ML20069N780

Improvements in Training & Requalification Programs as Required by TMI Action Items I.A.2.1 & II.B.4,Calvert Cliffs,Units 1 & 2, Technical Evaluation Rept

ML20069N780
Person / Time
Site:	Calvert Cliffs
Issue date:	08/31/1982
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20069N779	List: ML20069N780
References
CON-NRC-03-82-096, CON-NRC-3-82-96, RTR-NUREG-0660, RTR-NUREG-0737, RTR-NUREG-660, RTR-NUREG-737, TASK-1.A.2.1, TASK-2.B.4, TASK-TM SAI-186-029-19, SAI-186-29-19, NUDOCS 8212070115
Download: ML20069N780 (13)

Text

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i== , s SAI-186-029-19 TECHNICAL EVALUATION REPORT IMPROVEMENTS IN TRAINING AND REQUALIFICATION PROGRAMS AS REQUIRED BY TMI ACTION ITEMS I.A.2.1 AND II.B.4 .

for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (Dockets 50-317 and 50-318) es-August 17, 1982 Prepared By:

Science Applications, Inc.

,1710 Goodridge Drive

-McLean, Virginia 22102 Prepared for:

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract NRC-03-82-096 8212070115 821124 DR ADOCK 050003 Science Applications,Inc.

TABLE OF CONTENTS ,

Section Page I. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . 1 II. SCOPE AND CONTENT OF THE EVALUATION . . . . . . . . . 1 A. I.A.2.1: Immediate Upgrading of R0 and SR0 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 R0 and SRO Training and Qualifications ..... 8 B. II.B.4: Training for itigating Core Damage. . 10

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V. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . .' 10 VI. REFERENCES. . . . . . . . . . . . . . . . . . . . . . 11 9

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s I. INTRODUCTION Science Applications, Inc. (SAI), as technical assistance contrac-tor to the U.S. Nuclear Regulatory Commission, has evaluated the response by Baltimore Gas and Electric Company for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (Dockets 50-317 and 50-318) to certain requirements contained in post-TMI Action Items I.A.2.1, Immediate Upgrading of Reactor Operator and Senior Reactor Operator 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 licensee's operator training and requalification programs satisfy the requirements. The evaluation pertains to Technical Assignment Control (TAC)

System numbers :

I.A.l.2 II.B.4

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Unit 1 44148 44498 Unit 2 441'49 44499 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 Upgrading of R0 and SR0 Training and Qualifications The clarit u.ation of TMI Action Item I.A.2.1 in NUREG-0737 incer-porates a letter and four enclosures, dated March 28, 1980, from Harold R.

Denton, Director, Office of Nuclear Reactor Regulation, USNRC, to all power

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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 requirements on power reactor 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 requ.irements are elaborated in Enclosures 2, 3, and 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.

• Enclosure 1 of NUREG-0737 and NRC's Technical Assistance Control System distinguish four sub-actions within I.A.2.1 and two sub-actions within I I .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 11.8.4.1. ,

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

Program Elem nt h4C Requirements" Enclosure 1. Item A.2.c(1)

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

OPERATIO'45 Enclosure 1. Item A.2.c(2)

PER$0N4EL Training proarams shall be modified. as necessary to provide training in the TRAIN!r*

use of insta$ led 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.)

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

Training programs shall be modified, as necessary to provide increased emphasis on reactor ano plant transients.

Enclosure I. Item A.2.e INSTRUCTOR Instructors shall be enrolled in appropriate requalification programs to assure SCE7JALIFICATION they are cognizant of current operating history. problems, and changes to pro-cedures and administrative limitations.

! Enclosure 1.ItemC.1 Content of the licensed operator recualification programs shall be modified to in'clude instruction in heat transfer fluid flow. thermodynamics, and ritiga-tion of accidents involving a degraded core. (Enclosures 2 and 3 provide guide-

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lines for the minimum content of such training.)

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Enclesare 1. Item C.2

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m. i The criteria for rec iring a licensed individual to participate in accelerated .

, recaalificatien shall te modified to be consistent =itn the new passing grace

, for issuance of a license: 80 overall and 70'. each category.

fEnclosure1.IteaC.3 i Programs should be modified to recuire the control manipulations listed in Enclosure 4 herral control manipulations, such as clant or reactor startups, must be performed. Control manipulations during abnormal or emergency opera-tions rast te walked through with, and evaluated ty. a me-iter of the training staff at a minimum. An appropriate simulator may be used to satisfy the requirenents for control manipulations.

'Tre recuirew nts shown are a subset of those centained in item I.A.2.1.

