ML20093G820

Forwards Responses to Listed Draft SER Outstanding Issues. Items Discussed During 840424 Meeting

ML20093G820
Person / Time
Site:	Beaver Valley
Issue date:	10/01/1984
From:	Woolever E DUQUESNE LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
2NRC-4-156, NUDOCS 8410160046
Download: ML20093G820 (103)

Text

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Duquesne Lidit 2g,g ge 2

Nuclear Construction Division geog 2 Robmson Plaza Building 2 Suite 210 Pittsburgh, PA 15205 October 1, 1984 United States Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Mr. George W. Knighton, Chief Licensing Branch 3 Office of Nuclear Reactor Regulation

SUBJECT:

Beaver Valley Power Station - Unit No. 2 Docket No. 50-412 Outstanding Issue / Question Response Gentlemen:

This letter forward s responses to the outstanding issues listed below. These items were discussed with the reviewer during a meeting which began April 24, 1984.

Attachment 1: Re sponse to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.l(1).

Attachment 2: Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.l(5).

Attachment 3: Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.l(5).

Attachment 4: Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Sect ion 13.2.1.2.

Attachment 5: Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.3.

Attachment 6: Res ponse to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.3(3).

Attachment 7: Re sponse to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.3(4)(b).

Attachment 8: Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.3(5).

Attachment 9: Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.4 (I.A.2.1).

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.Unitid Stctas Nuciscr Rigulctory Conunicsien Mr. -G23rgs W. Knighton, Chief

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Attachment 10: Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.1.4 (I.A.2.3).

'Attechment 11: Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.2 (STA).

Attachment 12: Response to Outstanding Issue of the -Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Section 13.2.2 (Fire Protection).

DUQUESNE LIGHT COMPANY By O. -

E.l/J . Woolever Vice President GLB/wjs Attachments cc: Mr. H. R. Denton, Director NRR_(w/a)-

Mr. D. Eisenhut, Director Division of Licensing (w/a)

Mr. G. Walton, NRC Resident Inspector (w/a)

Mr. E. A. Licitra, Project Manager (w/a)

Ms. M. Ley, Project Manat,er (w/a)

COMMONWEALTH OF PENNSYLVANIA )

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COUNTY OF ALLEGHENY )

On this ch day of _c , /ff , before me, a Notary Public in and for said Commonwealth and County, personally appeared E. J. Woolever, who being duly sworn, deposed and said that (1) he is Vice President of Duquesne Light, (2) he is duly authorized to execute and file the foregoing Submittal on behalf of said Company, and (3) the statements set forth in the Submittal are true and correct to the best of his knowledge.

s&Pa . Notary Public ANITA ELAINE REITER. NOTARY PUBUC ROBINSON TOWNSHIP. ALLEGHENY COUNTY MY COMMISSION EXPIRES OCTOBER 20,1986

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ATTACHMENT 1 Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report' Draft SER Section 13.2.1.1(1): Initial Training Program (excerpt)

Phase 1 - Academic and Nuclear Fundamental Training This training course of formal classroom study will be approximately 14 weeks long; it is designed to provide individuals with basic knowledge in science and technology of power plant operations. The major areas to be covered are mathematics, basic nuclear physics, reactor principles, radiological fundamentals, chemistry, instrumentation and control, electrical theory, safety analysis,- fluid flow, thermodynamics, and heat transfer.

With respect to instructions in the topics of fluid flow, the rmodynamics and heat transfer, the staf f requires the applicant to provide a program in accordance with the guidelines as outlined in Enclosure 2 of H. R.

Denton's March 28, 1980, letter. The staff will review the program when it is docketed and report its findings in the final SER.

Response

The current lesson plan LP-TMO-O (attached), " Thermodynamics -- Intro-duction" provides an outline of subjects shich satisfy the topics of fluid flow, theromodynamics and heat transfer as outlined in Enclosure 2 of the Denton letter. This course has been evaluated by the American Council on Education and has been recommended for upper division bacca-laureate category, three semester hours in Nuclear Technology.

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Figure 0.7 DUQUESEE LIGHT COMPANY Nuclear Division (

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LESSON PLAN 183 Thermodynamics - Introduction Course Course Hours .

May 7, 1982

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Slavic tak / Roehlich I Data Instructor

[ jg LP-TMO-O (1 hr.)

Lesson Plan No. (Sequentially Fron U J

g Approved By:

Raferences To Be Quoted: INPO Standards Itsas Issued: (Actsch copy of all passouts, quizzas, etc.)

1) Text: BVPS Thermodynamics: 2) Course. Letter: 3) INPO Standards:

4) Course Schedule: 5) Steam Tables: 6) Lesson Plan Handouts

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Introduction:

1.

Purpose:

To delineate the obiectives. content. and schedule for che BVPs Thermnd mom 4ce Course.

2. Motivation: (Discuss how you plan to motivate students)

Explain that a lack of knowledge of Thermodynamics can lead to serious safetv problems, e.g., TMI; also a significant fraction of the NRC SRO and RO Licensing Exam covers Thermodynamics.

3. General Outline: (T.ist detailed' outline Section I) eni,re. e rh o a r,1.

Courne obieceiven. entirne enneane.

4. . .Getieral Student Go'els: (List detailed student objectivesSection II)

Upon completion of this lesson, the student will be aware of: the course objectives, how the course will be conducted, and the course schedule.

- 0.Ik ISSUE.3~

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DUQUISNE LIGHT COMPANY

/' Nuclear Division - -

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Nuclear Support Services Department _,

APPROVAL SHEET - LESSON PLAN AND TEXT REVISIONS Document

Title:

LP-TMO-O R v. Subjects Revised Revised Approval No. (Brief Description) by Signature Date 1 - Revised detailed outline to S. Slavichak July 19, 1982 l

- Course description d v

- Absence policy

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- Added student objectives

- Added INPO Standards to handouts _ _ . . . . _ _

g 2 - Added revision approval form S. Slavichak June 15, 1982

- Changed detailed outline format

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- Revised detailed outline to include more detailed:

- Performance requirements

- Absence policy

- Changed Student Objectives format [l to terminal objectives and - g .

enabling objectives.

- Added course objectives, descrip-tion, and temporal breakdown to course letter.

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- Changed course schedule to lengther time spent on lessons ,6 & 7 while shortening time spent on lessons 1 & 5.

- Abbreviated Thermodynamics Formulas ,

conversions and constants handouts,

< - Added exam policy statement to ku' handout. ,

- Modified transparency numbering ---

system. _ _ , _ _ _ _ , _ . , ,,_

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I Instructor's Lesson Plan LP-TMO-O Pne. 1 of 5 Lesson Plan Outline Instructor Notes and References

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I._ Issue of Materials ) Students may keep;.tell them to

( read the course letter now.

A. Course letter B. Text C. INPO Standards D. Course Schedule E. Lesson plan handout J F. Steam Tables -- Students must return; record copy number II. Introduction of Instructor (s)

A. Name(s)

B. Office location (s)

} C. Background (s) (if asked)

III. Introduction to Course -- Preview lesson objectives

(' A. Scope of the course

1. Meet or exceed INPO standards Refer to handout; show TP-TMO-0-1, Course Letter

a. Technical specifications also are 1 earned

2. Help students understand heat transfer and fluid flow in plant systems

a. During normal operations

b. During emergency conditions Prepare students for NRC exams 3.

a. Both RO and SRO exams contain thermodynamics problems i

B. Course objectives (upon completion of this -- Show TP-TMO-0-1, Course Letter course)

1. Students should be able.to describe fluid flow and heat transfer processes in the plant i

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Instructor's Lesson Plan Page 2 of 5 Lesson Plan Outline Instructor Notes'and References

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2. Students should be able to describe the heat and energy cycles involved with plant operations

3. Stu' dents should be able to explain the reactor thermal and hydraulic limits

_C. Course content: Thermodynamics, Heat Transfer - These are the three major topics and Fluid Flow covered by this course

1. Chapter 1 - Fundamentals

a. Units and conversions

b. Properties of matter

c. Pressure / vacuum scales

d. Forms of energy

2. Chapter 2 - Heat and the First Law of Thermodynamics

, a. Heat (m.[-

! b. First law

c. Heat tranfer (1) Radiation (2) Conduction

3. Chapter 3 - Convection i a. Convection

b. Fluid flow

c. Heat exchangers

4. Chapter 4 - Systems, Pumps and Valves

a. Systems

b. General energy equation

c. Bernoulli's equation

/ d. Flow measuring devices

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Instructor's Lesson Plan Page 3 of'5

-LP-TMO-O Lesson Plan Outline Instructor Notes and References-(^

e. Pumps -

f. Pump laws and curves

g. Pipes and valves

h. Integrated fluid system behavior

5. Chapter 5 - Behavior of Steam and Gases

a. Entropy

b. Steam tables

c. Processes.

d. Moisture separators

e. Ideal and real gases

f. Steam / air mixtures

6. Chapter 6 - The Conversion of Heat to Work: The Stesm/ Water Cycle k.

a. Nozzles

b. Air ejectors i c. Turbines (1) Impulse (2) Reaction (3) Efficiency

d. Condensers

e. Cycles

f. Cycle efficiency

g. Calorimetric

7. Chapter 7 - Nuclear Power Plant Charact-eristics

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a. Program Tavg

b. Pressurizer

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Instructor's Lassen Plen Page 4 of 5 LP-TMO-O Lesson Plan Outline Instructor' Notes'and References

c. Thermal sleeves i d. Level indication .

e. Core thermal limits

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f. Boiling heat transfer

g. Core peaking factors 5

h. Technical specifications

1. ' Natural circulation D. Conduct of the Course

1. The course is broken up into' lessons which correspond to the chapters of the text

2. The lessons vary from 1 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in i length

3. Lessons are presented as a lecture O a. Prior to the lecture (s) on a lesson,  ;

the student will be issued lesson  !

objectives and given a text reading assignment l

b. Each lecture is approximately one hour long followed by a ten

. minute break 4

c. During the lecture (s), the student should take notes i .

4 Subsequent to the lecture (s), the student will have text problems to complete

5. Prior to each exam, the text problems will be reviewed by the instructor E. Exams

1. . Total of six (6) exams during course 4

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Instructer'o Lancen Plcn Pa83 5 of 5 LP-TMO-O Lesson Plan Outline Instructor Notes and References

a. Exam 1 covers Chapters 1-3, Exams 2 through 5 covers Chapters 4 through 7, respectively; Exam 6 is a comprehensive final exam

b. Exam weighting (1) Exam 1 through 5 - 12% each (2) Exam 6.- 40%

c. Exam conduct -- Refer to handout on the Conduct of Training Dept.

d. Exam content Exams (1) Definitions (2) Essays (3) Short answers F. Performance

1. Failure of the course will result in an Academic Warning (< 70%)

2. Failing any quiz o.r test (< 70%) cy;

< 72%) the course or

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marginallypassing(ltinaReportof final exam will resu Counseling.

G. Absence

1. Students will have to makeup for lost --

Stress that it is the students time respr.asibility to meet with his instructor on the day he returns

2. Catch-up time will be on a one for to work. Together they will one basis (no overtime!) arrive at a schedule for completion of the missed work.

a. e.g., A student who missed four (4) days of class will have four (4) days after his return to work to make-up all he missed. Concurrently' he must learn the new material

, taught during this make-up period H. Course schedule

1. Briefly review course schedule with the -- Refer to handout; show TP-TM0-0-2, students Course Schedule

2. Emphasize that this schedule is only tentative 5.

IV.. Summary A. Review Objectives --

Review lesson objectives with B. Make problem assignment C. Make reading assignment

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LP-TMO-O STUDENT OBJECTIVES Terminal Objectives Upon completion of this lesson, the student will be aware of the course objectives, the conduct of the course, and the course schedule.

Enabling Objectives

1. The student will be able to list the course objectives.

2. The student will be able to list the three major topics covered by the courae.

3. The student will be able to describe the format of the course.

4. The student will know how absences will be resolved.

5. The student will know the number and frequency of exams given during the course and each one's percentage of the final grade.

6. The student will know the consequences of exam or course failure or near failure.

( 7. The student will be able to interpret the course schedule.

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ATTACHMENT 2 h

Respon'e s to Outstanding Issue-of the

. Beaver Valley Power' Station _ Unit No. 2

' Draft' Safety Evaluation Report Draft:SER Section'13.2.1.1(5): Initial Training Program (excerpt)-

Phase 5 --Plant Manipulations Training This phase ~ of the training program ' is approximately 13 weeks long; and

- will provide license -candidates with hands-on training in the areas - of reactivity manipulations. The- applicant has indicated -thatthis-training-will~ be conducted ca either -cae'of the-Beaver Valley units, the Beaver Valley simulatorf or~an;_offsite simula tor. _However, the appli-cant has _not provided :the Laimulator training program ~ for staff review.

As 'specified in Enclosure ~ 1 of H. R. Denton's letter of March 28,.1980, the. staf f -requires all license candidates to participate in a simulator training program as part of the long-range- training program. Therefore,

-the _ staff requires that the applicant . submit- a detailed simulator training program' for NRC review. The staf f will report the results of-its review in the final SER.

Response

Attached is a description of the reactor operator startup certification course for. experienced hot licensed candidates. This course is being used - for operators now being trained for BVPS-1. Enclosure 4 of the Denton letter does not specify the topics to be covered in tne initial operator training simulator course, however, -it does describe the requirements for requalification training. Attached is a description of the simulator retraining course presently used, Aich meets the require-ments of Enclosure '4 of the Denton letter except all items are performed on -a' two year cycle due to the limited amount of simulator ~ time available in the industry.

Individuals to license on Unit 2 will be- either of- two categories, experienced licensed operators from Unit 1 or individuals completing the initial;1icense training program.

Beaver Valley i4 currently constructing a plant simulator which is planned to be available for training prior to any individuals being licensed'on Unit 2. In any case, all candidates being examined for an operating license _ on Unit 2 will meet the requirements of the Denton lette either by being experienced on BVPS-1 and completing both simula-tor ' programs described' in Paragraph 1 or by completing the license

- simulator' training progran as desc ribed in the Beaver Valley Simulator Training Plan Section III (attache'd).

