ML20052E558
| ML20052E558 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/06/1982 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.2.1, TASK-2.B.4, TASK-TM NUDOCS 8205110230 | |
| Download: ML20052E558 (13) | |
Text
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e TENNESSEE VALLEY AUTHORITY CH ATTANOOG A, TENNESSEE 37401 400 Chestnut Street Tower II May 6, 1982 e,
Director of Licensing 4
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Attention: Mr. Domenic B. Vassallo, Chief s
Operating Beactors Branch No. 2 c'
NO 3
U.S. Nuclear Regulatory Conmission lO[3d
.,,,II.9I332 " Q Washington, DC 20555 i
g t r:-
Dear Mr. Vassallo:
' '. N y
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w In the Matter of the
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Docket Nos. 50-259
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'knnessee Valley Authority
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50-260
'N 50-296 Your letter to H. G. Parris dated April 2,1982 requested that TVA provide additional information regarding our response to NURBG-0737 items I.A.2.1 and II.B.4 for the Browns Ferry Nuclear Plant.
Enclosed is our response to this request.
A copy of this response is being sent to R. T. Liner of Science Applications, Inc., as you requested.
Very truly yours, ES VALLEY AUTIORITY L.
. Mills, er Nuclear Licensing Subscribed sworn to be me this //> _-
day of
/-
1982.
Y b
Notary Public
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My Conmission Expires Enclosure cc: See page 2
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8205110230 820506 PDR ADOCK 05000259 P
PDR An Equal Opportunity Employer
. Mr. Damenic B. Vassallo May 6, 1982 cc:
U.S. Nuclear Regulatory Conmission Region II ATIN: James P. O'Reilly, Regional Adninistrator 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. R. J. Clark (Enclosure)
U.S. Nuclear Regulatory Conunission Browns Ferry Project Manager 7920 Norfolk Avenue Bethesda, Marylard 20014 Dr. R. T. Liner (Enclosure)
Science Applications, Inc.
1710 Goodridge Drive McIean, Virginia 22102 l
4 ENCLOSURE RESPONSE TO D. B. VASSALLO'S LETTER TO H. G. PARRIS DATED APRIL 2, 1982 NUREG-0737 ITEMS I.A.2.1 AND II.B.4 BROWNS FERRY NUCLEAR PLANT Question 1.
The enclosure from your November 10, 1980, submittal includes lectures which appear to have the potential for covering the subjects of heat transfer, fluid flow and thermodynamics for training and requalifications as called out in enclosure 1 of i
Denton's March 28, 1980, letter. Do these lectures in fact cover this material and is the coverage at the level of detail I
siecified in enclosure 2 of the Denton letter?
Response: The TVA cold and hot license and requalification programs provide training in heat transfer, fluid flow, and thermodynamics. Attachments 1 and 2 outline the training provided for TVA operations employees.
Question 2.
The enclosure from your November 10, 1980, submittal includes lectures which appear to have the potential for addressing the subject of using installed plant systems to control or mitigate an accident in which the core is severely damaged.
This requirement is called out in enclosure 1 of Denton's l
Do these lectures address the topic at the level of letter.
detail specified in enclosure 3 of Denton's letter?
Response: The TVA cold and hot license and requalification programs include training in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged. The training outline that covers the areas in enclosure 3 of Mr. Denton's letter is shown in attachment 3.
Question 3 Are the lectures and quizzes on the subject of accident mitigation given to shift technical advisors and operating personnel from the plant manager through the operation chain to the licensed operators? If they are, would you please provide the titles of the people who are trained and an organization chart which illustrates their position in the operations chain?
Response: Accident mitigation training is given to shift technical advi9 ors and operations employees from the operations super-visor to the licensed operators to comply with enclosure 3 of Mr.
Denton's March 28, 1980 letter. An abbreviated program has been given to managers and technicians in the Health Physics, Plant Chemistry / Radiochemistry, and Instrumentation and Controls sections commensurate with their responsibilities in the event of a core damaging accident. Attachment 4 (Browns Ferry organizational chart) shows the employees trained in accident mitigation.
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Question 4.
The enclosure from your November 10, 1980, submittal claims to cover all the manipulations as called out in enclosure 4 of Denton's letter. Does the requalification program in fact cover all these manipulations and are these manipulations performed within periods specified in enclosure 4 of Denton's letter?
Response: All control manipulations mentioned in enclosure 4 of Denton's letter are performed by requalifying operators at the required frequency (annually or on a 2-year cycle). These manipu-lations are shown in attachment 5.
i Question 5.
