ML20051L919

Forwards Addl Info Requested Re TMI Items 1.A.2.1 on Upgrading Senior Reactor Operator Training & Qualifications & II.B.4 on Training for Mitigating Core Damage

ML20051L919
Person / Time
Site:	Davis Besse
Issue date:	05/10/1982
From:	Crouse R TOLEDO EDISON CO.
To:	Stolz J Office of Nuclear Reactor Regulation
Shared Package
ML20051L952	List: ML20051L919 ML20051M340
References
RTR-NUREG-0660, RTR-NUREG-0737, RTR-NUREG-660, RTR-NUREG-737, TASK-1.A.2.1, TASK-2.B.4, TASK-TM 816, TAC-44152, TAC-44502, NUDOCS 8205170322
Download: ML20051L919 (20)

Text

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _.

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TOLEDO

% EDISON Docket No. 50-346 RcHmo P. CROUSE License No. NPF-3 gowne v

Serial No. 816 Y

May 10, 1982

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• 4}" Y /g 2A Mr. John F. Stolz, Chief Operating Reactors Branch #4 6'

h Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.

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Dear Mr. Stolz:

Af ter completing your initial review of our August 4,1980 submittal on information concerning post Three Mile Island action items set forth in NUREG 0660 and NUREG 0737 items:

1.a.2.1 Immediate Upgrading of Senior Reactor Operator Training and Qualifications and II.b.4 Training for Mitigating Core Damage, additional questions requiring our response were generated.

Attached is the additional information requested on your questions 1 through 6.

Also included as an enclosure for Item II.b.4 is an outline of the training program for the Mitigation of Core Damage and other related outlines of programs relating to the training on mitigation of core damage.

Very truly yours, f

0 RPC/JMH/bj s

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Enclosures:

1.

Operator Training - Mitigation of Core Damage 2.

Simulator Training Program Documentation 3.

Simulator Training for Mitigation of Core Damage cc: DB-1 NRC Resident Inspector THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652 820 517 0%L P

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ATTACHMENT Docket No. 50-346 Serial No. 816

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Page 1 of 4 License No. NPF-3 May 10, 1982 1.

The use of installed instrumentation was addressed in the core damage training by the following methods:

a.

Use of the Self-Powered Neutron Detector System (SPND) to determine that high temperatures exist within the reactor core. This is accomplished by observing the alarm summary status for the SPND System. The limitations of this method were also stressed to the operator in that the thermionic emission currents and space charge release currents are very small and only meant to be an indication that high temperatures occurred and not a means of determining actual core temperature, b.

The use of the core exit installed thermocouples was discussed in detail. This included normal methods of reading the incore thermo-couples via the plant computer and the use of the local readings of the thermocouple taken by the 1&C Department.

c.

Methods of interpreting out-of-core source range nuclear instrumen-tation readings to core uncovery and inadequate core cooling were described. This included a discussion of how coolant temperature, coolant level and coolant void distribution can be related to in-adequate core cooling using out-of-core nuclear instruments.

d.

Methods of determining effective core flowrate utilizing various heat balance procedures were provided. This enables the operator to calculate an expected core flow and compare it to installed indications.

e.

Void size calculations were provided to demonstrate how the operator can determine approximate sizes of steam or non-condensible gas bubbles. This will better indicate the actual coolant volume present in the system and whether the core is actually covered, f.

Failure modes of instruments were addressed, especially radio-logical monitoring instruments for the Containment and letdown systems.

Included in this discussion were the results of the Sandia National Laboratory study of the TMI Dome Monitor.

g.

The Davis-Besse Station also ran tests on determining pressurizer level using the change in resistance af the heater banks when uncovery of the bank occurs.

2.

To date a quiz specifically addressing accident mitigation has not been administered. The requalification quizzes are given in the subject titles of the NRC examination. Therefore, questions concern-ing safety systems and procedures are given when these lectures are presented.

For example the present requalification lecture series discusses the OTSG tube leak emergency plan along with the actual transients that occurred at Oconee and Ginna Nuclear Power Stations.

The quiz on this lecture is required to be taken by all licensed operators and shift technical advisors.

ATTACHMENT Docket No. 50-346 Serial No. 816 Page 2 of 4 License No. NPF-3 May 10, 1982 f'

l The Operations Manning Chart is as follows:

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Assistant Station Superintendent (SOL)

Technical Engineer (SOL)

Operations Engineer (SOL)

Operations Operations Supervisor (SOL)

Shift Tech. Advisor Tsch.

Coordinator (SOL)

Shift Supervisor (SOL)

Tsch. Proj. Sup.

