evaluation was to determine whether the litensee's operator training and requalifitation programs satisfy the requirements.-* The evaluation pertains to the following Technical Assignment

• Control (TAC) System numbers:

TAC Nos.

I.A.2.1 II.B.4 Unit 1 44201 44551 Unit 2 44202 44552

. As de l i neat e d be 1ow, the e v.a 1 u at i an covers on 1y some as p e ct s of item I.A.2.1.4.

The detailed evaluation of the licen*see'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 Upgrading of Reactor Operator and Senior Reactor Operator 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 (hereafter referred to as Denton's letter). This letter and enclosures imposes a number of training requirements on power rea~tor licensees. This evaluation specifically addressed a subset of the require-ments 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

.instructor requalification; and Section C, which addresses operator requali-fication. Some of these requirements are elaborated in Enclosures. 2, 3, and

• Enclosure 1 of NUREG-0737 and NRC's Technical Assistance Control System distinguish four sub-actions ~'t'ithin I.A.2.1 and two sub-actions within II.B.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.

1

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-ments 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 shodld consist of at least 80 contact hours* in both the initial. training and the re4uali~

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 i s not i nt e nde d t o be a pp l i ed r i g i d l y; r at he r , i t s pur po s e i s to *p r ov i de greater assurance of adequate course content when the licensee's training courses are not described in detail.

Since the licensees generally have their O\'ln unique course out-1i nes, 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 concluded from the licensee's description of his train-ing material that the items in the enclosures are covered.

The Institute of Nuclear Power Operations (INPO) 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 s am e re qu i rem en ts and are more th an ad e qu ate, i . e .., t r a i n i ng pr ogram s based specifically" on the complete INPO documents are .expected to satisfy all the requirements pertaining to training material v1hich* are addressed in this evaluation.

• The licensee's response 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 the nature of the *increase, or, if it addresses both -normal and abnormal transients (without necessarily indicating an increase in emphasis) and the requalifi-cation program satisfies the requirements for control manipulations, Enclo-sure 1, Item C.3. The latter requirement calls for all the manipulations listed.in Enclosure 4 (Figl}re 4 in this report) to be performed, at the frequency indicated, unless they are specifically not applicable to the licensee's type of reactor(s). Some of these manipulations may be performed on a simulator. Personnel with senior licenses may be credited \'Jith these acti~ities 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. requires about 32 specific manipulations over a two-year cycle 1<1hile 10 CFR 55 Appendix A requires only 10 manipulations over a tl'lo-year cycle.

• A contact hour is a one-hour period in which the course instructor is present or available for instructing or assisting students; lectures, seminars, discussions, problem-solving sessions, and examinations are considered contact periods. This definition is taken from Reference 4.

2

• e

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

Program Element NRC Requirements**

-** Enclosure 1, Item A.2.c(l)

I I Training programs shall be modified, as necessary, to provide training in heat transfer, fluid flow and thermodynamics. (Enclosure 2 provides guidelines for I the minimum content*of such training.)

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

PER~O:,NEL TR/;INING I Training programs shall be rr.cdified, as necessary to provide training in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged. (Enclosure 3 provides guidelines for the minimum content of such training.)

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

Training programs shall be modified, as necessary to provide increasec em~hasis I on reactor and plant transients.

I 1

I Enclosure l, Item A.2.e INSTRUCTOR Instructors shall be enrolled in appropriate requalification programs to assure REQUALIFICATION they are cognizant of current operating history, problems, and changes to pro-cedures and administrative limitations.

I Enclosure l, Item C.l Content of the licensed operator requalification programs sha11 be modifiec to l in"clude instruction in heat transfer, fluid flow, thermodyr.arr.i:s, and mitiga-tion of accidents involving a degraded core. (Enclosures 2 and 3 provide guide-lines for the minimum content of such training.)

I PERSOm!EL I Enclosure l, Item C.2 The criteria for requiring a licensed individual to participite in accelerated REQUALiFJCATIOti I requalification shall be modified to be consistent with the new passing grade for issuance of a license: 80~ overall and 70~ each category.

I Enclosure l, Item C.3 I Programs should be modified to require the control manipulations listed in I Enclosure 4. Normal control r.~nipulations, such as plant or reactor startups, must be performed. Control manipulations durin~ abnorr.~l or em~rgency opera-tions rr.ust be walked through with, and evaluated by, a member of the training I staff at a minimum. An appro;:iriate simulator m.;y be used to satisfy the requirements for co,ntrol manipulations.

i

• ihe requirements shown are a subset of those contained in Item I .A.2.l.