Do the training and the requalification program elements which
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include heat transfer, fluid flow, thermodynamics and accident mitigation use 80 contact hours? (A contact hour of instruction is a one-hour period in which the course instructor is present or available for instructing or l
assisting students; lectures, seminars, discussions, problem-solving sessions, and examinations are cor.sidered contact periods under this definition.)
Response: Browns Ferry operator requalification for 1980 and 1981 includes a total of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> of heat transfer, fluid flow, and thermodynamics (attachment 1).
Additionally, the operator initial training program (NOTP) has been updated to include the requirements of Denton's March 28, 1980 letter.
Question. 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 relates to the training for mitigating core damage.
Follow the guidelines given in enclosure 3 of H. R. Denton's letter dated March 28, 1980 and INPO Guidelines for Training to Recognize and Mitigate the Consequences of Core Damage (Document Number STG-01, Rev. 1, January 15, 1981). NRC requires a minimum of 80 contact hours of training for mitigating core damage.
Response: The mitigating core damage training outline is shown in attachment 3.
This pecgram consists of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of specific mitigation core damage training. Additionally, mitigating core damage training elements are covered in the Nuclear Operator Training Program, Pre-License Training, and Requali-fication Training for TVA licensed operators. TVA believes the total combined hours of training meet or exceed the recommended 80 contact hours and is sufficient to satisfy the intent of the NRC letter and INPO recommendations.
RFC:COH 4/30/82
I Listing of Ehclosed Attachments - BNR Requalification Training Table of Contents - BNR Requalification Training 'Ihermodynamics 'Ihermal Limits - BNR Training for Mitigating Core Damage - Browns Ferry Ebnctional Organization Chart - Required Reactivity Control Manipulations - Browns Ferry Simulator l
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ATTACHMENT 1 BWR REQUALIFICATION TRAINING TABLE OF CONTENTS Page Number Unit 1 - The Steam Power Cycle 1.1-1 1.1 Th e Ba sic Cycle...................,
1.2-1 1.2 Temperature, Pressure and Volume............
1.2-1 1.2.1 Temperature..............
1.2-3 1.2.2 Pressure................
1,2-11 1.2.3 Volum e.................
1.2-11 1.2.4 Universal Gas Law...........
1.3-1 1.3 Heat and Its Effects..................
/1.4-1 1.4 Enthalpy and Entropy..................
1.4-1 1.4.1 ^ '
Work and Energy............
1.4-2 1.4.2 Enth alpy................
1.4-4 1.4.3 Entopy.................
1.5-1 1.5 Cycle Diagram..................... 1.6-1 1.6 Steam Table s......................
Unit 2 - Thermodynamics: Heat at Work _
The First Law of Thermodynamics: Potential and Kinetic 2.1 2.1-1 En e r g y.........................
2.1.1 The First Law of Thermodynamics....
2.1-1 2.1-3 2.1.2 Potential Energy (PE)..........
2.1-4 2.1.3 Kinetic Energy (KE)...........
2.1-4 2.1.4 Water Hammer.............
2.2-1 Internal Energy, Flow Work, Mechanical Work and Heat 2.2 2.2-1 2.2.1 Internal Energy (U)...........
2.2-1 2.2.2 Flow Work (Pv).............
2.2-3 2.2.3 Mechanical Work (W)..........
2.2-3 2.2.4 Heat (Q)................
2.2-4 2.2.5 Use of the Energy Equation.......
4 9
s TABLE Or CONTENTS (Continued)
Page Number
_2.3-1 2.3 Energy Conversion.
2.4-1 2.4 The Second Law and Efficiency 2._5'- 1 2.5 Vapor Compression Refrigeration Cycle Unit 3 - Steam Boilers 3.1-1 3.1 Basic Heat Transfer Principles 3.1-1 3.1.1 Conduction.
3.1-2 3.1.2 Convection.
3.1-3 3.1.3 Radiation 3.1.4 Heat Transfer in the Plant 3.1-3 3.2-1 3.2 Physical Parameters of Basic Heat Transfer 3.2-1 3.2.1 Temperature Difference 3.2-3 3.2.2 Area 4
3.2-3 3.2.3 Material.
3.2-4 3.2.4 Flow 3.3-1 3.3 Boiling Heat Transfer.
3.4-1 3.4 Physical Parameters of Boiling Heat Transfer 3.4-1 3.4.1 Pressure.
