ML20002E317

Forwards Emergency Procedures & Training Program Re Recognition & Mitigation of ATWS Event.Procedures Provide Emergency & Off Normal Responses to ATWS & Inability to Insert Control Rods

ML20002E317
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	01/22/1981
From:	Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
IEB-80-17, NUDOCS 8101270373
Download: ML20002E317 (31)

Text

,

Consumers

Power s

company I

D General OfHces: 212 West Michigan Avenue, Jackson, M6chigan 49201 + (517) 788 0550 January 22, 1991

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Director, Nuclear Reacter Regulation Att Mr Dennis M Crutchfield, Chief Cperating Reactors Branch No 5 US Nuclear Regulatory Concissien

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l Washington, DC 20555 a

ECC IT 50-155 - LICENSE DPR 'f BIG ROCK PCINT PLANT - E'ERGE' ICY PECCEDURES AND TRAINING PROGPR4 FII.ATING TO THE RECOGNITICN AND MITIGATION OF ANY A'ITICIPATED TRANSIE:C WITHOUT SCFR4 (AT4S)

EVE'C Amendment No 33 (dated 1/15/61) to Facility Operating License No LPR-6 for the Big Rock Point Plant required Constrers Pcwer Cenpany to submit, prior to re-start from the 1980 refueling cutage, the energency procedures and training program that relate to the recognition and citigation of an AT4S event.

The folleving attached procedures provide both the emergency and off-normal procedural respcnses related tc AT4S and the inability to insert control red (s),

respectively: - Procedure #EMP 3.5A Rev 2 (7/18/60), Anticipated Transients Without Scram ( A""4S) - Procedure #0NP 2.9 Rev 3 (7/17/80), Multiple Red Insert Failure (Two or More) - Procedure #0NP 2.b Rev 1, Single Rod Insert Failure Attachments 1 and 2 have been updated to address the recent failure to scrP1 event at 3rovns Ferry 3 (IE Bulletin No 80-17).

Training in accordance with NRC regulations in the use of the above procedures is provided to all licensed operations personnel (SR0s and R0s); some non-licensed Auxiliar'/ Cperators, all Shift Technical Advisors (STAS); and, all en-duty Superin-tendents.

"'his training occurs in the form of onsite classroom training and off-site si=ulator training at the GE BWR TC simulator (Morris, Ill.).

The ATWS train-ing at the simulator involves the AT4S sequences of events first in actual speed; A o*'

f,9 n o mo 393 y

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Director, Nuclear Reactor Regulation 2

January 22, 1931

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j Big Sock Point Plant i

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i then in stop actien; and, again at actual speed. 3ig Ecek Pcint operators have

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consistently perfor: sed the actions required during the ATWS event (ie, recogni-(

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tien of the event and performance cf required actiens) in less than one =1nute

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during simulater training. The following attached Cperater Requalification j

Progra=, Three-Day Simulater Pregram and Procedures Manual Revision review docu-I sents provide the basic training applicable to ATWS events:

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Attachment L - Appendix 5: - Big Rock Point Plant - Three-Eay Simulator

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Program - GE EVE TC Simulator f

Attach =ent 5 - 1950 Operator Eequalification - Eequalification Classroom l

Reviev 6-90 i

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i - Censumers Power Company - Big Ecek Point Huelear Plant -

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Record of QA Review for Licensed RO/SRO and Prospective Licensees - 07/1S/80 (for EMP 3 5A Rev 2) r (NOTE: ATWS references have teen underlined in these documents) s l

Attachments 1 through 6 were previcusly given to the NEC staff during a deeting i

en January 19, 1991 as an infernal sut=ittal of the infor stion requir 2 ty j

Arendment No 38.

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4 1

David P Hoffman (Signed) t I

David P Hoffnan i

l Uuclear Licensing Administratcr I

CC Director, Eegicn I~I, USUE0 i

SEC Eesident Inspector - Eig Rock Point 9 pages l

t

3 A Z ACH' C""

l Anticipated Transients Uithout Scram (ATWS) EMp 3.5A Rev. 1 2/25/80

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EMp 3.5A AIWS An ATWS is a primary system transient in which a reactor trip setpoint is exceeded but a cenplete scrs= doesn't occur (ie, an insufficient number of drives or notches insert to lower reactor power sufficiently to reduce reactor pressure).

The consequences will vary with : The initial power level, nu:ber of rods that fail to insert, their relative location in the core and the initiating event.

The nost severe ATWS postclared is caused by the main steam isclation valve closing or a turbine trip with fe lure of the bypass valve to open at 100% power and all rods fail to scram. Under these conditions neutron flux vill spike to 200% and then steady out at v120%. The energency condenser comes into operation and four drum reliefs lif t causing containment pressure to reach 1.5 psig in less than 30 seconds.

g' During this accident our normal water supply would be depletal in j

a 5

approximately 6 minutes.