• • References to Enclosures are to Denton's letter of March 28, 1980, which is contained in the clarifi" cation of Item l.A.2.1 in NUREG-0737.

3

e e Figure 2. Enclosure 2 from Denton's Letter TRf,JN!NG IN HEAT TRANSFER, FLUID FLOW AliD THERMODYNAMICS

1. Basic Properties 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, specific weight, buoyancy, viscosity and other properties of fluids.* A working knowledge of steam/tables should also De included. Energy movement should be discussed including such fundamentals as heat exchange,

~pecific heat, latent heat of vaporization and sensible heat.

2. Fluid Statics.

This *section should cover the pressure, temperature and volume effects on fluids. Exar:iple of these parametric changes should be illustrated by the instructof and related calculations should be per.formed by the students and discussed in the training sessions. Causes and effects of pressure and temperature changes in the various components and systems should be discussed in the training sessions. Causes and effects of pressure and temperature changes in the various compone~ts and systems should be discussed as applicable to the facility with particular emphasis on safety significant features. The characteristics of force and pressure, pressure in liquids at rest, principles of hydraulics, saturation pressure and temperature and subcooling should also be included. *

3. Fluid Dynamics.

This section should cover the flow of fluids and such concepts as Bernoul11*'s principle, energy in moving fluids, flow measure theory and devices and pressure losses due to friction and qrificing.

Other concepts and terms to .be discussed in this section are NPSH, carry over, carry under*, kinetic energy, head-loss relationships and two phase flow fundamentals. Practical applications* relating tc the reactor coolant system and steam generators should also be included.

4. Heat Transfer by Conduction, Convection and Radiation.

Tnis section should cover the fundamentals of heat transfer by conductions. Tnis section should include discussions on such concepts and terms as specific heat, heat flux and atc:nic action. Heat transfer 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-tion should be discussed as applicable to the various systems at the facility. The convection current patterns created by expanding fluids in a confined area should be included in this section. Heat transport and fluid flow reductions or stoppage should be discussed due to steam-and/or r.oncondensib1e gas formation during*normal and accident conditi~ns.

This section should cover the fundamentals of heat transfer by thermal radiation in the form of radian:

energy. The electromagnetic energy emitted by a body as a result of its te~pera .ure should be discussed and illustrated by the use of equations and sample calculations. Comparisons should be made of a black body absorber and a white body emitter.

5. Change of Phase* Boiling.

This section should include descriptions of the state of matter, their inherent characteri' ics and thermodynamic properties such as enthalpy and entropy. Calculations should be performed i.rvolving stea:n quality and void fraction properties. The types of boiling should be discussed as appiicable to the facility during normal evolutions and accident conditions.

6. Burnout and Flow Instability.

This section should cover lescriptions and mechanisms for calculating such ter;:;s as critical flux, critical power, ONB ratio and hot channel factors. This section should also include instructions for

• preventing and ~onitoring for clad or*fuel damage and flow instabilities. Sam;,le calculations shculc be illustrated by the instructor and calculations should be performed by the studects and discusse~ in the training sessions. Methods and procedures for using the plant computer to de:ermine quantita:ive values of various factors during plant operation and plant heat balance determina:ions should also De covered in this section.

7. Reactor Heat Transfer Limits.

This section should include a discussion of heat.transfer limits by examining fuel rod and reactor design and limitations. The basis for the limits should be covered in this section along with recommended methods to ensure that limits are not approached or exceeded. This section should cover discussions of peaking factors, radial and axial power distributions and changes cf these factors due to the influence of other variables such as moderator temperature, xenon and control rod position:

4

e e Figure 3. Enclosure 3 from Denton s Letter 1 TRAJNJNG CRITERIA FOR MITIGATING CORE DAMAGE A;* Jncore Instrumentation

1. Use of fixed or movable incore detectors to determine extent of core damage and geometry changes.

2. Use of thermocouples in determining peak temperatures; methods for extended range readings; methods for direct readings at terminal junctions.

3. Methods for calling up (printing) incore data from the plant computer.

B. Excore Nuclear Instrumentation (NIS) .

1. Use of NIS for determination of void forma.tion; void location basis for NJS response as a function of core temperatures and density changes.

C. Vital lns'trumentation

1. Instrumentation response in an accident environment; failure sequence (time to failure,* method of failure); indication reliability (actual vs indicated level).

2. Alternative methods for measuring flows, pressures, levels, and temperatures.

a. Determination of pressurizer level if all level transmitters fail.

b. Determination of letdown flow with a clogged filter (low flow).

c. Determination of other Reactor Coolant System parameters if the primary method cf measurement has failed.

D. Primary Chemistry

1. Expected chemistry results with severe cor~ damage; consequences of transferring small quantities of liquid outside containment; importance of using leak tight systems.