3.4-3 3.4.2 Temperature 3.4-3 3.4.3 Flow 3.5-1 3.5 Steam Boiler Characteristics 3.5.1 Water Circulation 3.5 I 3.5-2 3.5.2 Steam 3.5-4 3.5.3 Level Changes.
Unit 4 - Turbine Generator 4.1-1 4.1 Turbine Cycle.
4.2-1 4.2 Energy Conversion 4.2.1 Critical Pressure Ratio.
4.2-1 4.2-2 4.2.2 Orifices
.y TABLE OF CONTENTS (Continued)
Pace Number 4.2-6 4.2.3 Types of Turbinea.
4.3-1 4.3 Superheat and Reheat Cycles 4.3-1 4.3.1 Superheat Cycle
" 4.3-4 4.3.2 Reheat Cycle 4.4-1 4.4 Turbine Precautions Unit 5 - Condenser 5.1-1 5.1 Condenser Theory 5.1-1 5.1.1 Purpose of the Condenser.
5.1-2 5.1.2 Parts of the Condenser.
-5.1-4 5.1.3 Condenser Operation 5.2-1 5.2 Condensers and Cycle Efficiency 5.3-1 5.3 Improving Condenser Efficiencies 5.3-1 5.3.1 Condenser Efficiency
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5.3-2 5.3.2 Condensate Depression 5.3-3 5.3.3 Condenser Design Features 5.3-4 5.3.4 Condenser Fouling 5.3-4 5.3.5 Air Binding 5.3-5 5.3.6 Air Leakage.
5.3.7 Other Effects on Condenser Vacuum 5.3-5 5.4-1 5.4 Turbine Extraction and Feedwater Heating 5.5-1 5.5 Condenser Cooling Systems Unit 6 - Pumps and Fluid Flow 6.1-1 6.1 Hydraulic Systems 6.2-1 6.2 Positive Displacement Pumps 6.3-1 6.3 Eductors and Jet Pumps 6.4-1 6.4 Contrifugal Pumps 6.4-1 6.4.1 Radial Flow Pumps 6.4-3 6.4.2 Axial Flow Pumps.
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4 TABLE OF CONTENTS (Continued)
Page Number 6.4-4 6.4.3 Mixed Flow Pumps............
6.4-5 6.4.4 Centrifugal Pump Precautions 6.4-6 6.4.5 Graphing Centrifugal Pump Operation...
6.5-1
.6.5 Net Positive Section Head.............., -
Ur.it 7 - Steam Plant Calculations 7.1-1 7.1 Steam Cycle Efficiency................
7.2-1 7.2 He a t Bala nce s.......................
7.3-1 7.3 Improving Cycle Efficiencies 7.4-1 7.4 Reducing Heat Waste.................
Unit 8 - Reactor Thermal and Hydraulic Performance 8.1-1 8.1 Performa nce Obj ectives................
8.2-1 8.2 Departure from Nucleate Boiling 8.3-1 8.3 Temperature and Pressure Limitations.........
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ATTACHMENT 2 BWR REQUALIFICATION TRAINING THERMODYNAMICS THERMAL LIMITS I.
Sta CORE IEERMAL LIMITS A.
Technical Specificatien Limits CMTLPD - Core Maximum Fraction of Limiting Power Density 1.
2.
"R" Factor _
3.
MCPR - Minimum Critical Power Ratio 4.
LHCR - Linear Heat Generation Rate MAPLHCR - Maximum Average Planar Linear Heat Generation 5.
Rate Non-Technical Q ecificatio'n Limits B.
1.
CMPF - Core Maximum Peaking Factor FCIOMR_- Preconditioning Interim Operating Management 2.
Recommendations II. PROCESS COMPUTER l
A.
Special Programs I
1.
LPRM Calibration (CD 1/2)
Cc=puter Outage Recovery Monitor (OD-15) 2.
Other On-Demand (OD) Programs OD-7, OD-8, OD-10, OD-11 CD-14, OD-16, and OD-17 3.
B.
Trequently Used Programs Thermal Data in a Specified Bundle (OD-6) l 1.
Core Thermal Power and APPJi Calibration (OD-3) 2.
Periodic Core Evaluation (P1) 3.
explanation of terms I
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ATTACHMENT 3 BWR TRAINING for MITIGATING CORE DAMACE A.
Incore Instrumentation 1.
Review of Neutron Monitoring System 2.