(RO)

Ass't. Shift Supervisor (SOL)

Nuc. & Perf. Eng.

(RO)

Reactor Operator (RO) also Nuclear Operations Training Supervisor (SOL)

Qualification Instructor (SRO) i Senior Operator !.icense NOTE: SOL

=

Operator License RO

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ATTACHMENT Docket No. 50-346 Serial No. 816 Page 3 of 4 License No. NPF-3 May 10, 1982 3.

The Davis-Besse Operator Training Program has been significantly changed to include much more emphasis on plant transient behavior.

Prior to TM1 the licensed operator training program Pressurized Water Reactor Technology (PWR) Program consisted of 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> of lectures en plant systems with most of the training on the individual system configurations. The program now consists of 280 hours0.00324 days <br />0.0778 hours <br />4.62963e-4 weeks <br />1.0654e-4 months <br /> classroom instruction. The additional time is used to dis-cuss not only the systems but their effects on all aspects of operation normal and abnormal.

The simulator training for license candidates has also been significantly modified. The first 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> lecture, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> on the simulator) deals primarily with startup certifi-cation and low power plant operations.

Following this the student commences part of the Control Room watchstanding under instruction.

The student thc, is given lectures on emergency and abnormal procedures, along with occident diagnosis and mitigation.

Following these lectures an additional 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> at the simulator is provided where the license candidate maneuvers the plant for power operations, transients, emergencies, and accidents. After simulator training the operator returns to the Control Room for final watch training under instruction.

The last phase of the operator program (Specialized License Training) consists of an audit examination similar in scope to the actual NRC license examination.

Based on the results of this examination, the Specialized License Training lecture series is presented which is approximately 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of advanced PWR, technical specifications, reactor theory, and thermodynamics.

4.

The instructors at Davis-Besse who have administered licensed operator training hold a Senior Reactor Operator license. They are cognizant of operating history, problems and procedures by the fact that they initiate the required reading and lectures for the operators. Also, they review all Licensee Event Reports, Transient Assessment Reports, etc. for operation of Davis-Besse and of other Babcock & Wilcox plants. Currently one instructor holds a Reactor Operator license; however, he is in training for the next Senior Reactor Operator license examination in December 1982. This instructor will not be teaching systems or plant responsec addressed in the Denton letter.

Also, all examinations and lectures given are reviewed and approved by the Nuclear Operations Training Supervisor who does hold a Senior Reactor Operator license. All training instructors are required to take part in all aspects of the requalification program which in-cludes a minimum actual plant watchstanding requirement of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per quarter.

5.

The Davis-Besse Requalification Program does mandate an accelerated requalification for any operator scoring less than 80% overall or any category less than 70%.

In addition, the operator is removed from license duties during the accelerated requalification. All operators acoring less than 80% in any category is required by procedure to attend that lecture series for the category below 80%.

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ATTACHMENT Docket No. 50-346 Serial No. 816

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Page 4 of 4

- License No. NPF-3 May 10, 1982 6.

The Babcock & Wilcox simulator now has a computerized-program listing j

each operator by name and by position at the control panel (ie, RO, Ass't. RO, SS, etc.). ' The program lists the events performed by name and by reference number to the Denton letter.

Several events cannot be performed on the simulator such as " loss of instrument air".

For these events the operators are given an in-plant walkthru by the training staff or other designated examiner, and these are entered

~in appropriate records. The combination of simulator and in-plant walkthrus allow completion of the Denton letter items in the allowed time frame.

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1 Babcock &Wilcox Power ceneration Grcup P.o. Box 1260, Lynchburg, Va. 24505 Telephone: (804) 384-5111 December 8,1980 TRAINING INFORMATION 110TICE SIMULATOR TRAINING PROGRAM C0CUMENTATION The requirements to perfom certain control manipulations are specified in 10CFR Part 55 as part of the operator requalification program.

These requirements are further defined in Enclosure 4 of H.R. Centen, NRC, letter to All Power Reactor Applicants and Licensees dated March 28, 1980, and in draft ANS 3.1 Standard en Qualification and Training of Personnel for Nuclear Power Plants.

Provisions are made to allow ccmoletion of these control manipulaticns at a simulator facility.

The Babcock and Wilcox Simulator Training Center is capable of providing simulation for most of the required centrol manipulations.

The Babcock and Wilcox Training Center provides documentation to each customer de-scribing what drills and evolutions eacn individual performed during training at the simulator facility.

Effective with training completed the week starting December 8, 1980 a revised fomat for the reporting of training will be used.