FSAR 13.2.1.1 will be revised to clarify Phase 4 and Phase 5 of the licensed operator training program as shown on the attached Page 13.2-2.

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- Rea::cr Litotus Certification Limalater Course for Experier.:e:

15 :*' 35 ' nct Licensec Candidates Puroose This course is'specifically designed for hot license candidates having significant control room experience.. By means of simulator

-training, the license candidate is exposed to a variety of conditions and transients which might not be experienced during actual operating conditions.

In order to be eligible for the NRC license exam, the hot license candidate must have achieved two criticalities during his/her training.

Also, the candidate is required to take the plant reactor to critical during the NRC test. In consideration of these requirements, WNTC offers this course in order to provide the simulator operational experience as stated above. This program is specifically designed to give the hot license candidate a broad spectrum of control room operations, ranging from cold solid shutdown to plant malfunctions in the power range.

Also, each trainee will perform three simulator reactor startups throughout the duration of the program.

The final day of the program consists of a startup certification examination performed on the simulator. The NRC will waive its requirements for two training startups of tne plant reactor as well as the startup during the actual licer. sing exam it a student attends this .

course and passes the startup certification examination (Nu. Reg. 0094, App. F).

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~inis course lasts a C. ration of seven cays. Soecifically, the coarse consists of 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> ir, t;tual.three man operation cf the simulator anc 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> in group ciscusilons aimed at preparing for'tne next day's evolutions. Utilities are encouraged to have the students use their o n procedures and technical specifications wnere applicaole, especially. in the areas of reactor startuo, ECP and 1/M calculations.

The initial day of the course consists of control board familiarization and basic system operation with a substantial number of

. demonstrations by the instructor. Subsequent days allow the student to bring the simulator to criticality and to perform a wide range of operations. The final day is utilized for the Startup Certification Examination.

Objective The student shall demonstrate upon the simulator a knowledge level and operational proficiency adequate to pass the Startup Certification '

Exam.

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'; DAY 1 DAY 2 DAV 3 DAY 4 DAY 5 DAY 6 DAY 7

-Courso Introduction -Review Previous -Review Previous -Review Previous -Review Previous -Review Previous -Review Previous

-Tech f. blodes Operations Operations Operations Operations Operations Operations

-Systems Review -Review Rx Startup -Discuss Secondary -DAscuss and Corr;ute -Xenon Effects on Rx -Reactivity T.ffects

-Rx Startup RCS Proceduro Plant Starttap and W .

Startup cn Rx Startup

-Rx Startup Tech Specs CVCS -P wer Reduction and -thergency Boration -Written Examination

-Lxplain and Cornputo . Power Incresso Paant Shutdown -Rx 1 rip

-h Startup Forms ICP SC815 an I.CC -Constant Axial Offset (2 ilours)'

-Coo 1Jown Frors llot -Sci f . Study

-Plant lleatup From Doubling Effect Progrms Starx!by to Cold Cold h tdown to SGI

-Automatic Rod Control hth llot Standby 1At Plot System -Steam 11 :p System

-Scif Study -Review NIS e

-n neory -Dropped Rod Recwcry Subcritical utltipli-cation Rx Criticality Doppler Effect Igortance Factor Point of Mding flest Rx Trips Associated

, with Ihr Srnrtun

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InUI LtN01 ' 18 0! IROI - IROI Inug mg

-Control Room Tour -Conduct Plant IIcatup -Conduct threc Rx -Rx Startup/ Secondary .-Reduction in Plant -Conduct h ree Rx -Individual Rx from Cold h tdown to Plant Startup with Power with Rx h tdou Startups with Simlator llot StandayConditions Startups Pou r Increase startup Certification

-Cooldown to H5'F Hixin a Xo transient Crcquter Room .g. sec d ry (ICTI T" "" IOI II E -Dropped Rod and g conditions Examinations system (Time permitting during the First Instructor Booth Recovery tarty

-Rx Startup to 21 Pwr Rx Startup)

-Rx Startup to 21 Pwr (Time Permitting) 9

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I - ..-- WESTINGHOUSE fiUCLEAR TRAltiltiG SERVICES SIM 415 SIMULATOR RETRAltiltiG COURSE l

(5 DAY.0PTI0ti)

ItiTR000CTORY STATEMEtiT '

A. OVERVIEW ,

The Westinghouse Simulator -Retraining Course (Sim 415) is designed to refresh the licensed operator's knowledge and proficiency. Tnrough a varied level of sim'ulator evolutions, the reactor operator or senior reactor. operator can respond to transients and malfunctions not normally encountered during actual plant operations.

3. PURPOSE This coarse has ceen designed to satisfy all the current annual and bi-annual control manipulations required by the tiRC.

C. PREREQUISITES f_ Participation in tnis course snali ce limited oy the following i prerequisites:

1. The student should nold a current Operator or Senior I Operator License, or

2. The student should have satisfactorily complete a license certification program, or j 3. The student should show enrollment in a retraining program designed for renewal of an expired license, or

'4. The stuaent snall De selected by nis training department for enrollment into this program.

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/ D. COURSE ORGAf112ATION *')

Tnis course is coa. prised . of - five units. - Eacn unit represents a i~

' combination ~ of simulator sessions supplemented by' classroom seminars and' critiques. Each unit .is outlined in separate!

assignment sheets (attached) containing an introduction and -

specific assignment. Thus, eacn unit specifies tne- stucent

4 i

objectives, reading assignments, ~ course presentations, and

[ requirec simulator operations..

o E. COURSE OBJECTIVES

{

Terminal Objective:

With the aid: of -a simulator, tne student shall demonstrate an ai. 'lity. to identify, describe, analyze, and respond to a variety

!- of transients and. malfunctions witn a level of proficiency equal to or exceeding regulatory and safety standards.

I Enablino Objectives:

Upon ccmpletion of this course, the student snail be able to:

DESCRIBE the plant response and required operator action for a large loss of coolant accicent.

DESCRIBE orally the plant response ano required operator action for a large steam generator tune rupture.

DETERMINE that adequate core cooling exists.

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DESCRIBE orally the . plant response and required operator action for a major loss of secondary coolant.

DIAGNOSE anu SOLVE operational problems associated with tne [

failure of plant protection and control systems. l

- DIAGNOSE and SOLVE operational problems associated with the loss of power sources or buses.

- _DIAufiOSE and SOLVE operational proolems associated with tne malfunctioning of automatic control systems effecting

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reactivity.

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b F. CONTENT AND S' COPE-

' Unit ' Title 1 . Reactivity Manipulations 2 Accident Assessment / Minor Plant-Transients (Part 1)-

3 Accident-Assessment / Mince Plant Transients

.(Part 2) 4- Major Plant Transients 5 Major Plant' Transients / Demonstrations

- G. BASIS OF EVALUATION At the end of each c0urse, the instructor is7 required to write up a .

formal evaluation on each student. This evaluation is then reviewed

, by the Training Center management and then forwarded to the student's training supervisor.

The following is a list of areas considered = by the instructor in 1 m3 king his evaluation.

- Class pcrticipation Individual knowledge of plant systems, controls, and operating procedures / limitation Use of reference materials Leadership (senior license personnel only)

. Control room operations Communicaticn 0052C IOI y e -w y -...w- g.,-9 -..p ---s, y .-

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SIM415 d5 DAY 2 DAY 3 DAY 4 OAY 5 -.

DAY 1 Classroo:n Review Classico.s RevicW. Classrcom Review Intru to Program (I fir) Classrous Revicw Classroc: Review (1llour) (1lledr) (I Hour). (1 lloaar)

R4 Startup 1001 Power; Eq. Xenon 1001 Power; Eq. Xenon 30Z Power; Xenon t,ullding 100% lwer; Eq. Xenon syncl.ronize to Grid Major Accident Di. gnosis Hajor Aci.ident Diagnosis Continue Plant Startup Nifunctions increase Power to 20t 100% Power; Eq. Xenon 100% Power; Eq. Xenon Hal functions . Inst. Air System Shift to Auto Systems Major Accident Diagnosis Major Accideat Diagnosis liod Control System' . Chemistry Turbine Halfunction 1001 Power; Eq. Xenon Plant Stabilization CVCS Rod Control System Plant Shutdown Major Accident Diagnosis 50% Power; Xenon t,ullding Component Cooling [IlC System Plant N1 function Diag. Service Water Hakeup Control .Syst' a :

Inst Air System Elect. System: .

l'aln Aux feed Systue. Aux, feed Systcm Plant Cooldoun (3liaurs) (3 Hours) (3 Hours)

LUNCH (1/2llour)

Rt Startup Power increase Curing Xenon Plant S/U from 10-8 amps Plant Startup from SOE with' Startup During Xenon Trans.

Synchronize to Grid Transient (50 - 100%) Hal functions Transients Halfunctions.

Increase Power to 20% Halfunctions Main feed System Halfunctions Pressurihr iysten Siiif t to Auto Systems Ciectrical System Pressurizer Systen 1;od Control Main feed System Pressurizer Half. Comp. Cooling System Condenser & Off-Cas Boric Acid System RCS Plant Shutdown Main Steam Syston Service Water System 5/G System Steam Generators' CVCS Rod Control /RPI FH3 System l'.aln Condenser Rod Control System Main Generator Electrical System 100% Power; Eq. Xenon Nuclear Inst System Rx Coolant Systua Protection System Charging System Loss of Coolant Plant Shutdown Main reed System Rad. Monitors (3 liours) (3 Hours) (3 lioJrs) (3Iours) (3Ifours)'

Critique (1/2 Itour) Critique (1/2Ilour) Critique (1/2 Hour) Critique (1/2 Hour) Critique (1/2Ilour)

P III. LICENSE TRAINING PROGRAM A. Course Descriotion The License Training Program is'a systematic train-ing program consisting of six (6) sections totaling-twenty-five (25) days (150 simulator hours).

o Sections 1-3 are directed to programs which allow the students to operate the individual control sys-tems of the plant in manual and automatic and observe control system functions and interrelations utilizing exercise guides and demonstrations.

i Section 1 is 2 days (12 simulator hours) in duration

- and will be conducted at'or near the end of the secondary systems qualification lectures.

Section 2 is 2 days (12 simulator hours) in duration and will be conducted at or near the end of the pri-mary systems qualifications lectures.

- - Section 3 is 2 days (12 simulator hours) in duration and will be conducted at or near the end of the reactor protection and control systems qualifica-tions lectures.

Detailed discussions of the systems covered in each of these modules will be accomplished through the system qualification lectures conducted as part of the classroom phase of the license training program.

Classroom lectures during the simulator phase will be in support of the simulator activities for that day.

III - 1 NUS CCAPCAATICN

'4[

o Section 4 is 4 days (24 simulator hours) in duration and will be conducted following Section 3. This section will be utilized to prepare the student for-and administer the startup certification portion of

. the license operator examination. Practice start-ups will be conducted from various initial condi-tions of.burnup and xenon, with emphasis placed on core behavior, plant control and operation and interactions.of applicable control and instrumenta-tion systems. A startup examination will be admin-istered to each candidate at the end of the section.

Classroom lectures will emphasii:e ' applicable opera-tions and administrative procedures including-tech-nical specifications and limits and precautions.

Reviews of related areas of reactor theory, kine-tics,. and control and instrumentation systems may be conducted if deemed necessary.

o Section 5 is 12 days (72simulat$rhours) in dura-tion and will be conducted following the license review ~ training portion of the qualification pro-gram.

Days 1-8 will be directed to combinations of normal and abnormal operating conditions. This program provides the candidate with various casualties and emergencies that could occur in the operating plant.

The candidate must demonstrate the ability to recover the plant from various conditions utilizing approved procedures.

Days 9-11 will be directed to accident mitigation.

This portion of the program will provide the simula-tor support for the Beaver Valley Mitigating Core Damage Course taught by the Beaver Valley Classroom III - 2 NUS CCAACAATICN

+

Training Staff. During these three (3) days, the candidates will'be given the opportunity to look at .

and respond to particular emergency situations that could occur in the plant and result in eventual core damage.

- Day 12 will provide for operational audit exams.. ,

The classroom instruction for Section 5 is intended to support the activities occurring on the simulator floor for that particular day.

o Section 6 is reserved for NRC license examinations.

o An average daily schedule would consist of the following:

Classroom Instruction / Discussion - 2 Hours Simulator Operation -

6 Hours No Scheduled Lunch Break B. Training Obiectives

1. To provide and document the training required

.. o for a candidate to systematically acquire the basic and specific operating knowledge neces-sary to safely and effectively operate the Beaver Valley Nuclear Power Station Units 1 and

- 2 as a Reactor Operator.

2. To provide and document the training required

- for a candidate to systematically demonstrate the basic and specific operating skills neces-sary to safety and effectively operate the l Beaver Valley Nuclear Power Station Units 1 and

- 2 as a Reactor Operator.

