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REGULATORY FORMATION DISTRIBUTION SY M ('RIDS)
At;CESSION NBR: 8204190066 DOC ~ DATE: 82/04/06 NOTARIZED:, NO FACIL,:50 244 Robert Emmet Ginna Nuclear Planti Uni<<t li Rochester G
'AUTH BYNAME AUTHOR AFFILIATION MAIERiJ ~ E ~
.Rochester Gas 8 <<Electric Corp'E<<C I P ~ NAME REC IP IENT A F F IL'I ATION ORUTCHFIELD~D ~
Operating Reactors Branch 5
DOCKET 05000244
SUBJECT:
Forwards clarification of util responses re NUREG 0737,Items I ~ A ~ 2 ~ 1~ "Upgrading Reactor Operator 8 Senior Reactor Operator Training Qualifications"-
8 Item II'E 4i
.Tr aining to Mitigate Core, Damage DISTRIBUTION CODE:
A046S COPIES:RECEIVED:LTR ENCL J SIZE,-
TITLE: Response to NUREG -0737/NUREG"0660
=TMI Action Plan Rgmts (OL's)
NOTES:NRR/DL/SEP 1cy,.
05000244 RECIPIENT ID CODE/NAME ORB 05 BC 01 INTERNAL; ELD IE/DEP D IR 33 IE/DEP/EPLB NRR/DE/ADCSE 22 NRR/DE/ADSA 17 NRR/DHFS/DEPY29 NRR/DL/ADL 16 NRR/DL/ORAB 18 NRR/DSI DIR 24 NRR/DS I/ADPS 25 NRR/DS I/AEB NRR/DS I/RAB NRR/DST/ADT 32 RGN1
'EXTERNAL: ACRS 34 INPOiJ ~ STARNES NRC PDR 02 NTIS
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ROCHESTER GAS AND ELECTRIC CORPORATION
~ 89 EAST AVENUE, ROCHESTER, N.Y. 14649 STATE JOHN E. MAILER VKO Pfosldo$ $$
TC$.CPHONC ARCA CODE T$ $$ 546.2700 April 6, 198 g
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4$$$, e'a Divector of Nuclear Reactor Regulation Attention:
Mr. Dennis Crutchfield, Chief Operating Reactor Branch ¹5 United States Nuclear Regulator Commission Washington, D.C.
20555 RE:
Mr. Crutchfield letter of March 23, 1982 concernin TMI Action Plan Items I.A.2.1 and II.B.4.
Deav Mv. Cvutchfield:
Rochester Gas
& Electric Corpovation responded to the Harold Denton letter of Mavch 28, 1980 and NUREG 0737 item I.A.2.1 and item II.B.4 through several correspondences.
The Ginna Training Pvogvams weve revised to address the recommen-
- dation, but our training programs do not reference contact hours on individual topics.
Both the above documents did not reference contact houvs for specific topics, and, therefore, our response addressed the recommendation of the NRC.
Each question will be individually addressed in an effort to clarify what was included in pvevious training velevant to your request.
Sincerely, Vice President Electric and Steam Production XC:
R.
Mor rill B.
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PAGE 1
1.
Attachment A associated with Che August 25, 1980 letter claims that Administrative Proceduve A102.13, R.E.
Ginna NRC Licensing Tvaining Program teaches the subjects of heat transfer, fluid flow, thermodynamics and Che use of installed plant system Co mitigate an accident in which'he core is sevevely damaged.
Is Che level of instruc-tion compavable to Chat; detailed in enclosuves 2 and 3 of Denton's Mavch 28, 1980, letter?
Do these subjects involve 80 contact hours?
(A contact hour is a one-hour period in which Che course instructor is pvesent or available for instvucting ov assisting students;
- lectures, seminars, discussions, problem-solving sessions, and examinations are considered contact periods under this definition.)
The level of instruction for the topics was comparable with those detailed in Mr. Denton's March 28, 1980 letter.
The Table of Contents for Che topic of Introduction to Physics, Thermodynamics, Fluid and Fluid Flow Principles, and Mitiga-ting Core Damage are attached as part of'uestion 86.
The training program does not address 80 contact hours.
The mater1al was covered as part of License Training and Requal-ification 1n Training at Che noted times.