A copy of this new fomat is attached.

Each plant training staff will be able to keep track of individual participation in the required evolution and drills by airect correlation of B&W report

iI with the individual's annual training requirements.

The Training Summary Sheet represents our effort to provide documentation of training performed in an imoroved femat that reduces confusion and allows you to readily determine the star.us of each operator in meeting the recuired control manipulaticns fcr his requalification program.

Any questions or coments conccrning this me: hod of documentation should be to W. H. Odell, Manager, Instruction Services.

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.j. Elliott anager Tra\\ining Services NSE:hcv RECE!VED

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DEC 121980 4

NUC TRNG DEPT The Babcock & Wi!Cox Company / Established 1867 E r.t I. 2

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SIMULATOR TRAltilitG

SUMMARY

SHEET has completed a

-week training program consisting of hours of classroom time and hours of simulator operations. The time spent in the simulator consisted of performing the evolutions listed below in the indicated capacity with the remainder of the time devoted to manual and automatic ICS power operations.

SS = Shift Supervisor ARO = Assistant l'eactar Operator F = Foreman STA = Shift Technical Advisor

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R0 = Reactor Operator i

Evolutions Performed SS F

R0 ARO STA (1) Plant or reactor startups to include a range that reactivity feedback from nuclear heat addition is noticeable and heatup rate is established.

-0 (a) Reactor Startup (to 10 Amps) l l

(b) Reactor Starutp (to 5k Power)

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' '(c). Reactor Startup (to 15;; Power) l l

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(d) Reactor Startup (to 100,i Power) j j

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l (e) Plant Temp Cnange Due to Heatup/

Cooldown (

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(f) Calculate a ECP l

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(g) Plot a 1/m l

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l (2) Plant Shutdown.

(a) Power Reduction (100~, to IS" Pcwer) l l

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(b) Power Reduction (155 to not Snutcown) l l

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R0 ARO STA 1.L Evolutions Performed l 0 f

t (3) Manual Control of Ste

,1 Generators and/or k,

feedwater during startuo and shutdown.

(a) Main and Startup Valves in Manual During Startup (b) Main Feeodater Pumps in Manual During Startup l

(4) Boration and or Dilution Durino Power Ooeration. __

(a) Adjust Rod Height for Rod Index Curve Limits

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(5) Any significant (> 10%) Power Changes in Manual Rod Control.

(a) Defeat Neutron Error Signal from ICS (b) Diamond in Manual

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(6) Any Reactor Power Change of 10% or Greater Where Load is Perfomed with Lead Limit Control or Where Flux, Temperature, or Speed Control is on Manual.

(a) Reactor Power Level Change > 10% Power With Either Main Turbine, Rod Cemand, Diamond, Main Feedwater Valves and/or Main Feedwater Pumos in Manual.

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age 3 SS F

R0 ARO STA Evolutions Perfomed (7) Loss of Coolant.

(a) Significant 0.T.S.G. Tube Leak l

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a capacity (

1. /sec)

(b) Reactor Coolant Leak Inside Primary l

Containment (Capacity

1. /sec)

(c) Reactor Coolant Leak Outside Primary Containment (Capacity

1. /sec) t (d) Leak-Rate Determination (e) Saturated Reactor Coolant Response (PWR) 1 I

I (8) Loss of Instrument Air (If Simulated Plant Specific)

Not Simulated.

(9) Loss of Electrical Power (and/or Degraded Power Sources).

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(a) Blackout l

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l (b) Fail 6900 Vac Bus l

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l (c) Fail Startuo Transfor:ter l

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-(d) Fail Diesei Generator I

(10) Loss of Core Coolant Flow / Natural Circulation.

(a) Loss of Power to All Reactor Coolant Pumps (b) Actuation of Safeguards System Which Re-quire Shutting Off RCP's due to Low Pressure Condition (Leak, Overcooling, etc.)

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Page 4 SS F

R0 ARO STA Evolutions Perforned J

(11) Loss of Condenser Vacuum.

(a) Failure of Condenser Vacuum Breaker l

(0 pen / Closed)

(b) Fail Air Ejector l

(c) Fail Condenser 3-Way Valve (d) Fail Gland Seal Steam Supoly Valve (e) Fail Condenser CW Pumps I

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(12) Loss of Service Water if Required for Safety.

NOTE: The Canal Water System Cools Components Associated with Service Water.

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I (a) Fail Canal Water Pumps

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(b) Fail SU Transformer #1 l

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-1 (13) Loss of Shutdown Coolina.