III - 3 NUS CCAPCAATCN

7

~

C. Tyce-of Training

1. Classroom Instruction-(Sections 1-3) o 2 Hours Per. day

- o 2 Days Per Section

2. Classroom Instruction (Section 4)

~

o 2 Hours Per Day

'~

o 4 Days

3. Classroom Instruction (Section 5) o 2 Hours Per Day o 12 Days

4. Simulator Training

a. Systems o 6 Hours Per Day o 2 Days Per Section

"~

6. Startup Certification o 6 Hours Per Day o 4 Days

c. Operations and Accident Mitigation o 6 Hours Per Day o 12 Days III - 4 NUS CO A AC A ATICN

D. Curriculum

1. The Basic Curriculum for Sections 1-3 is as follows:

o Main Steam o Condensate ,

o Extraction Steam

, o Heater Drains o Feedwater o Main Generator and Transformer

o. Main Turbine and Condenser o 4 KV Ststion Service Transformer o Reactor Coolant o Chemical and volume Control o Boron Recovery o Residual Heat Removal o Safety Injection o Containment Depressurization o Liquid Waste o Gaseous Waste

o Area Ventilation

.. o Reactor Control and Protection o Reactor Excore Instrumentation o Incore Instrumentation o Plant Process Control .

o Main Computer

2. The Basic Curriculum for Sections 4 and 5 is as follows:

i o Suberitical Multiplication o Reactivity Coefficients o Reactivity Balance Procedure o Station Startup o Station Shutdown III - 5 NUS COPACAAT:CN

o Power Operations Procedure o - Technical Specifications o Limits and Precautions o Accident Mitigation

~

o License Events Reports o Operating Procedures o Emergency and Abnormal Procedures .

o ECCS Actuation o Loss of Reactor Coolant o S/G Tube Rupture o Total Loss of Feedwater o Reactor Trip o Turbine and Generator Trip o Station Blackout o Loss of Component Cooling o High Reactor Coolant Activity o High Activity - Radiation Monitoring o Loss of Instrument Air o Loss of Containment Vacuum o Loss of Reactor Plant River Water

.- o Loss of Reactor Coolant Flow o Loss of RHRS E. Instructional Resources

1. Resources o NUS Thermal Science Course o Beaver Valley's Westinghouse NSSS Docu-ments o Beaver valley's Administrative Procedures o Code of Federal Regulations o Beaver valley's Emergency Plan o Beaver Valley's Health Physics Manual o Beaver Valley's One-Line Diagrams o Beaver Valley's Flow Diagrams o Beaver Valley's Limits and Precautions III - 6 NUS CCAPC AATICN

.- _ , ,_ . _ . _ _ _ _,_ . _ - . - -_ - - , . . . _ - _ - _ _ _ . . ~ _ -- _ . _ _ _ _ .

. l 1

a o Beaver Valley's FSAR o Beaver valley's Technical Specifications '

o Beaver Valley's Alarm Response Manual o Beaver Valley's Mitigating Core Damage Program o License Event Reports o I and E Bulletins .

o ANSI Standards o NRC Regulations o Beasar valley's System Descriptions o NUS PWR Core Physics Course o NUS Strength of Materials Course

- o Beaver Valley's Procedures F. Schedule o The Licen.se Training Schedule is as follows:

e 9

-e 9

III - 7 NUS CCAPCAATION

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g, Procedure Ibview &

A Discussion in Support g 3 of Sinalator Evolu- t 3 tions l R o

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1. min Steam and 1. Main 'Ibrbine and Rain Steam Isola- Auxiliaries tion 2. DiC Systan

2. Reheater Control 3. Main Generator 3 and Voltage Regu-y Syston g 3. Ileater Drains lator System Systun 4. Iturgency Diesel i U y, 4. Ccalenser Steam Generators l

3 Disup 5. Electrical Distri-

5. Peodwater and bution System

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1. Pressurizer Control 1. Residual Heat Re-Systens moval Systan j 2. Reactor Coolant 2. Safety Injection g limps and Seals Systan

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Procalure Revicw and A Discussion in Support '

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1. Iluclear Instrunen- 1. Protection and l tation Control Systans i 2. Incore Instrunen- 2. Ibd Control Systen 3 tation I

3. Radiation Monitor-

M 1"3 u 4. reactor vessel g, Ievel Indication A 5. Core Cooling Fbn-

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1. Reactor Startups 1. Reactor Startups Startup Certifica-

1. Reactor Startups i

(let standby to (hot standby to 5% (hot standby to 15% tion Examinations

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10 anps) Repeat tine permits, tine pennits, y as t2ne permits. 2. Reactor Trip M lboovery.

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I. Prrywbre Review arvi

!A Discussions in Support of Sinulator i g

S S Evolutions 3 11 O

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1. Reactor Startup and 1. Operations at Power 1. Operations at 1. Operation at Power

l. Plant Startup (cold standby to Power Increase 2. Feeduater mlfunc- Power 2. Pressurizer Control hot stardby). 2. Boron Concentration tions 2. Reactor Plant Coolin, i Systons mlfunc-Qanges 3. Condensate m 1func- mter Halfunctions tions 3 1. Onnical arxl voltme 3. Ini control Systan

3. Iluclear Instrumen- tions I

tation Halfunctions 4. 'Ibrb;ne Plant Control Syston ml- mlftux:tions g functions

4. It actor Trip Coolirv3 m ter u Systan Malfunctidna i 1.

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O Accident Mitigation I. Procedure Review A ard Discussion in '

1 Support of Sinulator

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1. Operation at 1. Operations at 1. Shutdown and Cool- 1. Operations at Ibwer 1. IOCA Power down 2. A'lW3 2. Steam Generator I'tmer

2. 2. Ioss of Sluit&wn 3. Dnergency Doration 'Ibbe Rapture

2. ICP Malfunctions Electrical Systan

4. Blackout 3. Stall Dreak IOCA S 3. Instnsnent Malfunctions Cooling I Failures 3. Snall BCS Icaks 3. RilR Systan M 4. Capressed Air 4. S/G 'Ibbe Icak Malfunctions i

U Systan Halfunc- 4. Feedwater Ralfune-tions I. tions A

5. NIS R11 functions 1

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L Accident Mitigation Audit Examinations A

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Feedwater i 2. Overcooling

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G. Training Materials

1. Classroom o Overhead Projector o Transparencies o Chalkboard ,

o Chalk o Program Schedule o Lesson Plans ,

o Student Handouts

2. Simulator

.. o Site Specific Simulated Control Room o Simulator Drill Guides H. Prerequisites

1. Prerequisites

- o- The candidate must meet the extensive operating experience requirement estab-lished by ANSI 3.1 (1981).

I. Performance Criteria To successfully complete the simulator portion of 1 the License Training Program, the candidate must meet the following conditions:

o The operational / oral audit exam results reflect a satisfactory level of competence.- )

(Due to the subjectivity of this type of evalu-ation, the documentation required to determine a candidate's " level of competence" will be l

III - 8 NUS CCAACAATICN  !

d reviewed by both NUS Corporation and Duquesne Light Company, and will also be determined on '

an individual basis).-

J. Evaluation Procedure

- 1. Classroom .

o Written quizzes and examinations will be periodically administered by the Beaver Valley Training Department as_part of the overall license training program.

o A comprehensive written audit exam utilizing the NRC format will be adminis-tered by a knowledgeable, independent audit team at the completion of the pro-gram prior to the NRC exam.

2. Simulator c.

o Reactor startup examinations will be administered by the instructor staff. A minimum of two startups will be performed by each candidate prior to the examina-tion.

o oral exams will be administered, by the instructor staff, periodically throughout the program. Each candidate will receive at least two oral examinations.

o Daily student evaluation sheets will be )

filled out by the instructor to document the daily surveillance of each student.

l III - 9 NLJS CCAPCAATICN l l

l

'* g ..

o Overall student evaluation sheets will be filled out at the completion of the week ,

(or section if less than 1 week) .

o A comprehensive- operating audit exam utilizing the NRC format will be admints-tered by a knowledgeable independent ,

audit team at the completion of the pro-gram, prior to the NRC exam.

3. Test Evaluation o The oral quizzes, reactor "* rtup examin-ation and daily operation.. evaluations results will be graded as "satisf actory" or " unsatisfactory." These qui:=es and exams will contribute to the overall stu-dent evaluation.

o The written 'd operational audit exams will parallel the evaluation process

'- adopted by NRC.

K. Documentation, Records, and Forms

1. Documentation o At the completion of each week or section (if less than 1 week), student evaluation sheets will be filled out. These evalua-tions will include the result of:

o Oral Quiz Results o Daily Student Evaluation Sheets o Reactor Startup Examination Results I

III - 10 NUS CCAAQAATICN

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e o Daily attendance records will be kept and filed with the weekly schedule.

o Simulated control room reactor operators log book must be kept to document reactor startup and major evolutions occurring in the control rcom.

o The comprehensive written and operating audit exam results will be filed into the class files and the individual student files.

2. Records The following documents will be maintained in the class and/or individual personnel file as permanent records for the time requirement established by NRC Regulatory Guides:

.,,. o Attendance Records o Daily Evaluation o Weekly (Section) Evaluation o Oral Examinations o Reactor Startup Exam Results o Program Schedule o Audit Exam Results, Test and Answer Key o Reactor Opertors Log (simulated) o Simulator Evolution Summary Sheets l 3. Forms o The following are forms to be used for the License Training Program:  ;

o Daily Attendance I

o Daily Student Evaluation 1 III - 11 NUS CCAPC AATICN

~

o Weekly (Section) Overall Trainee Evaluation .

o Oral Examinations o Reactor Startup Exam o Simulator Evolution Summary o The actual forms to be used will be in- .

cluded in the simulator Facility Instruc-tions.

1 4

4 1

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i 8

III - 12 NUS CCAPCAAT CN L ..

BVPS-2 FSAR L

-i, f

Individuals in the licensed operator _ training program receive .m -

training commensurate with their previous education, training and experience. All operating personnel required to hold a license, V ^

according to regulatory requirements stated in 10 CFR 55 such as Reactor. Operators (RO) and Senior Reactor Operators (SRO), are  ;

provided the necessary training in order to qualify. ,

4 13.2.1.1 Liciensed Operator Training Program The normal training for operationc personnel follows:-

Phase 1 - Academic training consisting of approximately 14 weeks '

of formal classroom study, depending upon job position, covers training in mathematics, physics, reactor principles, heat transfer, radiological fundamentals, electrical fundamentals, materials, safety analysis, and chemistry.

Phase 2 - A study of all plant _ systiems for approximately 30 weeks. A period of time tracing out systems,- identifying h specific equipment locations, observation of plant evolutions, and reviewing the station operating and equipment instruction manuals is included in this phase. The material presented in '

this phase is directed towards the unit on which the individual J will be applying for a license.  !

[

Phase 3 - Qualification Standard Checkoffs for approximately

• i 76 systems are performed during Phase 3. The checkoffs require ( ) l detailed knowledge of BVPS systems and the ability to perform v q certain operatione using plant control devices or demonstrating L knowledge by simulation. This period requires approximately '-

49 weeks and is directed towards the unit on which the individual will be applying for a license.

slMOLtlT N i Phase 4 - Offsite g training covers a 1 week period. Offsite  ;

training will be conducted in reactor startups and shutdowns to i familiarize the operator with reactor operations when the Hot i License Exam is required without a start-up demonstration. I OD ~ G th FT y a

Phase 5 -(Elant manioulation7b training provides the operator with

" hands-on" training in the area of reactivity manipulations.

.This. training will be provided on either one of the BVPS units, -

I the BVPS simulator, or an offsite simulator. This phase requires f 13 weeks. '

t Phase 6 - Review lectures designed to sum up the entire program i are given to prepare the candidate for the licensing exam. This [

phase requires 8 weeks or more, depending on the individual's i' background.

i-The _ details of the Licensed operator Training program are [

contained in Table 13.2-1. Each candidate's previous experience  !

i l;

13.2-2 T L

m,-

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ATTACRMENT 3 Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft' Safety Evaluation Report Draft SER Section 13.2.1.1: Initial Training Program (excerpt)

In addition, the applicant has not provided information of the simulator-to be used for training. As indicated in the Standard Review Plan, Section 13.2, the simulator used for training should meet the guidelines of Regulatory Guide 1.149. We will review this information when it is received and will report our findings in the final SER.

Response

The response to . Question 630.7 and the discussion provided in FSAR 1.8 provide this information.

I i

c LATTACHMENT 4 Response to Outstanding Issue of the ,

Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 13.2.1.2: Beaver Valley Operating Cross-Training Program i The applicant has indicated that the BVPS operator cross-training pro-gram is.. designed 'to prepare . operators licensed or ' licensable - on BVPS-1

, for. licensing on BVPS-2 to meet the needs .of the operating organization.

The BVPS operator cross-training -program is approximately three to four months in length and includes classroom training in system dif ferences

, and system checkout s , on those systems with significant- differences l between the units. Technical specification difference lectures are also included in the program. The applicant has not provided for our review the details. of the cross-training program. As described in NUREG-1021,

! " Operator Licensing Examiner Standard," for a reactor operator or-senior.

+

reactor operator to be' eligible to hold simultaneous valid licenses on  !

< 'more than one nuclear facility, we require the applicant to provide the justification to demonstrate that the dif ferences between the units are not so significant that they impact the' ability of the licensed person-nel to operate - safety and competently both facilities. Further, the applicant must submit for NRC review the de tails of the training and

. certification program. The analyses and summary of the dif ferences that j must be performed should include:

• Facility design and systems relevant to control room personnel

• Technical Specifications

• Procedures, primarily abnormal and emergency operating procedures

• Control room design and instrument location

• Operational characteristics

, The applicant also should describe the expected method of rotating j personnel between units and the refamiliarization to be conducted before responsibility on a new unit is assumed.

. We will review the details of the applicant's cross-training program

!. when they are received and report our findings in the final SER.

t

Response

The use of NUREG-1021 as an acceptance criterion is beyond the guide-lines of the standard review plan and no basis in the regulations has been provided to justify this request as an outstanding issue in - the safety ' evaluation report. In a memorandum for all NRR employees from Harold R. Denton dated April 28, 1982, he states, " Staff reviewers-should 'not decrease nor go beyond the scope ' and requirements of- any specific'SRP section." The memo closed by saying:

i l

l i.

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'" Implementation of this approach with res pect to . the SRP use ' and revision procedure will add greater stability . to the licensing process and increase confidence that requirements imposed by NRC are congruent _with the regulations and are commensurate with the safety .

value to be expected. Your careful consideration of this memorandum and its. consistent implementation should enable NRR to carry ot**. its statutory function with full consideration of the public interest."

In the absenceL of (1) a description of the regulatory basis, and - (2) standard review plan acceptance criteria for this item, it is necessary for NRR to implement the ' backfit procedure described in Generic Letter 84-08 if this is to remain a SER outstanding issue.

BVPS-2 has recognized the need for an operator cross-training program which is in draft form.

4 6

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ATTACHMENT 5 Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 13.2.1.3: Requalification Training Program (excerpt)

A requalification training program conducted by the applicant for all licensed reactor operators and senior reactor operators will be imple-mented following . the initial licensing. This program will consist of the following:

(1) Lectures The applicant has indicated that a total of six pre planned requal-ification training lectures will be scheduled throughout the ye ar .