SPRING 1982 LICENSING CLASS Heat Transfer and Fluid Flow with Pertinent NRC Type Exam Questions January/February 1982 10 Days Mitigating Core Damage as part of training before starting License Training June August 1982 4 Days SUMMER 1981 LICENSING CLASS Mit1gating Core Damage June 1981 Heat Transfer March 1981 4 Days 6 Days
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PAGE 2
WINTER 1980 LICENSING CLASS Plant Transient and Emergency Procedures involving Radiation Monitoring, Containment Isolation and specif'ic Emergency Procedure Responses July 1980 t
Introduction to Thermodynamics September 1980 I
Natural Circulation~and Heat Transf'er
'ust prior to beginning L'icense Training April 1980 5 Days 4 Days 1 1/2 Days
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PAGE 3
2.
Are the lectures and quizzes on the subject of accident mitigation given to shift technical advisors and opevat-ing personnel from the plant manager through the operations chain to the licensed operators?
Xf they are, would you please provide the titles of the people who are trained and an ovganization chart which illustvates their position in the operations chain'?
Yes, all of mentioned personnel participated in Mitigating Core Damage during the Summer of 1981.
Superintendent Assistant Superintendent Operations Engineer Operations Supervisor Shift Supervisor Head Control Operator Control Operator Technical Assistant for Operational Assessment Shift Technical Advisor
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Does the tvaining progvam described in Administrative Pr ocedure A102.13 have an increased emphasis on dealing with veactor and plant transients as called for in enclosure 1 of Denton's March 28, 1980 letter'? lf Chere is, does Che program addvess both normal and abnormal (accident)
Cvansients'?
- Yes, Che program presently addresses additional time devoted to plant transients with the most significant changes being simulator use.
Prior to 1980,,the simulator was used only for startup certification.
Presently, we have added two additional weeks Co the program for manipulations under normal and transient conditions.
The overall progvam has been increased in scope and length to add more material dealing with plant responses.
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PAGE 5
I 4.
Does the requalification program which addresses heat transfer; fluid flow, thermodynamics and the use of installed systems for accident mitigation involve 80 contact hours?,
I The program does not address 80 contact hours.
- However, the below topics address the Heat Transfer, Fluid Flow, Thermo-dynamics and the use of installed systems for Accident Mitigation.
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Annual Exam and Review Involves three days of review and a one day exam where above topics are two of the eight areas.
"'1980' 1981 l day 1982%
l day 2.
Mitigating Core Damage.
See attached index in-cluded in question ¹6 4 days 3.
Introduction to Physics, Thermodynamics, Flu1d and Fluid Flow Principles See attached index in-cluded in question ¹6 Heat Transfer Intro-duction.
See attached 1ndex included in question ¹6 3 days 3 days 2 days 5.
Accident Analysis involving Emergency Procedures.
To address topics such as Inadequate Core Cooling and Natural Circulation 4 days
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6.
Radiation Monitoring System/Containment Isolation System.
Addressing the use of RMS to evaluate and mitigate accident and function of Containment Isolation to controlling release 7.
Sping Radiation Monitor.
Addressing the new installed high range monitor for evaluation and pro)ection of doses 1980' day
'1982<<
<<Actual days completed prior to April 1, 1982.
Additional training shall be provided during 1982.
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PAGE 7
5.
Administrative procedure A102.14 lists in Section 2.1 control manipulations which are part of-the requalifica-tion program.
Manipulation 22 is titled "Loss of Instrument Bus."
Does this include "Loss of Protective System Channel" which is item (17) of enclosure 4 of Denton's March 28, 1980 letter' At Ginna Station we have four instrument buses.
Each feed from a separate source-These buses each feed one reactor protection channel Therefore, loss of an instrument bus is equivalent to loss of a protective system channel.
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PAGE 8
6.
For item II.B.4 provide an outline of the training program for Mitigating Core Damage, including the number of training hours involved.
Your outline can include any training program which r'elates to the training for Miti-gating Core Damage.
Follow the guidelines given in the enclosure 3 of H.
R. Denton's letter dated March 28, 1980 and INFO Guidelines for Training to Recognize and Miti-gate the Consequences of Core Damage (Document Number STG-01, Rev.
1, January 15, 1981).
NRC requires minimum of 80 contact hours of training for Mitigating Core Damage.
Attached are the lesson outlines used to address the topics relevant to Mitigating Core Damage.