2. Expected isotopic breakdown for core damage; for clad damage.

3. Corrosion effects of extended illl11ersion in primary water; time to failure.

E. Radiation Monitoring

l. Response of Process and Area Monitors to severe damages; behavior of detectors when satL1rated; methoc for detecting radiation readings by direct measurement at detector output (overrangej detector); expected accuracy of detectors at different locations; use of detectors .to determine extent of .core damage.

2. Methods of determining dose rate inside containment from measurements taken outside containment.

F. Gas Generation

1. Methcds of H2 generation during an accident; other sources of gas (Xe, Ke); techr.iques fer *venting or disposal of non-condensib1es.

2. Hz flam,;ability arid expfosive limit; sources of Oz in containment or Reactor Coolant _System.

5

e e Figure 4. Control Manipulations Listed in Enclosure 4.

• =l I l CONTROL MANIPULATIONS. i *'

• l. P1ant or reactor startups to include a range that reactivity feedback from nuclear heat is noticeable 6nd heatup rate is estab1ished.

2. P.1 ant shutdo*,m.

'*3. *Manual control of steam generators and/or feedwater during startup and shutdown.

~. Boration and or dilution during power operation.

I

• 5. Aryy significant (greater than 10%) power changes in manual rod ,control or recirculation flow.

6. Any reactor power change of 10% o*r greater where load change is performed with load limit control or where flu~, temperature, or speed control is on manual (for HTGR).

• 7. Loss of coolant including;

1. significant PWR steam generator leaks

2. inside and outside primary containment

3. large and small, including leak-rate determination

4. saturated Reactor Coolant response (PWR).

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

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

'*10. Loss of core coolant flow/natural circulation.

11. Loss of condenser vacuum.

12. Loss of service water if required for safety.

13. Loss of shutdown cooling.

lt. Loss of component cooling system or cooling to an individual component.

I"

15. Loss of normal feedwater or normal feedwater syste~ failure.

.. 16. Loss of all feedwater (norma 1 and emergency).

17. Loss of protective system channel.

18. Mispositioned control rod or rods (or rod drops).

19. Inability to drive control rods.

20. Conditions r~quiring use of emergency boration or standby liquid control syste~.

21. Fue1 cladding failure or high activity in reactor coolant or offgas.

22. Turbine or generator trip.

I

23. Malfunction of automatic control system{s) ~nich affect reactivity. I

24. Malfunction of reactor coolant pressure/volume control system. I

25. Reactor trip.

26. Main steam line break (inside or outside containment).

27. Nuclear instrumentation failure(s}.

• Starred items to be performed annually, all others biennially.

6

e e B. II.B.4: Training for Mitigating Core Damage Item II.B.4 in NUREG-0737*requires that 11 shift technical advisors

~nd operating personnel from the plant manager through the operations chain to the licensed operators 11 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 the content of this training. 11 Plant Manager 11 is here taken to mean the highest ranking manager at the plant ~ite.

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

However, II.B.4 applies also to operations personnel who are not licensed

• and are not c and i d at es for 1i c ens es. Th i s may i nc l ude one or more of the highest levels of management at the plant. These non-licensed personnel are not 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 i'n mitigating core damage and related subjects.

-*so~e 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 commensurate with their responsibilities. Since this .imposes no additional demands on the program itself, we do not address it in this evaluation. It woul.d be appropriate for resident inspectors to verify that non-operating personnel receive the proper training *.

The re q u i red i mp l em en tat i-o n dates for a 11 item s have pas .s ed.

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

Mor*eover, the evaluation does not cover training program modi.fications that mi~ht have been made for other reasons subsequent to the response to Denton's letter.

III. LICENSEE SUBMITTALS The licensee (VEPCO) has submitted to NRC a number of items (letters and various attachments) which explain their training and requalification programs. These submittals, made in response to Denton's letter, form the information base for this evaluation. For the Surfy Power Station, Units 1 and 2, there were three submittals with attachments, for a total of nine items, *which* are listed below. The last six submittal items were in response to a request for additional information sent the licensee*

on April 29, 1982 (Reference 6).

1. Letter from B.R. Sylvia, Manager - Nuclear Operations &Maintenance, Virginia Electric & Power Co., to H.R. Denton, Director, Office of Nuclear Reactor Regulation, NRC. July 15, 1980. {l pg, with enclosure: item 2). NR'C Ace No:

8007170348. (re: Response to NRC 1 etter dated March 28, 1980).