"#eferences to Enclosures are to E4nton's letter of March 28, 1950 which is centained in the clarifi-cation of item .A.2.1 in NURIG-0737.

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Figure 2. Enclosure 2 from Denton's Letter TRAINING IN H[AT TRAN5fER. FLUID FLOW AND THERMODYNAMIC 5

1. Basic Proeerties of Fluids and Matter.

This section should cover a basic intrcduction to matter and its properties. This section should include suc't concepts as temperature measurements and ef f ects, density and its ef f ects, specific weignt, tucynacy. viscesity and et%er properties of fluids. A working knowledge of steam tables should also be incli,dec. Energy movement should be discussed including sucn fundamentals as heat exchange, speciftc heat laten. heat of vaporization and sensible heat.

2. Fluid Statics.

This section should cover the pressure temperature and volume effects on fluids. Exampit of these parametric changes shoJld be illustrated by the instructor and related calculations should be cerformed 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 components and systems should be discussed as applicable to the f acility with particular emphasis on saf ety significant f eatures. The characteristics of force and pressure, pressure in ligulds 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 Bernoulli's principle, energy in moving flutes, flom measure theory and devices and pressure losses due to friction and orificing.

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

l 4 Heat Transfer by Condaction. Convection and Radiation.

This section should cover the fundamentals of heat transfer by conductions. This section should include discussions on such concepts and ta: as as specific heat, heat flux and atomic action. Heat transfer characteristics of fuel rods and heat exchangers should be locluded 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 f acility. The convection current patterns created by expanding fluids in a confined area should be included in this section. Heat transport and fluid flow redsctions or stoppage should be discussed due to steam and/or noncondensible gas foreation during normal and accident conditions.

This section should cover the fundamentals of heat transfer by thermal radiation in the form of radisat energy. The electromagnetic energy emitted by a body as a result of its temperature shou ; ** t

. discussed and illustrated by the use of equations and sample calculations. Comparisons should t e m' e .

of a olack body atsorber and a white body emitter. ,

I 5. Cmame e' une . Boilir:.

This secticn smculd inctuce descriptions of the state of matter, their innerent characte 1stics and theratodynamic proDerties such as enthalpy and entropy. Calculations should be performed involving steam QJality afd void fractiCn pr0perties. The types of boiling should be discussed as applicable to the f acility dudng normal evolutiCns and accident Conditio95.

6. Bu r neut aad Flo. Instability.*

Inis section snou'Id Cover descriptions and mechanisms f or calculating such terms as critical flux, critical go.er. Dh8 ratio and het channel f actors. This sectierr should also include instructions for preventing and monitoring for clad or fuel damage and flow instabilities. Sample calculations should be illustrated by the in'tructor and calculations should be performed by the students and discJssed in the tra1Mng sessions. Methods and procedJets for using the plant computer to determine quaetitative values of various f actors during plant operation and plant heat balance de.erminations should also be covered in this secticn.

7. Reactor Neat Transfer Limits.

This section should include a discussion of heat transf er 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 f actors, radial and antal power distributions and changes of these f actors due to the influence of other variables such as moderator tercerature, nenon and control rod position.

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[ Figure 3. Enclosure 3 from Denton's Letter TRAIN!hG CRITERIA FDR MITIGAilh3 CDRE DAMAGE A. Incore instrumentation

1. Use of fixed or movable incere det'.4 tors to determine estent of core damage and geometry changes.

2. Use of thermocouples in determining peak temperatu*es; methods for extended range readings; methoos for direct readings at terminal junctions.

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

B. Escore Nuclear Instrumentation (NI$)

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

C. Vital Instrumentation

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

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

s. Determination of pressuriter level if all level transmitters fail,

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

c. Determination of other Reactor Coolant System parameters if the primary method of measurement p has f ailed.

D. Primery Cbeeistry

1. Espected chemistry results with severe core damage; conseqsences of transferring smal1 quantitles of Itquid ostside containment; im;:ortance of using less tight systems.

2. Esparted isotcpic creaadown for core damage; for clad data;e.

3. Corrosien ef fects of entended tru*ersion in p*imary water; time to failure, f' .

E. &tc*atic" W,aitering f 1. Restonse cf Prccess and Area Menitors to severe data;es; behavior of detectors = hen saturated; j method f or detecting radiation readings by direct measurement at detector output (overranged I detector); espected accuracy of detectors at different locations; use of detectors to determine i entent of core damage.