Use of NMS for determinatier. of void formation, void location basis for NMS response as a function of core temperatures and density changes.
Use of_NMS or TIP system to determine extent'of core 3.
damage and geometry changes.
B.
Vital Instrumentation 1.
Instrumentation response in an accident environment with particular emphasis on post accident monitors (PAM).
2.
Inderumentation to be discussed a.
NN3 b.
RMCS c.
RPIS d.
H2 and 02 analyzers e.
Reactor level 3.
Topics to be discussed Instrument failure mode during loss of instrument a.
power or other predictable failures (e.g., loss of reference leg).
b.
Effects on readings due to design transients of temperature, radiation, moisture, and pressure.
Expected degree of accuracy following parameter's c.
return to normal.
d.
Possible alternate means of determining approximate value for critical parameters assuming the primary method of measurement has failed.
Possible use and capability of plant computer in e.
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monitoring and analyzing critical parameters.
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Isolation philosophy, signals, instrumentation, and potential failure modes.
C.
Chemistry Results with Core Damage 1.
Initial detection a.
Chemistry parameters b.
Pretreatment monitor 2.
Determination of extent of damage Methods of sampling and analysis a.
(1) Gamma-ray isotopic (2) Gases b.
Time requirements for sampling and analysis c.
Indication of:
(1) Clad defects (2) Massive defects (3) Fuel melting l
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a.
d.
Additional sampling and analysis (1) Reactcr coolant (2) Containment 3.
Consequences of damage Activity levels in coolant and total activity in core a.
b.
Contamination considerations c.
Releases to containment d.
Releases to the environment Personnel exposure during sampling and analysis e.
4.
Corrosion effects a.
Zircaloy/ water reaction b.
Submersion of equipment and time to failure Emergency chemistry controls c.
D.
Radiation Monitoring 1.
Response of process and radiation monitors to severe core damage.
2.
Behavior of monitors and deteer. ors when saturated.
3.
Methods of detecting radiation readings by direct measurement at detector output signals.
4.
Expected accuracy of monitors at different locations.
5.
Use of munitors to determine extent of core damage.
6.
Radiation monitor failure modes.
7.
Methods of determining dose rate inside containment from measurements taken outside containment.
E.
Gas Problems Under Accident Conditions 1.
Hydrogen a.
Sources b.
Hazardous concentrations c.
Methods of measuring concentration d.
Venting e.
H2 recombiners
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2.
Oxygen a.
b.
Containment 3.
Other gases a.
Noncondensables--Xe, K, AR R
b.
Accumulation in containment c.
Venting and leakage
FUNCTIONAL ORGAKIZATION FOR BROWNS FERRY NUCLEAR PIANT i
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.~ ;. g ATTACHMENT 5 REQUIED EAi..auTY CCTTRCL MCIPL' ATIONS SW.M5 TDIT SIE" A!cR Trainee's Name Instructor's Signature Date Purpose of Evaluation NOTI:
1.
Star (*) itema shall be perfermed annually, all etner itess shall
,j be perf ormed on a two-year cycle.
1 I
2.
Personnel with senior license may be credited with these activities t
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f if they direct or evaluate control manipulations as they are performed.
E v.
v.
- 1)
- Reactor startvo and haatup f
- 2) Turbine /senerator startup
- 3) elC2 power change
- 4)
- Manual control of feedveter during startup or shutdown
- 3)
- Plant shutdown
- 6) Rosetor seres
- 7) LOCA *Inside dryvell Small break
- 0utside dryvell Small break
- Inside drvve11 1 arse break =
- 0utside drvuell tarse break Main steam line break
- 8) Fuel failure /high off-gas
- 9) Loss of one recire p ep
- 10)
- Loss of both recire pups
- 11) Malfunction of rectre speed control
- 12) Inabiltrv to drive control rods
- 13) Mieseettioned control rod (or rod drop)
- 16) Muclear inst a.tation failure
- 15) Loss of 175 bus
- 16) stC initiation
- 17) Tertine/senerator tri,
- 18) toes of electrical power or less of bus / busses l
- 19) pressure regulator malfunction f
- 20) feedvater svsten salfunction
- 21)
- Total loss of feedvatet (normal & emergenev)
- 12) toes of condenser vacuum
- 23) toes of 73CCW
&) Less of shutdown cooling system
- 15) Total loss of AC electrical power
- 26) Scram s* stem failure
- 17) Loss of insts w.t air
- 28) Loss of IIO: or ERR 5*'
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- 29) Other
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