Our programs do not address
- hours, but in question number 4 included are the days of instruction for each topic during the past two years.
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WESTINGHOUSE MITIGATINGCORE DAMAGE
TABLE OF CONTENTS
~To ic Tab Introduction Core Cooling Mechanics Potentially Damaging Situations Small Break LOCA's - No High Head Safety Injection 3
Potential ly Damaging Si tua tions Loss of Feedwater Induced Loss of Coolant Accidents Vital Process Instrumentation Recognizing Core Damage - Incore Instrumentation 6
Response of Excore'nstrumentation Post-Accident Primary Radiochemistry Radiological Aspects of Core Damage
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CORE COOLING MECHANICS
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Objective Overview Introduction Mechanisms of Heat Transfer Definitions Steady State Heat Conduction Steady State Heat Transfer in a Slab 27 Steady State Heat Transfer in a Cylinder 29 Convection 32 Nucleate Boiling 36 Thermal Limits 39 Hydrogen Generation Natural Circulation 42 47 Boron Precipitation 55 Non-Condensable Gas Formation
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SIPER IHTEIEIENT TRAINING COCRD INATIR ASSISTANT SIP KRIHtKIOEHT TECHNICAL ASSISTANT FOR OFEYATIOtNL ASSKSSisKNT.
ENGINEERS SVPERY ISal CH0II5 TRY AIO HEAL PHYSICS CPERAT IOH5 ENGINEER SRO F IRE PROTECTION AI0 SAFETY COORDe Hh INfENhNX ENGINEER OfF ICE SIP ERYISCR TECUMICAL ENGINEER OVALITT COelTRL EHGIHKER
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CHEN I 'St HEALTH PHYSICIST5 OPERAT IOHS
'IPERY I SOR SRO HATUlIALS COORO IHATOR IVIIHTEHAICE SIPERYI Sal ISO SIPERYISOR HVCLEAR ENGINEER RESVLtS 4 TEST SIPERYISal ENGINEER(5)
~4 IM CHKHISTRY TKagiICIANS RADlhfION PROTECt I ON TEasHICIANS SHlft TEaiHlgC,
/OYISCR I/SNlfT SillfT SIPERYI SOR I/SHIF'f r SAO I
HKAD QXITROL OPERATOR I/SH IF T AO COITRQ. OPERATOR I/SHIFT AO 5toaBCCH HAINTEHAICE FCREHEN F ITlEBS ma ANICS HAIOYIN:N llC ANO ELECTRIC FORENEN TEagIICIAHS 4 REPA IIVIKH ELKCTAICIANS TECHH IC IANS AW(ILIARY OPERATOR 2/SHIFT RKPORT IHG
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- STA not required during ref e11 ugonr co1d shutdown modes.
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POTENTIALLY DAMAGING SiTUATIONS,.
SMALL BREAK LOCA'S -
NO HIGH HEAD SAFETY ItMECTION
~To ic Pa<ac Objectives Overview Description of Loss of Coolant Accidents
- Minimum Safeguards Core Cooling. Mechanics - Natural Circulation 27 Small Break LOCA's - No High Head Injection
- One Inch Cold Leg Breaks
- Four Inch Cold Leg Breaks J
37 37 41 ICC Instructions 45 Summary 59 References 59 Self Assessment 60
'Tables and Figures 61
POTENTIALLY DAMAGING SITUATIONS LOSS OF FEEDWATER -
INDUCED LOSS OF COOLANT ACCIDENTS
~TO 1C Pa(ac Objectives Overview Loss of Feedwater
- Accident, Loss of Feedwater Induced LOCA's Zero Break
- Auxiliary Feed Initiation
- Bleed and Feed-SI Actuation
- Bleed and Feed-Delayed SI Actuation
- Feed and Bleed 9
11 13 26 Loss of Feedwater Coincident with Small LOCA 28 ICC Instructions 31 Summary
,40 References 40.
Self Assessment 41 Tables and Figures 43
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VljtOPROCESS 1HSTRUMEjlTRTION
~To ic Objectives
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Overview Temperature Measurement 4'ressure Measurement Level Measurement 26 Flow Measurement 30.