7

e e

2. 11 Li c e n s e d Ope r at or Re q u a 1 i f i cat i on Pro gr a. m*i, VEPCO, Revised July, 1980. (11 pp, attached

• to item 1). NRC Ace No: 8007170350.

3. Letter from R.H. Leasbur.g, Vice President, Nuclear Operations, VEPCO., to D.G. Eisenhut, Director, Division of Licensing, N~C. December 9, 1981.(1 pg). NRC Ace No: 8112230378. (re: Revision 2 to.

NUREG-0737 response dated 12/15/80). r .

4. Letter from R.H. Leasburg, Vice President, Nuc 1ear Operations, VEPCO., to S.A. Varga, Chief of Operating Reactors Branch #1, Division of Li c ens i n g, NRC. May 2 4, . 19 8 2. ( 1 pg, w i th enclosures: items 5, 6, 7, 8 & 9) .. (re: Response to NRC's RAI dated Apri 1 29, 1982)~

5. "Reply to Licensing Action Request for Additional Information". Undated (2 pp, attached to item 4).

6. 11 Station Organization for Training Identification".

Undated (1 pg, attached to item 4).

7. 11 SRO/RO 81-3 License Cl ass Outline". Undated 1 * (7 pp, attached to i tern 4 ).

8. "STA Trai'ning Course Outline". Undated. (11 pp, attached to item 4). (re: Course outline &

contact hours).

9. 11 RO Cl ass 81-3 Schedu 1e". Undated (5 pp, attached to item 4).

IV. EVALUATION SAI's evaluation of the training programs at Virginia Electric and

. Power Company's Surry Power Station, Units 1 and. 2, is presented below.

Section A addresses TMI Action Item I.A.2.1 and presents the assessment organiz.ed in the manner of Figure 1. Section B addresses TMI Action Item II.B.4.

A~ LA.2.1: Immediate Upgrading of Reactor Operator and Senior Reactor Operator Training and Qualification.

Enclosure 1, Item A.2.c(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 of detail specified in Enclosure 2 of Denton's letter~

• Submittal item 7, 11 SRO/RO 81:..3 License Class Outline," is a moderately detailed outline which lists eight .courses related to heat transfer, fluid flow and thermodynamics. These lectures cover the principal (numbered) topical areas in Enclosure 2 of Denton's letter. Based on the 8

~J;\,~~- e e limited details about the content of these lectures, it appears reasonable to conclude the required level of detail is provided.

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

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

Submittal item 7 lists eight courses re 1at i ng to mitigating core damage which cover the subject areas defined in Enclosure 3 of Denton's ietter. Submittal item 9 indicates the number of days devoted to training in the areas of heat transfer, fluid flow, thermodynamics and mitigating core damage*~* . Based on th i s we have est i mate d the tr a i n i ng i nvol ves 12 8 contact hours which is more than the required 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Therefore the NRC requirements are fulfilled.

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

I I,

Th~ requirement is that there be an increased emphasis in the I I

training program on dealing with reactor transients.

1:

Submittal item 9 indicates that five classroom days are devoted to I transient analysis and submittal item 5 ~tates, "Increased emphasis has be~n I placed on reactor and plant transients of both normal and abnormal nature. 1:I The use of the simulator has, in itself, provided the necessary increased emphasis but the contact hours have also been increased in this area." The I requireme11ts of this Enclosure 1 item have been met-.:

Enclosure 1, Item A.2.e The requirement is that instructors for reactor operator training programs be enrolled in appropriate requalification programs to assure they are cognizant of current operating history, problems and changes to procedures and administrative limitations.

Submit ta 1 item 5 states, "Instructors are kept apprised of current .

operating history, problem~ and changes to procedures and administrative limitations by required -procedure reviews and required reading which encompass the areas above plus many other subjects. All licensed instructors are in the VEPCO LORP. 11 This aspect of the requalification program meets the NRC requirement.

Enclosure 1, Item C.l The primary requirement- is that the requalification p~ograms have in*struction 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.

Submittal item 5, the most recent discussion of tLe reactor operator requalification program, stated that the requalification program uses the s~me lesson plans as the initial training progr~m evaluated 9

e e previously. Ten four-day training sessions are conducted periodically

~uring the calendar year and differ from the initial training program only 1n that there are fewer contact hours. The. decrease in contact hours is due to:

o The material covered is'. to recoup and reinforce previous knowledge, and

/

0 lhe press of the shift work schedule.

Because the technical course content was judged to be adequate for the training requirements it is similarly adequate here for the requalification requirements.

During the requalification program s~me 50 contact hours of training a(e devoted to heat transfer, fluid flow, thermodynamics and accident mitigation.