Lecture subjects and content will be based on the results of the acnual examination adminis tered tu licensed reactor operators and senior reactor operators. However, the content of the lectures described in the FSAR by the applicant does not cover all the s ubjects listed in Appendix A of 10CFR Part 55. We require the lectures to be modified to ' include the following subjects as listed in Appendix A of 10CFR Part 55 as well as in Enclosure 1 of H. R.

Denton's March 28, 1980, letter:

• Theory and principles or operation

• General and specific plant operating characteristics

• Plant instrumentation and control systems

• Plant protection systems

• Engineered safety systems

• Normal, abnormal, and emergency operating procedures (2) On-the-Job Training The on-the-job training portion of the requalification program will consist of the following segments:

(a) Control Manipulations The applicant has indicat ed that during each two year period, each licensed reactor operator is required to manipulate facil-ity controls through at least 10 evolutions and each licensed senior operator ' is required to manipulate, direct, or evaluate the manipulation on controls through a like number of plant evolutions from any combination of the following evolutions:

• Reactor start-up from suberitical to the point of adding heat-

• Manual control of steam generstor levels during reactor start-up or: shutdown

,

• Placing reactor in hot standby condition form at power condition

• Dilution of reactor coolant system (RCS) boron concentration to achieve A reduc. tion of at least 100 ppm or boron

• Boration of RCS to - achieve an . increase of at least 100 ppm or boron

• Operation of EHC turbine governor controls during' unit start-up

• Manual rod control operation prior to and during genera tor synchronization

• Control rod manipulation during reactor power level changes or greater than 10 percent

• Plant and reactor operation that involve emergency or transient procedures where reactivity is changing

• Rod drop timing test We find that the above applicant's commitment of control manipula-tions required for licensed operators hes not comply wi th the i requirements as specified in Enclosure 4 of H. R. Denton's le tter of March 28, 1980, which requires that, during each two year license period, each licensed reactor operator shall perform all'of

the following listed control manipulations and each licensed senior reactor operator shall perform, direct, or evaluate all of the following control manipulations

• • Plant or reactor startup to include a range such that reactivity feedback from nuclear heat addition is noticeable and heatup rate is established 4

• Plant shutdown

• • Manual control of steam generators and/or feedwater during start-up and shutdown

• Boration and/or dilution during power operation

• • Any significant (> 10%) power changes in manual rod control

• Any reactor power change of 10% or greater where load change is performed with load limit control

• • Loss-of-coolant including:

f a. Significant PWR steam generator leaks

[. b. Inside primary containment i

l

-3

c. Large and small, including leak-rate determination

d. Saturated reactor coolant response l-

• Loss of instrument air

• Loss of electrical power (and/or degraded power sources)

• • Loss of core coolant flow / natural circulation I

• Loss of'condenaer vacuum l

• Loss of service-waterfif required for safety

• Loss of residual heat removal-(RHR). system

• Loss of - component ' cooling . system or cooling to an individual -l component

• Loss of all normal feedwater and feedwater system failure

• • Loss of all feedwater (normal and emergency)

• Loss of protective system channel

• Mispositioned control' rod or rods (or rod - drops)

.

• Inability to drive control rods

• Conditions requiring use of emergency boration

• Fuel. cladding failure or high activity in reactor coolant or offgas

• Turbine or generator trip

• Malfunction of automatic control system (s) which affect reactivity

• Malfunction of reactor coolant pressure / volume control system

• Reactor trip

• Main steam line break (inside or outside containment)

• Nuclear instrumentation failure (s)

The _ starred (*) items shall be performed on an annual basis; all other. items shall be performed on a two year cycle.- An appropriate simulator, which ; reproduces the general operating characteriarics.

of and has similar instrument and control arrangement to BVPS-2, may be used to perform these control manipulations.

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Response

Table 13.2-2 of the FSAR will be . revised to speci fy lecture subjects which are consistent _with ' examination categories of NUREG 1021. . Th e content of these lectures covers those areas specified in Appendix A of 10CFR55 and_ includes the following:

• Theory and principles of operation

• General and specific plant operating characteristics

• Plant instrumentation and control systems

• Plant protection systems

• Engineered safety systems

• Normal, abnormal, and emergency operating procedures The requalification training program required for licensed operators will comply with the requirements as specified in Enclosure 4 of R. R.

Denton's letter of March 28, 1980, which requires that , during each two year -license period, each licensed reactor operator shall perform all of the following listed control manipulations and each licensed senior reactor operator shall perform, direct, or evcluate all of the following control manipulations:

• • Plant or reactor startup to include a range such that reactivity feedback from nuclear heat additior. is noticeable and heatup rate is established

• Plant shutdown

• • Manual control of steam generato rs and/or feedwater during start-up and shutdown

• Boration and/or dilution during power operation

• • Any significant (> 10%) power changes in manual rod control j

• Any reactor power change of 10% or greater where load change is

!. performed with load limit control f

• • Loss-of-coolant including:

l a.,Significant PWR steam generator leaks l b. Inside primary containment I c. Large and small, including leak-rate determination

d. Saturated reactor coolant res ponse f

• Loss of instrument air

' Loss of electrical power (and/or degraded power sources) l

• • Loss'of core coolant flow / natural circulation

• Loss of condenser vacuum

• Loss of service water if. required for safety

• Loss of residual heat removal (RHR) system

' ' Los s of component cooling -system or. cooling to an individual component

• Loss of all normal feedwater and feedwater system failure

• • Loss of all feedwater (normal and emergency)

• Loss of protective system channel

• Mispositioned control rod or rods (or rod drops)

• Inability to drive control rods

• Conditions requiring use of emergency boration

• Fuel cladding failure or high activity in reactor coolant or offgas

• Turbine or generator trip

• Malfunction of automatic control system (s) which affect reactivity

• Malfunction of reactor coolant pressure / volume control system

,

• Reactor trip

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

• Nuclear instrumentation failure (s)

The starred-(*) items shall be performed on an annual basis; all other items shall be performed on a two year cycle. An appropriate simulator, which reproduces the general operating characteristics of and has similar instrument and control' arrangement to BVPS-2, may be used to perform these control manipulations, i

-~.

ATTACHMENT 6 Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report.

Draft SER Section 13.2.1.3(3): Simulator Training (excerpt)

The ap plicant has indicated that some or all of the licensed operators and senior operators may participate in simulator training during .their requalification programs. A simulator may be used to meet the require-ments of the FSAR if the simulator reproduces the general operating characteristics of BVPS-2 and the arrangement of the instrumentation and controls of the simulator is similar to that of BVPS.

We find that the applicant has not committed to the requirement as speci fied in Enclosure 1 of H. R. denton's le tter of March 28, 1980, which requires all licensed operators to participate in a simula tor training program as part of the requalification program. Therefore, we require that the applicant submi t a simulator training program as part of- the requalification program for URC review. We will report the results of our review in the final SER.

Response

As discussed in Attachments 2 and 5, BVPS-2 operators will participate in a simulator requalification program.

I

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ATTACHMENT 7 Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 13.2.1.3(4)(b): Systematic Ob se rvation and Evaluation (c;ccerpt)

The applicant has not addressed the systematic observation and evalua-tion of perfo rmance and competency of licensed reactor operators and senior operators. As described in Appendix A to the 10CFR Part 55, we require the applicant to provide an evaluation program to include systematic observation and evaluation of the performance and competency of licensed reactor operators and senior reactor operators by super-visors and/or training staff members including evaluation of actions taken or to be taken during actual or simulated abnormal and emergency conditions. We will review the applicant's modification of the program to include these subjects and report our findings in the final SER.

Response

The pe rfo rmance and competence of licensed operators and Senior Opera-tors is evaluated at le as t annually by observation or a critique of the manner in which the operators responded in recognizing and managing such events as abnormal occurrences and res ponse to off normal operating conditions or simulated emergency or abnormal operating conditions.

Final evaluation is accomplished by observation while using the control panel of the Beaver Valley Power Station or station simulator control panel.

ATTACHMENT 8 Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft. Safety Evaluation Report Draf t SER Section 13.2.1.3(5): Accelerated Requalification Program (excerpt)

The applicant has not provided the criteria for requiring a licensed individual to participate in an accelerated requalification program. We require an accele rated requalification progr am to be implemented when the performance of a licensed reactor operator or senior reactor opera-tor falls below the following criteria:

• As specified in Enclosure 1 of H. R. Denton's March 28, 1980, letter, the passing grade for the written examination shall be 80% overall and 70% in each category.

• As required in Appendix A to the 10CFR Part 55, where the performance evaluations conducted pursuant to the above section (4)(b), " System-atic Observation and Evaluation," clearly indicate the need.

We will review the applicant's commitment to the above criteria for 4

requiring a licensed reactor operator or senior reactor operator to participate in an accelerated requalification training program, and will report our findings in the final SER.

As indicated in the above , we find that the applicant's requalification training program for licensed reactor operators and senior reactor oper-ators does not fully confo rm to the requirements as specified in the Appendix A to 10CFR Part 55 and in the le tter from H . R. Denton to all power reactor applicants and licensees dated March 28, 1980. Therefore, we have not been able to conclude that the applican' s requalification training program is acceptable.

Response

A licensed operator or Senior Operator whose scoring is less than 80% in any section of the comprehensive annual examination shall be required to attend lectures in those sections of the exam. Should the licensed operator or Senior Operator fail to attain an ave rage of at le as t 80%

overall, with a minimum of 70% percent in each category in the annual examination, he shall be removed from shif t duties and shall participate in accelerated requalification programs under the direction of the Station Supe rviso r of Training. He will be returned to shift duties af ter retesting and achieving an overall average of 80 percent. Lectures will be scheduled in those areas in which a grade of less than 80% was achieved. The NRC will be notified of satisf actory completion of train-ing prior to the individual's return to licensed duties. Provisions have also been made for licensed operators and senior operators to partici-pate in an accelerated requalification program when the results of the systematic observation and evaluation program required by 10CFR55, Appendix A, Paragraph 4.c, clearly indicate the need.

ATTACHMENT 9 Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report-Draf t SER Section '13.2.1.4 (I .A'.2.1): Immediate Upgrading of Reactor' Opera-tor and Senior Reactor Operator Training and Qualification (excerpt)

The applicant's training program includes topics in heat transfer, fluid flow, _ and thermodynamics. However, the applicant has not provided a 2

program for the instructions of ~ these topics in accordance with Enclo-sure 2 of H. R. Denton's March 28, 1980, letter. We require the appli-cant . to . provide this program for us to review, and we ' will report our findings in the final SER.

The applicant's training program does not' include topics in reactor and plant transients. As described ~ in Enclosure l' of H. R. Denton's March 28, 1980, letter, we require the applicant _to modify the training-program to provide emphasis . on reactor and plant transient. We will review the applicant's modification to the training program to include these topics when.it is received and report our findings in the final

+

SER.

The applicant has submitted an outline of a program for training in mitigating core damage. We have reviewed it and find that the outline does not provide us suf ficient information to determine that the appli-i cant's program is comparable in scope and depth - of- training in- various s ubject s to the mitigating core damage training program as outlined in Enclosure 3 of H. R. De nton's March 28, 1980,- letter. Therefore, we require the applicant to provide for us to' review a detailed description of the program in accordance with the guidance as specified in the above

, cited enclosure of H. R. Denton's March 28, 1980, letter. We will review the applicant's program when it is recieved and report our I findings in the final SER.

i Based on our review, we have not been able to conclude that the appli-

- cant of BVPS-2 has satisfied the requirements of this item of the TMI Action Plan.

Response

Refer to Attachment I with regard to heat trans fer, fluid flow ced ther-modynamics. The current lesson plan LP-ATA/MCD-O (attached), " Transient and Accident Analysis / Mitigating Core Damage -- Introduction," provides

, an - outline of subjects which satisfy the topics of (1) training in the use of plant systems to control or mitigate an accident in which the core is severely damaged and (2) reactor and plant transients required by Enclosure 1 of the Denton letter. The portion of this lesson devoted

to mitigating core damage would consist of approximately 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br /> of training for new operators , 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> for requalification of operators, and 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> for other plant staff. Lesson plans LP-NOMCD-1 and LP-NOMCD-15 describe the subject areas of gas generat ion and radiation monitoring, respectively, i

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

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194 Ce,CC2,004l . ,

DUQUESNE LIGET CCMPANT Figure 0.7 Nuclear Division

{' rr.4 4.. unua LZSSON F wi Transient and Accident Analysis / Mitigating 120 Core Damace - Introduction Course Course Hours Laughlin/ Russell January 26, 1984 tructor Data

,/ M/ h_ _,

LP-ATA/MCD-0 (1 Hr.)

Lasson Plan No. (Sequentially From 1)

Ayy m .4 Sy:

INFO Standards Rafarances To Be Quoted 1 Items Issued: (Actac!1 copy of all passauts, quizzes, etc.)

1) Text: BVPS Mitigating Core Damage 2) Course Letter; 3) INPO Standards:

4) Course Schedule; $) Lesson Plan Handouts. ,

Introduction:

I 1.

Purpose:

To delineate the objectives, content, and schedule for the BVPS Transient and Accident Analysis / Mitigating Core Damage Course

2. h eivation: (Discuss how you plan to :mcivata students)

Explain that a lack of knowledge can lead to sericus safety problems and a significant fraction of the NRC SFO and R0 Licensing Exam covers these areas.

3. General outline: (List deca 11ed' ou. cline Section I)

Course objectives, course content, course schedule.

4. . General Student Go'als: (List detailed student objectivesSection II) the course Upon completion of this lesson, the student will be avare of:

4 k objectives, how the course will be conducted, and the course schedule.

t 0.14 ISSCI.3'

- \-. . -_ . . _ . _ - - _ _ _ _

)

1 1

DUQUESNE LIGHT COMPANY

. Nuclear Divisica f

i Nuclear Support Services Department APPROVAL SHEET - LESSON PLAN AND TEXT REVISIONS Document

Title:

LP-ATA/MCD-O Revised Aporoval.

Rev. Subjects Revised.

No. ~(Brief Description) by Signature .Date 1 Added A >!CD course introduction P. R" g

) Jan. 26, 1984 p4

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LP-ATA/MCD-0 STUDENT OBJECTIVES Terminal Objectives Upon completion of this lesson, the student will be aware of the course objectives, the conduct of the course, and the course schedule.