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a) Failure of Decay Heat Pump During S/D l

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Cooling (b) Fail USCW Valve to DH Cooler l

l (c) Fail Open Decay Heat Bypass Valve (d) Fail NSCW Pumps I

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Page 5 Evolutions Perforred SS F

R0 AR0 STA (14) Loss of Component Cooling System or Cooling to an Individual Cemconent.

(a) Failure of Component Cooling and Con-tainment Penetration Valve (0 pen / Shut)

(b) Fail Component Cooling Water Pumps (c) Fail CCW Valve to Letdown Cooler l

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l (15) Loss of Nomal Feedwater or Nomal Feedwater System Failure.

(a) Failure of Condensate Pumo(s) which Cause Loss of Main Feedwater Pump (s)

(b) Loss of Main Feedwater Pump (s)

(c) Fail Main Feedwater Control Valve c.;

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(d) Fail Main Feedwater Block Valve a

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(e) Fail Startup Feedwater Control Valve (f) Fail Startup Block Valve (g) Vary Feed Pump Speed Feecback Signal (h) Degrade Feed Pump Speed l

(16) Loss of All Feedwater (Nomal & Emercencv).

(a) Fail Emergency and Main Feedwater-Pumps (b)

Fail Emergency and Main feedwater Block Valves (c)' Fail Emergency and Main Feedwater i

Control Valves l

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Page 6 Evolutions Performed (17) Loss of Protective System Channel.

(a) Defeat RPS Trip Functions (b) Defeat ESFAS Automatic Initiation l

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(18) Miscositioned Control Rod or Rods (or Rod Droos).

(a) Dropped Rod (b) Stuck Rod I

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(19) Inability to Drive Control Rods.

(a) Motor Fault l

l (b) Retard Rod Motion i

I (20) Conditions Recuirino use of Emeraency Boration.

Note: Dnergency Boration Will Occur Whenever the HPI Pumps Take a Suction From the BWST.

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1 Page 7 Evolutions Performed SS F

R0 ARO STA (21) Fuel Cladding Failure or High Activity in Reactor Coolant or Offaas.

(a) ' Failed Fuel Casualty l

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(22) Turbine or Generator Trio (a) Load Rejection (b) Turbine Trip (c) Defeat ICS Signal to EHC NOTE: Record for All Casualties When Result is a Reactor Trip (23) Malfunction or Automatic Control System (s) Which Affect Reactivity.

Note: These Malfunctions can be Satisfied Under Sections (18) or (19).

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

(a) Loss of Makeup Pumo(s) l l

(b) Failure of Pressurizer Spray Valve (0 pen / Shut)

(c) Failure of Letdown Isolation Valve (0 pen / Shut) j (d) Failure of Makeup Valve (0 pen / Shut)

(e) Failure of PORV

-(f) Failure of Pressuri er Palief Valve e

Page 8 Evolutions Performed SS F

R0 ARO STA (24) Continued (g)

Failure of 3-Way Valve (h)

Fail Pressurizer Level Control Valve i

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1 (25) Reactor Trio.

NOTE:

Record for all Casualties Which Result in a Reactor Trip.

1 (26) Main Steam Line Break.

(a)

Inside Containment (Capability f/sec)

(b) Outside Containment (Capacity f/sec) l (27) Nuclear Instrumentation Failure (s).

(a) Failure of Power Range NI (b)

Failure of CIC f

(c) Failure of P.C.

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Page 9 Evolutions Performed SS F

R0 ARO STA h

(28) Instrument Failures.

i (a) Fail Selected TH Instrument High/ Low l

(b)

Fail Selected TC Instrument hign/ Low ll 4

(c) Fail RCS Flow Signal to ICS f

(d) Fail Header Pressure Signal to ICS l

(e) Fail Feedwater Flow Signal to ICS (f) Fail 0TSG Startup Level to ICS l

l (g) Fail 0TSG Operate Level to ICS (h) Fail Neutron Error l

(1)

Fail Steam Generator Pressure Signal l-l l

(j) Fail Feedwater Temperature Signal l

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(k) Fail Power Range Signal to ICS l

(1) Fail Feed Ptrcp aP 4

l (m) Fail Generated Fegawatts (n) Fail Tave l

(c) Fail RCS Pressure Signal l

(p) Fail Pressurizer Level Signal li l

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(29) Unassioned casualties and Evolutions.