Submittal item 5 provides a description of the requalification program which is consistent with the sti pul ationC~19sures 2 and 3 of Denton's letter; however, the requirement f<(rt30 contad7hours training on heat transfer, fluid flow, thermodynamics a7ict--aee-toent mitigation has not been fulfilled. , Item C.2 The requir"ement for licensed operators to participate in the accelerated requalification program must be based on passing scores of 80%

ov~rall, 70% in eath category.

• Submittal item 2 states, 11 An overall grade of less than 80 per-cent, or a grade less .than 70 percent in any section,* shall require removal from licensed duties and participation in an accelerated requalification program." The requirement of this enclosure item is therefore fulfilled. , 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 nor~al and abnormal s~tuations. The specific manipulations required and their performance frequency are identified in Enclosure 4 of the Denton letter (see Figure 4 of this repoft).

Submittal item 2 identifies the control manipulations whic'., are to be performed annually and those to be performed biennially. The fllanipula-tions listed in Submittal item 2, while arranged in a somewhat different order, are identical to those listed in Enclosure 4 of Denton's letter. The requirements of Enclosure 4 to Denton's letter are met.

B. II.B.4 Training for Mitigating Core Da~age

• Item II.B.4 requires that training for mitigating core damage, as indicated in Enclosure 3 of Denton's letter, be gi_ven to shift technical advisors and operating personnel from the plant manager to the licensed operators. This includes both licensed and non-licensed personnel.

10

e e With regard to licensed operators, personnel wh~ have completed or are completing the initial training program discussed in connection with items A.2.c(l) and A.2.c(2) will have satisfied the training requirements of II.B.4. Those who received accident mitigation training in the requalification program for licensed personnel (Enclosure 1,. item C.1) do not meet the_ requirements of II.B.4 in that it does not provide the stipulated number of training contact hours and does not, therefore, fulfill the requi,r~ments of I I.B.4.

• The tr a i n i. ng of non - l i cens e d opera t i on s person ne l and Sh if t Technical Advisors as required by II.B.4 was also evaluated. The training program for Shift Technical Advisors (Sub~ittal item 8) contains materials which address all the requirements of Enclosure 3 and therefore exceeds the re qu i re ment s for co urs e cont e nt. Submitt a l item 6 prov i de s a 1 i s t i ng of non-licensed operations personnel having received training. The submittal does not indicate training having been received by the Station Manager. SAI concludes that this position, the high-est onsite operating manager, must be trained before this part of the I I.B.4 requirements are met.

V. CONCLUSIONS It is concluded that the Virginia Electric and Power Company training programs at Surry Power Station, Units 1 and 2 do not* fully meet the requirements of TMI Action _Items I.A.2.1 and II.B.4.

The jirst exception i~ that.the.ritjuisite numb2r of contact training,hours {80) in mitigating coie damage a~d related subjects re not given during the licensed operator requalification program, therefore the requirements of Item I.A.2.1, Enclosure 1, Item C.1 are not met.

The second exception is that if the training given in core damage mitigation, as required by TMI Action Item. II.B.4, was the same as for that subject provided during the licensed operator requalification p~ogram the requirements of II.B.4 have not been fulfilled for the reason stated above.

Finally, failure to provide training in ~itigatin~ core damage to the Station Manager precludes fulfillment of the II.B.4 requirement. All other aspects of the requirements are met.

11

e VI. REFERENCES

1.

• 11 NRC Action Plan Developed as a. Result of the TMI-2 Accident." NUREG-

• 0660, United States Nuclear Regulatory Commission. May 1980.

2. 11 Clarification of TMI Action Plan Requirements," NUREG-0737, United States Nuclear Regulatory Commission. November 1980. 1 ~

• * *** ** - - 1.* * *

3. The NRC requirement for 80 contact hours is' an Operator Licensing Branch techn i ca 1 position. It was int l uded with the acceptance criteria provided by NRC. to SAI for use in the present ev~luatton. See letter, Harley Silver, Technical Assistance Program Management Group, Division of Licensing, USNRC to Bryce Johnson, Program Manager, Science Applications, Inc.,

Subject:

Contract No. NRC-03-82-096, Final Work Assign~ent 2, December 23, .1981. * *

4. 11 Guidelines for Heat Transfer, Fluid Flow and Thermodynamics Instruction, 11 STG-02, The Institute of Nuclear Power Operations .

. December 12, 1980. *

5. 11 Guidelines for Training to Recognize and Mitigate the Consequences*of Core Damage, 11 STG-01, The Institute of Nuclear Power Operations.

• JanGary 15, 1981. * '

6. "Licensing Action Request for Additional Information" NRC to Virginia Electric and Power Company, 29 April 1982.* -

• 12