Enabling Objectives j 1. The student will be able to list the course objectives.

2. The student will be able to list the two major topics covered by the course.

3. The student will be able to describe the format of the course.

4. The student will know how absences will be resolved.

5. The student will know the number and frequency of exams given during the course and each ones percentage of the final grade, i

[' ' 6. The student will know the consequences of exam or course failure or

(. near failure.

I f

C

InstructIr's Larson' Plan LP-ATA/MCD-0 Page~ 1 of 4 -l

'\

I Lesson Plan Outline Instructor Notes and Referances I. Issue of Materials Students may keep; tell them to. read i the course letter now.

A. Text B. INPO Standards

'C. Course Schedule II. Introduction of Instructor (s)

A. Name(s)

B. Office location (s)

C. Background (s) (if asked)

III. Introduction to Course ,

A. Scope of the Course

1. Meet or exceed INPO standards

a. Technical Specifications are also covered.

2. Help students understand the plant response

a. During transienta

b. During accide'nts

3. Help students understand how to mitigate various accidents.

4. Prepare students for NRC exams

a. Both R0 and'SRO exams cover the plant response during transients and acci-dents and accident mitigation.

B. Course objectives (upon completion of this course)

1. Students should be able to describe the plant response to various transients.

2. Students should be able to describe the plant response to various accidents.

(

Instruct:r'c Larson Plan Page 2 of 4

• LP-ATA/MCD-0 Lesson Plan Outline Instructor Notes and References

3. Students should be able 'tc explain how to mitigate the consequences of various plant transients and accidents.

C. Course content: Transient and Accident Analysis and Mitigating Core Damage

1. Transient and Accident Analysis

a. Fundamentals Review

b. Power Distribution

c. Transient Anclysis

1) Normal

2) Abnormal ,

d. Accident Analysis

1) Reactivity Addition Accident

( 2) LOCA's

3) Miscellaneous

2. Mitigating Core Damage

a. Post Accident Cooling -

b. Potentially Damaging Operating Condi-tions

c. Small Break Loss of Coolant with No High Head Safety Injection

d. E-0 Procedural Review

e. E-1 Procedural Review

f. Loss of Feedwater Induced Loss of Coolant Accident

g. Main Steam Break Review

h. Steam Generator Tube Rupture Review

/ 1. Excerpt from Incident Evaluation

\w- Ginna SGTR

j. Steam Generator Overfill

Instruet:r's Le: son' Plan LP-ATA/MCD-0 Page 3 of 4 -

Lesson Plan Outline Instructor Notes and References L. Loss of All A.C., E0P-7

l. Incore Thermocouple Maps

m. Vital Process Instrumentation

n. Instrument Qualification and Accident'

Response

o. Accident Response of Excore Instru-mentation

p. Accident Response of Incore Instru-mentation

q. Post Accident Primary Radiochemistry D. Conduct of the Course .

1. The course is broken up into two areas:

Transient and Accident Analysis and Core Mitigating Damage

2. Lessons are presented as a lectur.e Cd a. Prior to the lecture (s) on a lesson, the student will be issued lesson ob-jectives and may be given a text read-ing assignment.

b. Each lecture is approximately one hour long followed by a ten minute break.

c. During the lecture (s), the student should take notes

3. Prior to each exam, a review will be con-ducted.

E. Exams

1. Total of three (3) exams during course

a. Exam 1 - Transient Analysis Exam 2 - Accident Analysis Exam 3 - Mitigating Core Damage and Transient and Accident Analysi s Comprehensive.

b. Exam weighting

\-

1) Exam 1 cnd 2-20% each

O InstructWO Lacson' Plan Page 4 Of 0 LP-ATA/MCD-0 Lesson Plan Outline Instructor Notes and References

) Exam 3 - Trans'ient and Accident Analysis - 20% Mitigating Core Damage - 40%

c. Exam conduct Refer to handout on the Conduct of Training Dept. Exams

d. Exam content

1) Essays

2) Short answers.

3) Calculations and graphs F. Performance

1. Failure of the course will result in an ,

Academic k'arning (< 70%) .

2. Failing any exam (< 70%) will result in a Report of Counseling.

G. Absence

l. Students will have to makeup for lost time Stress that it is the students responsibility to meet with his in-

2. Catch-up time will be on a one for one structor on the day he returns to basis (no overtime!) work. Together they will arrive at a schedule for completion of missed

a. e.g., a student who missed four (4) work.

days of class will have four (4) days after his return to work to make-up all he missed. Concurrently he must learn the new material taught during this make-up period.

H. Course schedule

1. Breifly review course schedule with the Refer to handout; show Course Sche-students. dule.

2. Emphasize that this schedule is only tentative IV. Summary A. Review Objectives Review lesson objectives with the students.

/

-(

i

l COURSE LETTER l' ' Transient and Accident' Analysis and Mitigating Core-Damage The Nuclear Support Services Department of the Duquesne Light Company is committed to meet the INPO Guidelines for Qualification and Training of Licensed Operators. The Transient and Accident Analysis and Mitigating Core Damage Course has been designed to meet or exceed these guidelines.

The Transient and Accident Analysis and Mitigating Core Damage course is presented in a lecture format. Questions and discussion are encouraged during the' lectures. The BVPS Mitigating Core Damage Text is the reference book used for this course. However, the text is supplemented with handouts taken from Westinghouse WCAP's, BVPS Technical Specifications, EVPS Setpoint Study and BVPS updated FSAR.

There are a total of three (3) exams given during this course. These include the final exam which is comprehensive. There is a review session prior to each exam. Subsequent to each' exam, the correct answers to the exam questions are presented to the students.

Accompanying this letter are a copy of the INPO Guidelines for training to recognize and mitigate the consequences of core damage, and the Course Schedule.

Course Objectives Upon successful completion of this course, the student will be able to:

describe the plant response to various transients, describe the plant response to various accidents, and explain how to mitigate the consecuences of various plant transients and accidents.

Course Description Topics include:

Fundamentals Review Power Distribution Transient Analysis

1. Normal

2. Abnormal Accident Analysis

1. Reactivity Addition Accident

2. LOCA's

3. Miscellaneous L

- > ~

Mitigating-Core' Damage

.x ,

1. Poat Accident Cooling

2. Potentially Damaging Operating conditions

3. Small Break Loss of Coolant with No High Head Safety Injection

4. E-0 Procedural Review

5. E-1 Procedural Review

6. Loss of Feedwater Induced Loss of Coolant Accident

7. Main Steam Break Review

8. Steam Generator Tube Rupture Review.

9. Excerp t from Incident Evaluation Ginna SGTR

10. Steam Generator Overfill

11. Loss of All A.C., E0P-7

12. . Incore Thermocouple Maps

13. Vital Process Instrumentation

14. Instrument Qualification and Accident Response

15. Accident Response of Excore Instrumentation

16. Accident Response of Incore Instrumentation

17. Post Accident Primary Radiochemistry Course Temporal Breakdown Course length 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> Lectures 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Student reading, study 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Student examinations 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Examination reviews 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> I

w

, . - . , ,.v .- - - ,-, , , y-, , +- ,~ y

CLASS SCHEDULE TRANSIENT AND ACCIDENT ANALYSIS MITIGATING CORE DAMAGE l

I t

Day 1 Introduction Fundamentals Review Power Distribution Day 2 Power Distribution (Cont.)

Reactor Trip Review Day 3 I&C Review Introduction to Transient Analysis Normal Transient Analysis Day 4 Normal Transient Analysis (Cont.)

Abnormal Transient Analysis Day 5 Abnormal Transient Analysis (Cont.)

I&C Failures Day 6 Study / Review Quiz 1 Day 7 Introduction to Accident Analysis Reactivity Addition Accidents Day 8 Reactivity Addition Accidents (Cont.)

LOCA's Day 9 LOCA's (Cont.)

Miscellaneous Accidents Day 10 Study / Review Quiz 2 1

1 j

DayJ11 Core Cooling Mechanics j

r Potentially Damaging Core Conditions T.

Day 12 Small Break LOCA with no HHSI

' Procedural Review E-0 and E-1 Loss of Feedwater Induced LOCA Day 13 Steam Break E-2 SGTR E-3 Review of Ginna SGTR SG Overfill Day 14 Loss of All AC E0P-7 Incore Thermocouple Maps Vital Process Instrumentation Day 15 Instrument Qualification Accident Response of'Excore Instrumentation Accident Response of Incore Instrumentation Post Accident Radiochemistry Day 16 Study / Review Day 17 Final Exam J

n J

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.a e .<:mv<1 , ...e .t, ,- .w.t ts ~ r 7: .y g, 194 C4,CC3,004 , ,

DUQUESNE LIGHT COMPANY Figure 0.7

/~' Nuclear Division k_-)i . Training Manual LESSON PLAN .

MCD - Post Accident Cooling 2 Course Course Bours Laughlin/ Russell __

May 7, 1984 ructor Date

/ //  % LP-NOMCD-1 Rev. I s' Approved by: Lesson Plan No. (Sequentially From 1)

J Refeiences To Be Quoted: SNUPPS, FSAR, Chapters 4, 6, and 16 (Technical Specifications); WCAP-9600, Sections 2.6, 2.9; NSAC Report " Analysis and Eval-

uation of St. Lucie Unit 1 Natural Circulation Cooldown,"; "Long Term Core Cooling-Boron Considerations, letter trom West. Elec. Corp. the USNRC, CLC-NS-309.

Items. Issued: (Attach copy of all passouts, quizzes, etc.)

Tab'1 of Beaver Valley Mitigating Core Damage Textbook 4

\ss/

Introduction:

I 1.

Purpose:

To make the student aware of various post accident cooling mechanisms.

2. Motivation: (Discuss how you plan to motivate students)

To understand the role the operator plays on the success or failure of

! post accident cooling.

3. General Outline: (List detailed outline Section I)

Introduction, Operator's role in plant safety, core thermal limits, natural circulation, non-condensable gas formation and effects.

4. General Student Go'als: (List detailed student objectivesSection II)

! -Understand operator's role, know ECCS acceptance criteria, understand

, natural circulation cooling.

0.1k ISSUE.3~

l

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l. -

}- DUQUESNE LIGHT COMPANY-

} ,.- .

Nuclear Division

( ,

Nuclear Support Services Department i

APPROVAL SHEET - LESSON PLAN AND TEXT REVISIONS Document

Title:

LP-NOMcn-1 Rev. Subjects Revised Revised Approval No. (Brief Description) by Signature Date 1 Developed lesson plan from MCD man- P. Russell ual. No change to manudl.

&Okmd O

._c. '

Si( '

LP-NOMCD ,

-"s - POST ^ ACCIDENT COOLING-

- () S tudent- Obj ect'ives u -

TERMINAL OBJECTIVE-p -

~

-The student'sho'uld understand 7 the key role he-plays in: mitigating-the iconsequences of any accident'that:could;1ead to core damage.---In addi-

- tion, he'should understa64 what effects his actions can have on the~

success;or failure of post accident cooling mechanisms.

[ . ENABLING OBJECTIVES p ,

U ' After studying the text'in conjunction with other'specified-references

~ and the lecture, the student.should be able--to:

1) Explain the operator's role, and the role of technical specifi -

i cations, in ensuring plant safety.

(" .

- \. 2) List the ECCS acceptance criteria, and explain how each item-relates to long term core cooling.

1-

, 3) Understand the operator's relationship with the ECCS accep-f:

e tance criteria.

i V

j 4) Discuss how the operator can determine the status of natural circulation, and what factors can promote / retard the effective-l ness of core cooling via natural circulation.

t f

i.

i I

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a.

, w. ~

e+w----,e- *FP*T

hat =uctor's LassonGlan Page 1 of 10 I Lasson Plan Cu

• a hatructor Nocas and Rafarances I

(

ost Accident Cooling I. Introduction A. Discuss various aspects of core cooling.

B. Know what cooling mech are available or

. should be available and what function he per-forms to enhance or alter these cooling mech.

C. Associated Technical Specs. ECCS acceptance criteria, natural circulation.

D. Operator must be able to recognize when the plant is not responding as predicted.

II. Operator's Role in Plant Safety A. FSAR Analysis

1. Plant operating in a given band via proper control systems.

(\

2.

1 Plant assumed to 'oe in steady state prior to the transient.

3. Plant operated within Tech. Spec. limits B. Operation Within Band LP-NOMCD-1.1

1. Operator acts as backup to auto control system.

2. Plant operation within certain deviations allowed.

3. Key deviations

- RCS leakage

- RCS activity

- Power distribution

- Safeguards equipment out of service G

1 1

Instructrr's I. arson' Plan i Page,2 of 10 l I,assca Plan Outlina Instructor Notes and Rafarences l t' N 1

V 4. Example

a. 100% power with several S/G safeties LP-NOMCD-1.2 inoperable

- Complete loss of load (condition ~

11)

- Plant will overpressurize

5. Tech. Specs. assist the' operator in en-suring the FSAR results remain valid.

6. Even with these operating restraints it is possible degraded core conditions coulc exist.

a. Operator must then take corrective action.

III. Core Thermal Limits A. 10 CFR 50.46 acceptance criteria (5) LP-NOMCD-Table 1.1

/3

('-~#) B. Cladding. oxidation - drives the success with LP-NOFCD-Table 1.2 which the other four are met.

1. Reaction - ZR ZR + 2H2O + ZRO + 2H2 + Heat

2. Eaker - Just Eq.

a. du/dt = .3937/r e

• RT w - clad thickness oxidized T - clad temp in *K R - 1.987 cal /g-mole, K t- time in seconds E - 45,500 cal /g-Mole

3. Important point is oxidation rate in-creases exponentially with temperature.

a. 1800*f - significant rate

b. 2000*f - 17% cladding oxid. limit vithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

['}

-v

c. 2200*f - rate is accellerated rapidly

d. 4800*f " auto-catalytic" L

4

Inst =uct;r's La:sor/!1an Page 3 of 10 l

Lesson Plan Outline Instructor Notes and References

( l

4. 17% oxidation of clad thickness

a. Loss of strength and ductility

b. Consider thermal shock during SI C. H, Generation - < 1% of the theoretical vol.

of H, that could be generated if all the cladaing reacted.