2. Math 0ds of determining dose rate inside Containment from Feasurements taken outside containment.

'. Gas Generatien

1. Metnods of H2 generation during an accident; other sources"of gas (Xe. Ke); techniques for venting or dispcsal of non condensibles.

2. H2 fla rattlity snd explosive limit; sources of D2 in containment or Reactor Ccolant System.

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a Figure 4. Control Manipulations Listed in Enclosure 4.

• 0NTROL MAft!PULAT!QNS

• 1. Pla .t c reactor startups to include a range that reactivity feedback from nucleae heat addition is r.cticeable and heatup rate is established.

2. Plant shutdown.

• 3. Manual control of steam generators and/or feed =ater during startup and shutdown.

4 Boretion and or dilusion during power operation.

• 5. Any significant (greater than 105) power changes in manual rod control or recirculation flow.

, 6. Any reactor power change of 105 or greater where load change is performed with load limit control or where fluz. temperature, or speed control is on manual (for HTGR).

• 7. Loss of coolant including:

1. significant PWR steam generator leaks

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2. inside and outside primary contairvren't

3. Ir.rge and small, including leak-rate determination 4 > turated Reactor Coolant response (PWR).

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

9. Loss of electrical power (and/or degraded power scurces;.

• g , Loss of core coolant flow / natural circulation.

11. Loss of concerser vacuum.

12. Loss cf se vice ater if re: wire: f or safety.

13. Loss of shutdown cooling.

14 Less of com:0re9% cooling system or cooling to an individual corponent.

15. Less cf nor al f eed.ater or normal feet.ater system f atture. ,

'15. Loss of all fee:.ater (ncreal a-c e~ergency).

17 Less of pectective system chaneel.

18. Misocsitionec control rod or rocs (or rec dreps).

! 19. Ina:111ty to drive centrol rocs.

20. Ccnciticas requiring use'of ereergency beration or standby liquid ccetrol system.

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21. Fuel cladding f ailure or high activity in reactor coolant or offgas.

22. Tu*btne or gene ator trip.

23. wa lfuncticn of automatic certrol system (s) which affect reactivity.

24 Malfunction of reactoe coolant pressure /vol m control system.

25. Reactor trto.

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

27. Nuclear instru-entation f ailure(s).

• starred iten to be performed annuatly, all others biennially.

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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 should consist 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 licensees generally have their own unique course out- ~

lines, adequacy of response to these requirements necessarily depends only

,on.w hether 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 (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 INPO documents are expected to satisfy all the requirements pertaining to training material which are addressed in this fgaluation.

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-s ure 1, I tem C.3. The latter requirement calls for all the manipulations listed in_ Enclosure 4 (Figure 4 in this report) to be performed, at the frequency indicated, unless they are specifically not applicabl.e to the licensee's type of reactor (s). Mme of these manipulations may be performed on a simulator. Personnel .witt, 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 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 present or available for instructing or assisting students; lectures, t seminars, discussions, problem-solving sessions, and examinations are considered contact periods. This definition is taken from Reference 4. -

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,. o B. II.B.4: Training for Mitigating Core Damage 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 the content of this training. " Plant Manager" is here taken to mean the highest 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 operator requalification program.

However, II.B.4 applies also to operations 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 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 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 commensurate with their responsibilities. Since this imposes no additional demands on the program itself, we do not address it in this evaluation. It would be appropriate for resident inspectors to verify that non-operating personnel receive the proper training.

sc-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 l

l The licensee (Baltimore Gas and Electric Company) has submitted to NRC a number of items (letters and various attachments) which explain their l

training and requalification programs. These submittals, made in response to Denton's letter, form the information base for this evaluation. For the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, there were three submittals with attachments, for a total of nine items, which are listed below. The last submittal and attachments thereto were in response to the NRC request for additional information (Refence 6).

1. Letter f rom A.E. Lundvall, Jr., Vice President, Supply, Baltimore Gas & Electric Co., to P.F.

Collins, Chief of Operator Licensing Branch, NRC.

June 18,1980. (1 pg, with enclosures: items 2, 3,

& 4). NRC Acc No: 8101070379. (re: Response to i NRC letter dated March 28,1980).

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2. " Heat Transfer, Fluid Flow & Thermodynamics" Undated. (1 pg, attached to item 1). (re: Course Outline)

3. " Mitigation of Core Damage". Undated. (1 pg, attached to item 1). (re: Course Outline).

4. " Reactor and Plant Transients". Undated.(1 pg, attached to item 1). (re: Course Outline).

5. Letter from A.E. Lundvall, Jr., Vice President, Supply, Baltimore Gas & Electric Co., to D.G.

Eisenhut, Director, Division of Licensing, NRC.