Summary 33 References 33'elf Assessment 35 Tables and Figures 36
R'ECOGNI ZING CORE DAMAGE IN-CORE INSTRUMENTATION
~TO 1C
~algae Objectives Overview Movable In-Core System Hardware Description Ability of Movable Detectors to Sense Gamma Levels 6
Low Level Detector Setup Movable Detector Surveillance in an Accident Condition Thermocouples f
Lessons Learned from TMI-2 12 17 Summa y 18 References
'19 Self Assessment 20 Tables and Figures 21 SS/3803 B/ D161B
EXCORE INSTRUMENTATION TO POST ACCIDENT CONDITIONS
~TO 1C
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Objectives
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0 Normal Response to Reactor Trip Source Range
Response
During Accident Conditions Core Voiding Effects on Reactor Kinetics Non-Homogeneous Effects 7
9 ll TMI-2 Excore Response 17 Recriticality Analysis Summary References 21 23 25 Self-Assessment 26 Tables and Figures 28
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0 POST-'ACCI DENT PRIMARY RADIOCHEMISTRY
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Objective Overview Baseline Data and Assumptions In-core Release Mechanism Rod Burst Effects on Coolant Radiochemistry Mechanism for Extensive Core Damage/Radiochemistry Effects 12 Radiological Hazards of Sampling 16 Summary 20 References 21 Self-Assessment'.2 Tables and Figures 24
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RADIOLOGICAL ASPECTS OF CORE DAMAGE
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Obj ectives 3
Overview'ostulated Plant Accident Concentrations of Fission Products in Containment Affects on Area Monitors Contamination Levels in Containment 12 Estimate of In-Containment tteasurement from External Readings Effects of Environmental Release 15 Emergency Action Level Guidelines 19 Summary 22 C
References Self Assessment 22 24 Tables 28 G
SS/3733B/DI50B
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THERMODYNAMICS~
FLUID AND FLUID FLOW PRINCIPLES Ginna 1980 Revision 3
06/25/81
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TABLE OF 'ONTENTS Mass, Acceleration, Force Weight Pressure/Pressure Due to a Height of Water Work/Energy
<PE KE)
Pressure Work.
Internal Energy
.Power Temperature/Specific Heat/Enthalpy Heat Capacity Enthalpy The Relationship Between q and h
Total System Energy (G.E.E.)
Density/Specific Volume/Specific Weight Temperature and Pressure Effects on Water Buoyancy Fluid Friction (Vicsosity)
Laminar/Turbulent Flow/Re Mass Flow Rate Flow Measurements Pressure Drops in a Flowing System Bernoullis Equation with.Losses The G.E.E.
(Studied Further)
Example Problems Using the'G.E.E.
Heat Transfer/Conduction Convection Forced Convection Two-Phase Flow Void Fraction/Steam Quality Radiation Heat Transfer Specific Heat Transfer Related to Nuclear Power Heat Exchangers Turbine Cycles/T S Diagram/The Carnot Cycle 3
4 6
7 9
10 11 12 13.,
14 15 18 20 21 22 23 24 25 26 17/28 29 31 35 36 37 38 43 44 45
. 47 52 Plants
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TABLE. OF CONTENTS Mollier Diagram (h s)
The Unavailability of Energy (Enthropy)
The Rankine Cycle The Turbine Turbine Efficiency The Condenser The Pump The Boiler The P V Diagram Rankine Cycle Efficiency The Reheat-Regenerative Cycle Nuclear Technology Reactor Heat Transfer Limits Westinghouse Handout on Hot Channel Factors 54 56 58 59 61 62 63 64 65 66 67 69 84
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OPERATOR REQUALIFICATION TRAINING 3/25/80 to 5/30/80 Heat Transer Introduction The following topics were covered:
Definitions of Terms Types of'Heat Transfer Types of Boiling Reynolds
- Number, Lamerian and Turbulent Flow Pipe Friction Overall Heat Transfer Coefficient Heat Flow from a Fuel Rod Q = cpmhT Q = UAAT Gal erome tric DNB, Burnout, Flow Instability Hot Channel Factors Quad to Average Power Tilt
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0 8/29/80 Radiation Monitoring System Containment Isolation
'Lecture includes the following topics and procedures:
Hydrogen Accumulation in Containment Principles of Operation G.M.
& Scintillation E-16.1, High Activity RMS E-16.2, High Iodine in Plant Vent E-28, RCS Leak SC-1.2, Local Radiation Emergency SC-1.6, Release to Lake T-35A, Ventilation System Startup a Shutdown Technical Specification Changes 33 and 34
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