1. Westinghouse 312 % 34,000 pounds of ZR 3

a. Potentially 270,000 ft of H STP)inRCSwouldrecombine$ue(at to radiation but in containment during LOCA could mix with free oxygen. LP-NOMCD-1.3

b. 4% flammable

c. 18-54% explosive

d. 54-75% flammable

,-_ e. > 75% not enough oxygen I

's_ / 2. Other H S urces LP-NOMCD-Table 1.3 2

a. Stainless steel

- 3Fe + 4H 2O + Fe 03 3 + 4H2+

- Accelerate at 2300*f (hegligable until this temperature)

- 5000 btu /lbeFe

- does not surround fuel sc not as big of a concern

- Melts at 2500*f

b. Aluminum N 2000 lbs

- A1 + 3H 2O + A1 033+H2

- 300*f - accelerated

! - Conduit, coolers l

l \

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Instructor'c Le: son' Plan Iage 4 of 10 Lesson Plan Que" e Instrue:or Notes and References w

( l

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• L**#

- 2al + 2NaOH (spray + A12 +syst) suppression 3H2 (#

- produce large amounts of H but 2

the amount of Al. is limited in-side containment.

c. Radiolysis of coolant-in system and in sump - significant source.

d. H9 expanded from RCS inventory (25-35 cc/KG in RCS)

D. Boron Precipitation

1. Fourth and fifth limits of Table 1 re-quire long term coolable geometry.

2. Assume LOCA and loss of natural circula-tion in good loops - would have core boiling with steam loss to ambient as heat removal.

,. a. Boric acid concentration would + to due to low voletility.

(\n,)

b. Flow blockage, and heat transfer,

c. Hot leg recire. - 14.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after LOCA to reverse core flow and quench boiling flushes out cold leg.

d. If hot leg break, would still quench boiling.

e. Computer calculation performed to predict boron concentration in core nt 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after cold leg break.

f. Assumptions

- Core vol. up to Th leg lip con-  ;

sidered

- Any H B0 v latility ignored 2 3 l l

- t in specific gravity as concentra- l tion + is ignored (this would cause boric acid to settle to lower

/N t

plenum). LP-NOMCD-Table 1.4

. _. ]

LP-NOMCD - 1.4

l Inst =uctcr's La: son' Plan Page 5 of 10 Lasson Plan Ouet4 e Instructor Notas and References e IV. Natural Circulation A. Flow due to fluid density difference within the fluid.

B. Requires several conditions LP-NOMCD-1.5

1. Heat source

2. Heat sink - (Temp. would equalize)

3. Height difference C. Thermal Driving Head - pressure difference LP-NOMCD-1.6 due to columns of fluids at different densities .

1. TDR - g/g 5 o (z) dz for censtant source and sink TDH = -g/g"p (apaZ) do = go - H

2. Head is defined as D/c

'O t / Head = TDH/D f t Ibf/ lbe N/ N.C Avg TDH/0 = IL *2 avg Loop . V (gg = flow resistance D. V = TDH = g/g .D AZ Loop p g c avg Loop p (g

1. The following equations can be found from the above equation.

a. TDH a AT

- I a AT 1/9

~

b. M 3/2

c. Q a AT l

E. Another obvious requirement for nat. cire. flov '  ;

is a complete unobstructed loop for flow.

1. Ah % 20' - T H n zzle t tube sheet (O) ,

2. Show TDH Z/p curve - discuss overfeeding S/G.

l Instructor's La::sen ePlan  !

Pace 6 of 10 Lasson Plan Ouc W e Instructor Notas and Rafarences

]

F. Design criteria to t efficiency of N.C

1. IZR level > 50%

2. PZR press. > 2000 PSIA

3. S/G level in narrow range in at least one S/G.

a. 2000 psia results in at least 15*

subcooled at core exit for 100% power T value, this ensures no voids H

forming so PZR level is a valid indication that the core is water filled.

b. The S/G requirement ensures a heat si nk.

(narrow range ind, above the tubes).

G. Indications of heat removal .

1. RCS At < full power At 3/2 s a. Q a at

\l b. 2/3 At a Q

c. Q=mC at p

- (8%) = (10%) (.8%) actual for N.C.

2. Core Exit T/C's constant or +

3. S/G press. + or + at a rate equiv. to rate of RCS temps. + while maintaining S/G levels.

- If N.C. stopped - steam press. would

+ quickly as S/G cools RCS water which is not flowing.

H. It is desirable to maintain RCS press. with PRZ heaters or "ith bubble in head controlling.

I. Voids are a concern if they collect in tube bend area, this could block flow. AFW should condense steam voids if level in narrow band.

t I

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lastructor's I.asson' Plan Pae, 7 of in Lasson Plan OutTd e Instructor Notas and Rafarances 7-i.

v

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J. St. Lucie Unit 1, July 11, 1980

1. Elect. failure lost cooling water to RCP's

2. Rx trip and RCP's secured - natural cir-culation cooldown.

a. 15-20 min. establish N.C. LP-NOMCD-1.7

b. 35 min. commensed cooldown LP-NOMCD-1.8

c. PRZ level irregular at 7 AM - CE LP-NOMCD-1.9 engineers agree steam bubble in head. -

3. "T.g Plants", W recommends a cooldown .

rate of 25'f/ hour compared to 60*f/ hour that they had "Tc" Plants" W recommends

~ '

50*f/ hour C/D rate maximum.

4. If it is apparent that expansion of the steam bubble is blocking N.C. flow, 4

/^ cooldown rate to collapse bubble by (s,)% cooling head more effectively.

5. The steam bubble in the head will control LP-NOMCD-1.10 press at sat (5 - 6:30 a.m.) - press was attempted to be+ to use low head systems but the steam void held P+.

K. If flow is +, the At would t, Tp would ap- LP-NOMCD-1.ll proach sat. - but TDH + (AP) this figure assumes Tp at sat.

1. Ex. if T at 300*f and RCS press, at C

800 psia system > psf TDH.

2. If these T "*****' H below T Cs 'ince only s 100 PSF TDH is requirea for natural circulation flow.

3. If N.C. flow was impeded the AT would

+ until boiling at T

• H i L. What could impede natural circulation flow?

i

1. Non-condensable gas formation.

O m-

b m etr.r'o Larson* Plan Page 8 of 10 I.asson Plan Queline Instructor Notas and Rafarances O) i

a. TMI - severely affected - plant specific LP-NOMCD-1.12 problem -

b. Collect in high points of system (RCS head, PRZ, U-tubes of S/G)

c. S/G req'd for heat removal

- 2" brk - needed for % 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

- 1" brk - needed for N 1 day V. Sources.of Noncondensable Cases

~

r; A. Dissolution of H 2- P + H, released assumed relea'se.d .when press. reaches saturation and l

H, conce'ntration is assumed at 50 cc/KG. (25 -

35 cc/KG):, .

B. Radiolysis

^

y,N

1. 2H2 O ; 2H2+02

2. P O Starts at 5 cc/KG concentration .1

\-- 3. Sump' assumed negligable i C. PRZ Vapor Space -

1. 600 scf H 2

2. Assumed ideal gas law expansion D. ZR - Water Reaction

1. Temp dependent

2. Computer calculated core temps. used.

E. Accumulator N 2

1. Assumed released under ideal expansion F. Fission Gasses

1. Released if fuel rods burst

2. Calculation in computer code.

(m

Instruct r'o Larson* Plan P 9 of 10 Lesson Plan Outid"e Instructor Notes and References s

i!

l G. Helium

1. If clad bursts l

l H. Dissolution of gasses from S.I. flow.

l. Assumed to be air saturated (18 cc/Kg of varicus gasses).

I. Conservative assumptions for computer code.

1. Perfect mixing of S.I. and RCS (max gas reaches core)

2. No reabsorption of released gas.

3. All non-cond. gasser released at point of saturation.

J. Breakdown Time Event

,- y 90 sec dissolution begins

( ) 144 sec radiolvsis starts

\

(Sce/Kg) 190 see PRZ empty 1350 sec Zr-H O begins 2100 sec min.2RCS press. (no more gasses released) 4050 see Bkr removes all decay heat

• accum did not inject ati all.

1. Calculates 1648 ft at STP gorrected LP-NOMCD-Table VI for s 800 psia yields % 50 ft 3

2.

Model53s70ft of bend radius (600 ft total volume of tubes).

3. If all gas was distributed evenly s 237.

of total bend area would be lost.

K. Noncondensable Gas Generation

- +s Is  !

\

~-

Inst =ucttr's La: son' Plan Page 10 of 10 Lesson Plan Outline Inst =uctor Notas and Rafarences

(~g

• t, Mechanisms  % Production Boiling, flashing, dissociation 26.5%

Radiolysis 39.1 PRZ Vapor Space 18.2 ZR-H2 O Reaction 16.2 Accum. N 2 0 Fission gas - helium 0 L; TMI-2 problem LP-NOMCD-1.12

l. Noncondensable collected in essentially noncoolable section of RCS (bend) r.

2 '. Compare to Westinghouse design.

VI. Summary:

A. Self Assessment t'

N .I Ou

mu sce u-w 3. a . .c o.o .. -

gr = .y . . . r, g mu c 194 C4,CC3,004 , ,

DUQUESNE LIGHT COMPANY Figure 0.7 -

'f ]

(y_ Nuclear Division

Training Manual-LESSON PLAN .

MCD ' Post Accident Primarv Radio Chemistry ~ 9 Course Course Hours i

4. Laughlin/ Russell vav 19: tora ctor Date V = - - - _ LP-NOMCD-15 Rev. 1

/ -Approved By: Lesson Plan No. (Sequentially From 1)

References To Be Quoted: Water Cool. Tech. of Power Rescrors? PWR Tach. h m1-

, Radiation Analysis Design Manual: Amer. Nat. Stan. Source Term Sonc.- NRAr-1?

) Tech. Staff Anal. Report of Core Damage; Wash-1400, Appendix VII: USNRC Ree.

Guide 1.109 .

Items Issued:, Aopendix !(Attach copy of all passouts, quizzes, etc.)

Tab 15 of Beaver Valley Mitigating Core Damage Textbook.

i O

Introduction:

j 1.

Purpose:

To make the student aware of the impact of core damage on primary radio-

. chemistry.

2. Motivation: (Discuss how you plan to motivate students)

To understand the potential hazards and information obtainable from 4

primary samples following an accident.

3. General Outline: (List detailed ~ outline Section I)

Introduction, Baseline data, Incore releases, Rod bursts, Radiological hazards.

l 1

-4. General Student Goals: (List detailed student objectivesSection II)

Describe incore release mechanisms. Discuss radiological hazards asso-ciated with primary samples.

l I

l 0.'14 ISSUE 3~

(

, - . _ _ . , - . . _ . . , _ . _ . . , _ . _ . , _ _ . . . . . . _ . . _ _ . . _ . ~ . _ . , , _ . . _ . . . _ . , _ . _ , - . . - _ . . . . . . _ _ _ , .

DUQUESNE LIGHT COMPANY Nuclear Division (p. Nuclear Support Services Department t

APPROVAL SHEET - LESSON PLAN AND TEXT REVISIONS l

Document

Title:

LP-NOMCD-15 Rev. Subjects Revised Revised Approval No. (Brief Description) by , Signature Date f_ -

h 1 Developed lesson plan from MCD man- P. Russell ..

ual. No change to manual. C -

fJG"A0 m_gif

~

I O

4 LP-NOMCD-15

. /T Post Primary Radiochemistry-

' d STUDENT OBJECTIVES Terminal Objective The student should understand the key role primary samples can. play in

,. determining the consequences of.any accident that could lead to core damage. In' addition, he should understand the radiological hazards

associated with post accident primary s'nples. a

-Enabling Objectives Afterstudyingthetext-inconjuncidonwithotherspecifiedreferences and the lecture,-the student should be able to:

1. Describe the incore relsase mechanisms for fuel failure and their effects on primary, radiochemistry.

[, 2. Estimate the' effects of rod burst on 8-y activity.

3. Estimate the'effect of fuel melt on 6-y activity.

4. Discusstheradiologicalj. hazards of sampling and how the hazards vary with time.

5. Relate consequence of transferring primary water outside the containment following core damage.

i l -

(

i i.

l nv

, . , . . . * - ---r- ~ ,c-~ w ' * ' ~ ~ ~ ' ~ ~ ~ * " ' ~ ' ' ~ ' ' ~ ~ ' ~ ~ ^ ' ' ' " ~~" ~ * ~

Instruct:r's I, arson' Plan Page 1 of 11 I,asson Plan OucWe Instructor Notes and Rafarences

1. Introduction A. Two levels of core damage

1. Fuel rod cladding cracked or ruptured

a. Cap release or rod rupture accident

2. Partial fuel meltdown B. 10% of fuel rods incore damaged II. Baseline Plant and Assumptions A. Assumptions

1. Core power rating (Mwt) 2900

2. Specific thermal power (MW/MTU) 40

3. Number of reactor coolant loops 3 f')

N_/

4. System water volume (ft ) 8910

5. Normal operating pressure (psia) 2250

, 6. Average temperature in core 590*F

7. Core time in life Middle

8. Design basis failed fuel fraction 0.01 B. Tables

1. Table 1 these assumptions. LP-NOMCD-Table 15.1

2. Table 2 normal activity levels LP-NOMCD-Table 15.2 by isotope

a. One-tenth of design basis

3. Table 3 total core radionuclide inventory LP-NOMCD-Table-15.3

4. Based on ORIGEN computer code

a. Time dependent concentrations of f~N (v )

Inst =uctor'c Larson* Plan Page 2 of 11

. Lesson Plan Que" e Instructor Notas and References

- 256 activation products

- 461 fission products

- 82 transuranics (elements having atomic number greater than Uran-ium (92))

5. Total core inventory of all fission pro-ducts if distributed evenly in coolant is 100Ci/ml

6. Principal contributors to normal activity

a. Iodine

b. Cesium

c. Sum approximately 1 micro Ci/ gram

- Actual plants experience LP-NOMCD-Table 15.4

.1 micro Ci/ gram g.