August 11, 1981. (2 pp) NRC Acc No: 8108180126.

(re: Response to NUREG-0737, Item II.B.4).

6. Letter from A.E. Lundvall, Jr., Vice President, Supply, Baltimore Gas & Electric Co., to R.A. -

Clark, Chief of Operating Reactors Branch #3, Division of Licensing, NRC. May 28, 1982. (1 pg, with enclosures: items 7, 8, & 9). NRC Acc No:

8206020511. (re: Response to NRC's RAI dated 04/26/82).

7. Enclosure (1). Untitled, undated.(2 pp, attached to item 6). (re: Responses to NRC's questions outlined in the RAI).

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8. " Training in Heat Transfer, Fluid Flow &

Thermodynamics", Enclosure (2). Undated. (4 pp, attached to item 6). (re: Course Outlines).

9. " Mitigating Core Damage", Enclosure (3). Undated.(9 ..

pp, attached to item 6). (re: - -

IV. EVALUATION SAI's evaluation of the training programs at Baltimore Gas and Electric Company's Calvert Cliffs Nuclear Power P.lant 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 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)

The basic requirements are that the training programs given to reactor operator and senior reactor operator candidates cover the subjects of heat transf er, fluid flow and thermodynamics at the level of detail specified in Enclosure 2 of Denton's letter.

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, ..s Submittal item 8 is the course outline for the instruction provided relating to heat transf er, fluid flow and thermodynamics. It consists of 80 classroom hours and covers the subjects in considerably more detail than specified in Enclosure 2 of Denton's letter, thereby satisfying the requirements of this Enclosure 1 item.

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 of this report).

Submittal item 9 is the course outline related to accident mitiga-tion training. It shows the subject covered in greater detail than is specified in Enclosure 3 of Denton's letter, provides 174 contact hours and therefore fulfills the requirements of this Enclosure 1 item.

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 7 indicates 64 contact hours are devoted to transient analysis and submittal items 4 and 9 present the course outlines for this subject. The three submittal items show the requirements of this Enclosure 1 item are met.

w 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. -

Submittal item i states: " Operations Instructors currently ' hold NRC licenses and are, theiefore, subject to the same requalification requirements as are other licensed operators. This requalification program assures that the instructors are cognizant of current operating history, I

problems and changes to procedures and administrative limitation". The foregoing indicates satisfaction of Enclosure 1, item A.2.e.

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.

Submittal items 2, 3, 4 and 7 present course outlines relating to

! heat transf er, fluid flow, thermodynamics and accident mitigation. The requalification program provides 142 contact training hours all of which is O

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v related to heat transfer, thermodynamics and mitigation of core damage.

These submittal items clearly indicate the requirements of this Enclosure 1 item are met.

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. Submittal item 7 shows the requirement of this Enclosure 1 item as company policy.

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). Submittal item 7 states conformance with this Enclosure 1 item.

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

• • - Submittal item 7 states that all licensed personnel have received training in mitigation of core damage (addressed in connection with Enclosure 1 items A.2.c(2) and C.1. Positions having received the training are as follows: Plant Superintendent alternates, General Supervisor -

Operators, Shift Supervisor, Senior Control Operator - also serves as STA, Shift Supervisor Assistant - also serves as STA, and Control Room Operator.

The Plant Superintendent (Plant Manager) has not received. core damage mitigation training under the Calvert Cliffs program but participated in an 8 - hour course on Core Damage Mitigation conducted by Combustion Engineering on November 11, 1981. Additionally, the Plant Superintendent, who previously held an SRO license, received training similar to that pro-vided by the Baltimore Gas and Electric Company as part of his license training program and as an officer in the Navy's Nuclear Training Program.

Therefore, all operations personnel (licensed and non-licensed) have received training in core damage mitigation.

V. CONCLUSIONS The Baltimore Gas and Electric Company training program for operator candidates provides a total of 288 contact training hours, 215 of which relate to heat transfer, thermodynamics and core degradation. Its licensed operator requalification program provides 142 contact training hours, all of which are related to heat transfer, thermodynamics and mitiga-tion of core damage.

SAI concludes that the Baltimore Gas and Electric Company has. met the requirements of ,TMI Action items 1.A.2.1 and II.B.4.

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= V. REFERENCES

1. "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 Commission. 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, 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 Assignment 2, December 23, 1981.

4. " Guidelines f or 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.

6. NRC request for additional information sent the Baltimore Gas and EgetricCompany, April 26, 1982.

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