(a) RC Pump Trip (b) Plant Temperature Changes > S0 F (c) TMI Demonstration (d) Solid Plant Ocerations l

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(e) Draw a Bubble in the Pressurizer I

(f) Fail Seal Water Control Valve (g) Fail NSRW Pump (h) Degrade Low Pressure Feed Heater (1) Degrade High Pressure Feed Heater (j) Fail Plant Cooling Water Pump l

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(k) Fail Turbine Plant Cooling Water Pump l

l (1) Fail.RB Emergency Coolers j

(m)

Fail RB Spray Pumps l

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Evolutions Performed SS F

R0 ARO STA d

(29) Continued L-(n) Fail HP Injection Valves (o) Fail Boron Addition Valve l

(p) Turbine Trip Locked Out on Reactor Trip l

l (q) Fail Turbine Bypass Valve l

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(r) Degrade Secondary Steam Relief Setpoint l

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l (s) Fail Heater Drain Pump l

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ADDIT 80?tAL PAGE FOR (29) U?iASSIGNED CASUALTIES AND EVOLUTI0ftS Evolutions Performed SS F

R0 ARO STA I

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TEC0 1981 REQUALIFICATION SIMULATOR MONDAY

1.. Reactor start-up safety rods withdrawn to 100% Power 7;

Fail CIC during S/U Fail F. W. during S/U Feedwater pump in manual during S/U Manual control' of reactor, feedwater and turbine during power increase (document manual operations)

TUESDAY 1.

Tube leak 500 gpm (also document plant shutdown and cooldown) 2.

Small break in RC System outside Reactor Building (letdown or makeupsystem) a.

Detennine leak rate b.

Locate and isolate 1

3.

Loss of Power to all RC pumps 0

Natural circulation cooldown to 500 1

(document natural circulation)

WEDNESDAY 1.

Loss of component cooling water to letdown coolers 2.

Loss of condenser vacuum (start with small steam leak then cause steam reducer valve to fail losing auxiliary steam to air ejectors) 3.

Loss of all feedwater normal and emergency (document solid plant j '_

t operation) 4.

Loss of RPS 5.

Reactor trip, turbine fails to trip 6.

ICS failures (temperature, levels, feedwater temperature, etc.)

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THURSDAY j

Adjust rod heights for rod index curve by changing boron concentration (make calculation necessary to support exercise)

Lct.s of cooling accident - leak large enough to make the RC reach saturation conditions and not repressurize (document emergency boration after actuation of HPI.)

(document operation at saturated conditions) carry out to long term cooling (suction of LPI from CVsump)

Reactor trip with secondary safety valve stuck open FRIDAY Blackout with both diesels failed (feedwater available) 15-20 min repair one diesel Return nonnal power 35-40 min.

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REQUALIFICATI0li TRAINING FOR 1

TECO 1981 CLASS ROOM SCHEDULE CONTROL ROOM SCHEDULE Reference Referencs N.

Day / Bats ilms Salist!

Instructor Ties Opstaties Instructor 07:30 REVIEW to CECAY llEAT & I ETil0DS TO REMOVE DECAY liEAT TRAINING 09:30 NATUPAL CIRCULATION SERVICES RC SYSTEH AT SAT DURING C00LDOWN b

GENhNb$[o$[G$rr[ChS$N OL NG D

D 11:30 BOR0tl PRECIPITATION CONCERS FOLLOWING LOCA l-07:30 CONSEQUENCES OF INADEQUATE CORE COOLING AND

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AND LIKELY CORE DAMAGE EFFECTS TOM THORNTON 1 30

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f 10:30 USE OF SPND'S IN R COGNITION OF DEGRALED TRAINING CORI C0tIDITIONS SERVICES 1 30 07:30 DETECTION #iD TRFATHENT OF INADEQUATE CORE TRAINING

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COOLING USING CORE EXIT THERMOCOUPLES SERVICES 09 30 09:30 THERMOCOUPLES #{D CORE FLOW BLOCKAGE RELATED TRAINING to SERVICES TO THI-2 10:30 lb$3b"'"h[Lk5[b$55IP bh bbT bF BORE SOURCE kkkGb hRAkHlNG to TO DEGRADED CORE COWDITION

' SERVICES 11:30 RELEASE OF FISS10:1 PRODUCTS & FISS10ti PRODUCT DON TRN4 SPORT NITTI I.

to RESPONSE OF GAMMA RADIATION MONITORS CHEMICAL AND RADIOC5EMICAL SAMPLING PROBLEMS 11:30 07:30 t

REVIEW OF OPERATING WITH CORE DAMAGE TRAINING 09 30 SERVICES 09:30 RIVIEW OF RELATED TAP'S THAT COULD EFFECT TPAINING f.)-

t 11:30 TECO SERVICES

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