( ) 7. Baseline will be 1 micro Ci/ gram v

III. Incore Release and Escape Mechanisms A. Normal release mechanism is by diffusion from fuel to gap (release fraction) then through manufacturing or corrosion defects in cladding (escape fraction)

B. Escape Coefficients LP-NOMCD-Table 15.5

1. Magnitude dependents on volatility of nuclide

a. Gaseous higher

2. Iodine nuclides gaseous and fuel rod temperatures but very reactive with zirconium and cesium

3. Strontium nuclides exist in metallic or oxide form and do not readily escape through cladding defects.

<x v

I

Instruct:x'o Larson' Plan Page 3 of 11 Lasson Plan Queid e Instructor Notes and References IV. Rod Burst Effects on Coolant Radiochemistry A. Local fuel rod temperatures reach point where internal pressure causes rod to rupture

1. Fuel does not lose integrity B. Increase in Coolant Activity fer 10% fuel LP-NOMCD-Table 15.6 rupture CA = "( }(

V where CA = coolant activity in uCi/ml '

NI = nuclide inventory in the core EF = escape fraction for nuclide class l V = coolant volume (or mass at 590*F, 2250 l psia)

FF = failed-fuel fraction l l

Gap Release Fraction = .03 (only 3% escapes  !

l r] to gap from fuel pellet).

l Q,)

, Cap Escape Fraction (EF) = .03 since 100% of

( the gas released from the fuel will escape to the coolant for a rupture of the fuel cladding.

For Xe, Kr NI = 30.85 x 10 C1 (data from coolant inven-tory list, Table 3; total of Kr, Xe isotopes)

V = 8.91 x 10 ft x 2.83 x 10 cc/ft x

.72 g/cc V = 1.82 x 10 grams of coolant at 590*, 2250 psia CA = (30.85 x 10 Ci)(.03)(.10) = .005 Ci/g =

1.82 x 10 g 5 x 10' Ci/g O

( >

v-

Instructor's La: son'?lan Pig'2 4 of 11 Lasson Plan Que Ne Instructor Notas and References or 5 x 10 uCi/ml (at sample temperatures) of noble gases in coolant. Much of this noblegasconcentrationisNm"Kr8Y*#Y5b# 8 ha Xe{

ifeggelides{geKr , , Kr , and Kr and Xe will likely be the only -

nuclides present at sampling time and thus the actual noble gas concentration will be about 2.5 x 103 uCi/ml. This calculation is summarized in Table 7, with the results show-ing that for ten percent failed'. fuel the Kr/Xe activity increases by about 104- . LP-NOMCD-Table 15.7 For Iodine: r.

. x . 01D ( . O CA = = .006.Ci/g f 1.82 x 10 g I

.or -

3 6 x 10 uCi/mi of Iodine. isotopes at sample conditions.

N i /"N For Cesium: .t

~-

6 CA = (13.8 x 10 C1)(.05)(.1) 8

= .0004 C1/g 1.82 x 10 g or 4 x 102uCi/mi of Cs at sample conditions.

The Cs contribution is still an order of mag-nitude below that from the I isotopes. The contributions of other isotopes such as Sr, Ba, and Te are not calculated for the gross activity since the estimated escape fractions are much lower for these nuclides and their total inventory is slightly lower. For exam-ple, consider Sr:

CA =

• (* ( "

8 1.82 x 10 g

• 0.1 uCi/ml, at sample conditions indicating that contributions due to Sr are low compared to the I and Cs activity.

["N

\

..,~

l k .

Instrccror's LarsonClan Page 5 of 11 Lasson Plan Out""e Instructor Notes and Paferences C. Total I and Cs activity

1. Approximately 7 x 10 micro Ci/ml

2. 7000 times normal D. Sample results vary with time after accident

1. Iodine decrease by factor of 3 after one day E. Compare with TMI-2 LP-NOMCD-Table 15.8

1. Results are comparable V. Mechanisms for Extensive Core Damage Radiochemistry Effects A. Fuel rods clad with zirconium alloy metal

1. At high temperatures react with water

"'}

'J Zr + 2H 2O + Zr02 + 2H2

2. Produces heat and hydrogen B. Oxidation limit for cladding can be reached in one hour at cladding temperatures of 2000*F.

C. At 3450*F oxidized Zr can melt.

1. UO, may dissolve in liquid oxidized zirconium and release its activity

2. Significant since melting temperature of UO2 is 5200*F.

D. Increase in coolant activity for 10% fuel LP-NOMCD-Table 15.9 meltdown 1

For Xe and Kr:

CA =

• b *

= .153 Ci/g 1.82 x 10 g l

or A

1.53 x 105uC1/ml of noble gas in the coolant.

(%. .)!

l

Inst =2ct:r'c Larson' Plan Page 6 of 11 Lasson Plan Ou" W e Instructor Notes ari References For Iodine:

CA = = .32 Ci/g 1.82 x 10 g or 5

3.2 x 10 uCi/mi at sample conditions.

For Cesium:

. x 0 CO (.8)(.D = .006 C1/g CA =

1.82 x 10 g or 6.0 x 10 uCi/ml at sample conditions.

For Strontium:

CA = = 0.1 Ci/g 1.82 x 10 g

%.s 1 x 10 uCi/n1 at sample conditions E. Sr contribution more than Cs since Sr escape fraction changed by several orders of magni-tude.

1. Helps letermine fuel rupture or melt.

F. Gross degassed activity increase by factor of 300,000 G. Serious accident will also release U-235 and Pu-299

1. Use spectral analysis

2. Check for alpha activity H. Dilution by safety injection water not includ 3d

1. Reduce by up to factor of 5 (D

N.

Instructor'c La son Plan Pagn 7 of 11 Lasson Plan Outline Instructor Notes and References I. The previous calculations also did not account for the effects of half-life between the time of the fuel damage and the time of sampling.

The short-lived nuclides of Kr and Xe would reduce noble gases by a factor of 2, short-lived I nuclides would reduce I by a factor of 4, and short-lived Sr nuclides would reduce Sr cencentrations by about 25 percent if the sample were taken one day after the accident.

J. TMI-2

1. Significant Sr fraction

2. Indicates some fuel melting VI. Radiological Hazards of Sampling A. Assumptions 9

1. 100 c1 depressurized sample collected

,r'y 2. Fuel rupture accident 10% fuel s

'~

3. Noble gas in coolant about 5 x 10 3 micro Ci/ml.

B. Radiological dose calculation LP-NOMCD-Table 15.10 For a sample taken one day after the accident, theactivitywillbeprgmarilyXe-133witha level of about 2.5 x 10 uci/ml. Total noble gas in the sample is thus:

Total Activity = 2.5 x 103 uCi/ml x 100 ml =

5 2.5 x 10 uci = .25 Ci This gas will escape the coolant and fill the 3

sampling room (3m x 3m x 4m or 36 m ) giving a noble gas concentration of (assuming no ventilation):

.25 C1/36 m = .007 C1/m 1

/

\v /

Inst =unt'x'c I.asson'?lan Pag: 8 of 11 Lasson Plan Ouch"a Instructor Notas and References From Table 10 the direct whole body dose rate from_gsemi-igfinitecloudofXe-133is2.94 mrem-m /pCi-yr and so the dose rate to x 10 the sampler can be calculated:

Dose Rate = (7 x 10 pCi/m )(2.94 x 10~ mrem-m /pCi-yr)

Dose Rate = 21 x 10 5mrem /yr The total dose to the operator can be calcu-lated assuming he spende 15 minutes in the area.

15 min. x 21 x 10 5mrem /vr se = 60 min /hr r 24 hr/ day x 364 day /yr Dose = 60 mrem or .06 Rem.

The direct radiation dose to the operator from the 100 mi sample bottle can be estimated from the Curie-Meter-Rem rule, or if the iso-(n) v topes are known, from the equation:

6 Dose Rate (R,' =

d~

The total activity in the sample considering just the I and Cs activity, as previously cal-culated, is (one day af ter accident):

Total Activity = 6 x 10 uCi/ml x 100 ml =

5 6 x 10 uCi or .6 C1 (note that only Cs activity was con-sidered since I has only low energy gammas and Sr90 is a beta emitter).

Using the Curie-Meter-Rem rule the dose rate would be .6 Rem /hr at one meter. This assumes the activity to be due to Co-60 (this gives conservative results). If the operator is actually one foot from the sample bottle the dose rate is:

r

, D - D, ( )2 .6(j)2 - 5.4 Rem /hr U,

Inst =1ctor'c La son' Plan p,g, 9 og 11 1

Lasson Plan Outid e Instructor Notas and References and the operator can receive 1.4 Rem from thi:

source during a 15 minute exposure.

Using the expression D.R. = 6CE/d the dose rate would be 2.4 R/hr, assuming the sample-bottle activity was predominantly due to Cs with a gamma energy of .66 Mev.

C. Dose rates very high special precautions are required.

D. Sample Spill LP-NOMCD-Table 15.10 LP-NOMCD-Table 15.11

1. 1-131, I-133, and Cs primary nuclides of LP-NOMCD-Table 15.12 concern.

2. Some go off with gases but 98.5% remain.

3. Assume 100 ml bottle spills one half of its contents over a 2 x 2 meter area.

I-131 contamination level is:

,rx

!  ! Total Activity = V x CA =

C/ sample 7

(7.7 x 10 C1) ( .017) ( .1) 8 1.83 x 10 g of coolant x 100 g of sample Total Activity = .07 Ci the contamination level is:

Cor.tamination = (1/2)(.07)(1 x 10 pCi/C1) 2m x 2m

= .9 x 10 pCi/m The direct radiation dose rate is calculated using the appropriate conversion factor from Table 11.

Dose rate = (.9 x 10 10 pCI/m )(2.8 x 10~

mrem-m /pCi-hr)

,_ Dose rate = 25 mrem /hr

!v) 1

Instruct:.r'o Lasson" Plan Peg: 10 of 11 Lasson Plan Outid-e Instructor Notes and Rafarences In 15 minutes the operator would receive 6 mrem due to I-131 activity. Calculations for other nuclides (i.e. I-133, Cs-136, C2-137) are left to the student.

Another exposure pathway is inhalation of par-ticulate radioactivity which gets into the air following the spill. Again, continuing as above, the airborne level is:

{

g9 g Activity spilled x fraction airborre j Volume of room l {

l where fraction airborne is .001 for a cold spill.

Thus for I-131 (assuming no ventilation):

Airborne activity = 5 x 01 -6 -

= 1 x 10 uCi/ml  !

l e l

(m) For the inhalation case it is important to

/

remember that different nuclides have differ- i ent critical organs. For I-131 the critical l organ is the thyroid and the annual dose to the thyroid per picocurie inhaled is 1.5 x 3

10 mrem /pci inhaleo (Table 12). The breath-l ing rate for an adult is assumed to be 1.5 x 10 ml/ min.

Total inhaled = 1 x 10 -6 uCi/ml x 1.5 x 10' ml/ min x 15 min

= .23 uCi or .23 x 106 pCi and the total dose to the thyroid is:

TD = .23 x 10 pCi x 1.5 x 10-3 mrem /pci 6

= .34 Rem.

VII. Summary A. Key Points e

\v/

Instructor's Larson* Plan Pags 11 of 11 Lasson Plan Outid"e Instructor Notes and References

1. Estimates of the extent and nature of a fuel damaging accident can be made fror the results of radiochemical samples following the accident.

2. Varying mechanisms exist for fission pro-duct radionuclides to reach the coolant.

These include diffusion through crack or I pinhole defects during operation, releasc through oxidized or melted fuel elements and cladding, and leaching from broken and oxidized fuel rods. ,

c.

3. There are extensive hazards associated  !

with the sampling operation. These haz- l ards must be minimized by using special sampling equipment which will contain the fission product- gases and particulatc s -

and will shield the operator from the direct radiation dose from the sample bottle itself.

/T B. Self-assessment .I .

v 1

n v

ATTACHMENT 10 Response to Outstanding Issue of the Beaver Valley Power Station Unit No. 2 -

Draft Safety Evaluation Report' Draft SER Section .13.2.1.4 (I.A.2.3): Administration - of Training Program (excerpt).

As specified in Enclosure 1 of H. R. Denton's March 28, 1980, letter, we require that all . instructors who teach systems, integrated responses, transient, and simulator corses shall be SRO certified and will' continue to participate in appropriate requalification programs. Vendor-supplied instructors who teach the above subjects shall also be similarly certi-fied. Other members of the permanent or nonpermanent training staff who -

are responsible ' for teaching technical subjects, such as reactor theory, heat trans fe r , fluid mechanics, thermodynamics , health -physics ,- chemis-try, and instrumentation are not~ expected to have an RO or SRO license.

Guest lecturers considered to be used on a limited bases shall be moni-

~

tored .by a qualified instructor. These guest lecturers'. are exempt from the SRO criterion.

Based on our review, we find that the applicant of the BVPS-2 has not committed to comply with the above requirements of this item of the TMI Action Plan.

Response

Beaver Valley training meets the requirements speci fied in Enclosure l~~

of the Denton le t t er . -All instructors who teach integrated responses, transients, and simulator courses are SRO . certified or licensed.

Instructors who teach systems are either SRO certified, licensed, .or designated and qualified system ex pe r t s . SRO licensed or certified instructors are enrolled in appropriate requalification programs.

s

e -

ATTACHMENT 11 Response to Outstanding Issue of the.

Beaver Valley Power Station Unit No. 2.

Draft Safety Evaluation Report Draft SER Section 13.2.2: Shift Technical Advisor Training The applicant has provided a training program for the Shif t Technical Advisors (STA). We have reviewed the program and find that it is not comparable in scope and depth of training in various subjects to the STA training program as outlined in NUREG-0737, Appendix C. Therefore, we require the applicant to provide for our review a de tailed training program for ETA in accordance with the guidance as specified in NUREG-0737, Appendix C. We will report our findings in the final SER.

Response

The STA training program is comprised of the following attributes as required by NUREG-0737, Appendix C.

B 6.1 Education 6.1.1 Prerequisites Beyond High School Diploma It is assumed that many candidates may have received the previous training and are qualified to begin the coursework prescribed in 6.1.2. Prerequisite education considered neces-sary for successful completion of the advanced coursework is identified below. This coursework may be waived without formal documentation of specific course completion.

Contact Hours Mathematics 90 Trigonometry, Analytical Geometry, College Algebra Chemistry Inorganic Chemistry 30 Physics 150 Engineering Physics (heat, mechanics, light sound, electricity and magnetism)

TOTAL: 270

6.1.2 College Level Fundamental Education Contact Hours Mathematics 90

. Engineering mathematics-through the introduction to ordinary dif ferential equations and the utilization of Laplace transforms to interpret control response Reactor Theory 100 Atomic and Nuclear Physics Statics, through 2 group Dif fusion Theory Dynamics, Point Kinetics, Reactivity Feedback Reactor Chemistry 30 Inorganic Chemistry (as related to reactor systems) Corrosion - Reaction Rates Nuclear Materials 40 Strength of Materials Reactor Material Properties (phase diagrams, fuel densification)

Thermal Sciences (for nuclear systems) 120 Thermodynamics Laws of Thermodynamics Properties of Water and Steam Steam Cycles and Ef ficiency Fluid Dynamics Bernoulli's Equation Fluid Friction and Head Loss Elevation Head Pump and System Characteristics Two Phase Flow Heat Transfer Methods of Heat Transfer Boiling Heat Transfer Heat Exchangers

Electrical Sciences 60 Electronics-(Circuit theory, digital electronics)

Motors , Generators , Trans formers ,

Switchgear Instrumentation and Control Theory .

Nuclear Instrumentation and' Control 40J Radiation Detectors Reactor Instrumentation Reactivity Control and Feedback Nuclear Radiation Protection and Health Physics 40 Biological Effects Radiation Survey Instrumentation Shielding TOTAL 520 6.2 Applied Fundamentals - Plant Specific In addition to the general education requirements described in Section 6.1, all STA's shall complete the following training at the-college level tailored to the speci fic plant at which the STA is assigned or a plant of similar de s ign. It may be presented separately from or may be integrated with the education described in Section 6.1 Subject / Topics Contact Hours Plant Specific Reactor Technology (including core physics data)

Plant Chemistry and Corrosion Control Reactor Instrumentation and Control Reactor Plant Materials Reactor Plant Thermal Cycle TOTAL 120 6.3 Management / Supervisory Skills Subject Contact Hours.

Leadership

Interpersonal Communication l' Motivation of Personn> '

l Problem and Socisional Analysis i

i

L Command Responsibilities and LimitD

! Stress f Human Behavior TOTAL 40 6,4 Plant Systems The training program shall cover the following sys tems along with others considered necessary for a specific plant.

System Contact Hours Emergency Core Cooling Emergency Cooling Water .l Emergency Electrical Power, AC and DC Reactor trotection Reactor Coolant Reactor Coolant Inventory and Chemistry Control Containment System (including Containment Cooling)

Closed Cooling Water Nuclear Instrumentation l Non-Nuclear Instrumentation Reactor Control ~

Containment Hydrogen Monitoring and Control Radioactive Waste Disposal (liquid, gas, solid)

Emergency Control Air Condensate and Main Feedwater Auxiliary Feedwater Steam Generator Level Control

, Main Steam Loose Parts Monitoring Status Monitoring (including Process Computer)

Seismic Monitoring Residual Heat Removal Radiation Monitoring Plant Ventilation Main Turbine and Generator TOTAL 200 4

4

n____-----_ _ - _ - _ _ _ _ _

6.5 Administrative Controls Subject Contact Hours Responsibilities for Safe Operation and Shutdown Equipment Outages and Clearance Procedures Use of Procedures Plant Modifications Shif t Relief Turnover and Manning Containment Access l{aintaining Cognizance of Plant Status Physical Security Control Room Access Duties and Responsibilities of the STA Radiological Emergency Plan Code of Federal Regulations (appropriate sections)

Plant Technical Specifications (including bases)

Radiological Control Instructions TOTAL 80 6.6 General Operating Procedures Subject Contact Hours Startup At Power Operations Shutdown Xenon Following While on Standby ECP and S.D. Margin Calculation TOTAL 30 6.7 Transient / Accident Analysis and Emergency Procedures Subject Contact

• Hours Transient and Accident Analyses Plant Abnormal and Emergency Procedures TOTAL 30 l

6.8 Simulator Training The plant evolutions , transients and events listed below shall be conducted along with 'any others deemed necessary. The primary objective should be to demonstrate plant and operator response to a given condition or event and not necessarily to develop the control

. manipulation expertise 'of the trainee. The trainee / instructor ratio should not exceed 4:1.

Simulator exercises should be preceeded by a period of discussion of the planned exercises addressing expected response of the plant and applicable plant procedures to be u sed . Approximately 100 contact hours are required with about 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> in the classroom and 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> on the simulator.

Following each exercise demonstrating a transient or emergency event, an incident critique discussion should be held ta enhance the trainees' understanding of that particular exercise. When the simulator is not plant-s peci fic , the training shall be tailored to

the specific plant as much as practical.

t Simulator Exercises Reactor and Plant Startup Load Changes at Power Shutdown to Cold Condition Demonstration of Steam Generator Level Manual Control Load Rejections of Greater than 10%

. Failure of Rod Control System Failure of Automatic Steam Generator Level Controls Failure of Pressurizer Level and Pressure Automatic Controls l Turbine Trip from Full Power Reactor Trip from Full Power Loss of Normal Feedwater at Full Power Failure Open of Power Operated Relief Valve Stuck Open Pressurizer Safety Valve Loss of Reactor Coolant Pumps at Full Power and Demonstra-tion of Natural Circulation Failure Open of One or More Turbine Bypass Valves While at (a) Full Power, (b) Hot Standby I

Loss of All Feedwater (normal and emergency)

Loss of Reactor Coolant (small and'DBA)

Steam Generator Tube Rupture (small and large)

Loss of RHR Shutdown Cooling with the RCS Temperature 200*

l to 300*F Inadvertent Safety Injection While at Power Loss of Offsite Electrical Power Loss of One Train of Onsite Electrical Power

6.9 Annual Requalification Training Subject Hours Req'd.

Review of transient and accident analyses of FSAR condition III and IV events emphasizing the individual's role in accident as se s sme nt . Review selected industry events and LER's that could have led to more serious incidents. 40 (lecture)

Simulator exercises related to the tran-sients in Section 6.8 conducted so as to emphasize the role of the STA. g (simulator)

TOTAL 80 i

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!- ATTACHMENT 12 Response to Outstanding Issue of.the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 13.2.2: Fire Protection Training The applicant has . established a fire protection training program to ensure that the capability to fight potential fires ,is maintained.

However, the applicant has not provided the details. of the program for us to review. We require the applicant to provide a fire protection program which will fully comply with the guidelines in SRP Section 13.2.2 and BTP CIEB ' 9.5-1. We will review the program when it is-received and report our findings in the final SER.

Response

As specified in 10CFR50, - Appendix R, and SRP 13.2.2, the fire brigade training program ensures that the capability to fight potential fires is

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esteblished and maintained. The program consists of an initial class-room instruction program followed by periodic classroom ins t ruc tion, fire fighting practice, and fire drills:

1. Instruction

a. The initial classroom instruction shall include:

1) Indoctrination of the plant fire fighting plan with specific identification of each individual's responsibilities.

2) Identification of the type and location of fire hazards and associated types of fires that could occur in the plant.

3) The toxic and- corrosive characteristics of expected products of combustion.

4) Identification of the location of fire fighting equipment for each fire area and familiarization with the layout of the plant including access and egress routes to each area.

5) The proper use of available fire fighting equipment and the correct method of fighting each type of fire. The types of fires covered should include fires in energized electrical equipment, fires in cables and cable trays, hydrogen fires, fires involving flammable and combustible liquids or hazardous process chemicals, fires resulting from construction or modifi-cations (welding), and record file fires.

6) The proper use of communication, lighting, ventilation, and emergency breathing equipment.

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7) The proper method for fighting fires inside buildings and confined spaces.

8) The direction and coordination of the fire fighting activities (fire brigade leaders only).

9) Detailed review of fire fighting strategies and procedures.

10) Review of the -latest plant modifications - and corresponding changes in fire fighting plans.

NOTE: Items (9) and (10) may be deleted from the training of no more than two of the non-operations personnel who may be assigned to' the fire brigade.

b. The instruction shall be provided by qualified individuals who are

' knowledgeab le , expe rienced , and suitably trained in fighting the types of fires that could occur in the plant and in using the

, types of equipment available in the nuclear power plant.

c. Instruction shall be provided to all fire brigade members and fire brigade leaders.

, d. Regular planned meetings shall be held at least every 3 months for all brigade members to review changes in the fire protection program and other subjects as necessary.

e. Periodic refresher training sessions shall be held to repeat the classroom instruction program for all brigade members over a two-

year pe riod. These sessions may be concurrent with the regular planned meetings, i

2. Practice Practice sessions shall be held fo r each shif t fire brigade on the proper method of fighting the various types of fires that could occur in a nuc le ar power plant. These sessions shall provide brigade.

members with experience in actual fire extinguishment and the - use of emergency breathing apparatus under strenuous conditions encountered in fire fighting. These practice sessions shall be provided at least once per year for each fire brigade member.

3. Drills f a. Fire brigade drills shall be performed in the plant so that the fire brigade can practice as a team, b.. Drills shall be performed at regular intervals not to exceed 'l months for each shift fire brigade. Each fire' brigade membcr l should participate in each drill, but must participate ~ in at least two drills per year.

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A . suf ficient- number of these ' drills, but- not less than one' for each shif t fire . brigade per year, shall be imannounced to de ter-mine the fire fighting i readiness 'of the plant fire brigade, brigade le ade r , and fire protection systems and equipment.-

Persons planning and authorizing an unannounced drill shall ensure that ' the responding shif t fire brigade members . are not . sware that a drill is beirg planned until it is begun. Unannounced drills shall not be scheduled closer than four weeks.

At le as t one drill per year shall be pe rformed on a "back shif t" .

for each shift fire brigade, f

c. The drills shall . be preplanned to es tablish the training 'objec-tives of the drill and shall be critiqued to determine how well the training objectives have been met. Unanounced drills shall be planned and critiqued by members of the isanagement staf f respon-sible for plant safety and fire protection. Performance deficien-cies of -a fire brigade or of individual fire brigade meisbes shall be remedied by scheduling additional training for the brigade or membe rs . Unsatisfactory drill perfermance shall be followed by a -

i repest drill within 30 days.

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d. At - 3 year intervale, a randomly selected unannounced. drill shall be critiqued by qualified individuals independent of the licen-I see's . staf f. A copy of the written report from such individuals shall be available for NRC review.

. e. Drills shall as a minimum include the following:

1) Assessment of fire alarm ef fectiveness, time required to notify and assemble fire brigade , 'and selection, placement and use of -

equipment, and fire fighting strategies.

2) Assessment of each brigade membe r's knowledge of his or her

. role in the fire fighting strategy for the area ' assumed to con-tain the fire. Assessment of the brigade member's conformance with established plant fire fighting procedures and use of fire

fighting equipment, including self-contained emergency breath-

} ing apparatus, communication equipment, and ventilation equip-ment, to the extent practicable.

3)'The simulated use of fire fighting equipment required to cope with the situation and ty pe - of fire selected for the drill.

The area and type of fire chose n for the drill should dif fer f rom those used in the previous drill so that brigade members are trained in fighting fires in various plant areas. The

situation selected should simulate the size and arrangement of a fire that could reasonably occur in the area -selected, l allowing for fire development due to the time required to respond, to- obtain eq uipme nt , and organize for the fire, j assuming loss of automatic suppression capability.

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4) Assessment of brigade leader's direction of the fire figh t-ing effort as to-thoroughness, accuracy, and effectiveness.

4. Records Individual . records of training provided to each fire brigade member, including drill critiques, shall be maintained for at least~3 years to ensure that each . member receives training . in all parts of the~ training program. These records of training.shall be available ' for NRC review. Retraining or broadened training for fire fighting within buildings shall be scheduled for all those brigade members whose performance records show' deficiencies.

Training is .also provided to satisfy additional guidelines of CEB 9.5-1 Paragraph C.3.d, Items (k) and (1) which are in excess of Appendix R and SRP 13.2.2. Local . - fire companies are- invited to attend the training.

program. Although this training is primarily of fered to the designated immediate response units, representatives from other units participating in the Mutual Aid Plan may also be invited to participate. The program-covers the following topics:

1) Interface with the Site Security Force during emergencies.

2) Basic health physics indoctrination and training.

3) Beaver Valley Power Station facility layout.

4) Onsite Fire Protection equipment (permanent and portable).

5) Dif ferences between onsite fire fighting equipment and fire company supplied equipment.

6) Communications Systems.

7) Review of the appropriate sections of the Beaver Valley Power

[ Station Emergency Preparedness Plan and Implementing Procedures.

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8) The onsite emergency organization with speci fic emphasis on the interface between the Beaver Valley Power Station emergency squad and the fire company personnel, i

3 Training related to fire protect ion is also provided to other station employees as part of their initi al Station Orientation Training and

. periodic General Employee Refresher Training (CERT) . These training sessions include the following subject areas:

1) Station Orientation ,

a. Fire Chemistry, Parts of Fire, Extinguishing of Fire

b. Types of Fires

c. Methods of Extinguishing

d. Use of Extinguishers

! e. Misuse of Equipment t

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f. Fi,re Doors and Fire Penetrations g'. Procedure for Reporting or Fighting Fires

h. Industrial Safety'CO2 and Halon Systems

2) GERT Module I

a. Peporting a Fire

b. Fire Doors

c. Fire Barriers and-Pipe Penetrations

d. Fire Preventior.

Training related to evacuation of outlying buildings is presented in the Fire Marshall Training Program. Designated fire marshalls receive l training in such other areas as firs taid , fire protect ion, and fire  !

protection systems within their areas of responsibility.' j l

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