3.5A.1 Symptons a.

Reactor scram setpoint exceeded with or without annunciation.

b.

All control rod drives not at "03" position.

c.

Flux level still above 4% power.

d.

Primary system pressure at or above 1360 psig.

3.5A.2 Autenatic Actions a.

Safety system may or may not trip depending on whether f ailure of the safety system is the cause of the ATWS.

b.

Emergency condenser comes into operation if reactor pressure exceeds 1435 psig, c.

Drum reliefs lift if reactor pressure exceeds 1535 psig., the Steam Drum Relief Valse Monitor High Alarm should annunciate and the high alarm lamp should

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illuninate for open relief valves.

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i Anticipated Transients Without Scram (ATWS) EMP 3.5A Rev. 2 l

7/18/80 f

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3.5A.3 Immediate Operator Action I

k-a.

Push ti.

nanual scram push button.

b.

Trip both recire pumps (reduces core power 60%).

c.

If the Channel 1 and Channel 2 scram annunciators are not alarming, trip the undervoltage breakers for the safety system.

I d.

If reactor pressure is greater than 1360 psig and not j

falling, inject the liquid poison by simultaneously closing MS 7003 and HS 7009 for five seconds.

(Circuit failure alarms' indicate proper operation.)

i Ref. SOP 4 - Liquid Poison System l

c.

Trip the clean up pump by either:

1.

Pull the cover of TSX-1507-2 located on the back of the plant temperature recorder and depress the relay or:

2.

Open breaker 52-1A-25 (clean up' pump) at 480 volt panel 4

1A in the electrical eqnipnent'. room.

f Until all control rods are verified to be fully inserted f.

g' l

monitor the reactor for decreasing neutron flux and indication of local areas of' high reactivity while per-J forming the following in order:

1.

If scram dump tank hi level alarm is in, bypass i

the SDT alarm by:

(a) Put the mode switch to BPDT.

i (b)

Insert key in HS7048, turn switch to right 4

i and hold.

(c) After tank drains, return mode switc'h to RUN.

2.

Individually scram those rods not-fully inserted starting with'those in' local areas of indicated high reactivity.

Reset each after it fully inserts and continue to scram rods until all rods are~ fully

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4 inserted.

If rode will not scram in:

(a) - Try nanual scram again.

r' (b) Insert rods manually as the rod sequence permits.

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EMP 3.5A Rev. 2 7/18/80

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3.5A.3 Immediate operator Actica (continued) g.

If drum reliefs were actuated as annunciated on the Relief Valve !!onitor and containnent pressure is 2 psig or greater, use primary containment sprays (MO-7064) to reduce prescure and; h.

Sound the plant siren for two minutes and initiate the

" Site Emergency Plan."

3.5A.4 Subsequent Operator Action a.

Attempt to maintain feed water system by transferring water from radwaste and the demin tank to the condensate storage tank, b.

Check emergency condenser for proper operation.

c.

If the scram valves do not indicate open, close the instrument air supply valve to the sphere (located in the equipment room).

d.

Containment sprays must be valved cut before drum Icvel jf

't goes off scale on the low side. This is to provide adequate margin of water for core spray in the event that a relief valve has stuck open or other type LOCA has occured due to l

over pressure.

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ATTACICEiT 2

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l Multiple Rod Insert Failure (Two or More) - ONP 2.9, Rev. 3 7/17/80

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ONP 2.9 NJLTIPLE ROD INSERT FAILURE This procedure will address three different conditions for rod insert f ailure; multiple rod insert failure following a scram with subsequent

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power above as well as below 4% and the case where multiple rods fall to insert when being driven in manually.

ONP 2.9.1 If two or more control rods fail to insert after a reactor

. scram and reactor power remains greater than 4%, perform the actions of EtiP 3.5A, " Anticipated Transients without Scram."

ONP 2.9.2 Two or more control rods fail to insert after a reactor scram and reactor power'isless than 4% and is stable or decreasing:

CAUTION: If reactor water level cannot be maintained and unable to maintain reactor suberitical, inject liquid poison (ref. SOP 4).

2.9.2.1 Symptoms

[

1.

Channel 1 and Channel 2' scram annunciators on v=

2.

Reactor power indicates less than'4%

3.

CRD positi6n indicators on two or more drives indicate a position other than "00".

2.9.2.2 Automatic Actions!

None.

2.9.2.3 Immediate operator Actions If two or more control rods fail to insert after an automatic reactor scram, depress'the manual scram button.

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NOTE: If the failure to scram occurs during a shutdown using the " manual scram," place' l

the mode switch in'" shutdown" momentarily to insert another scram signal and then if-necessary proceed on with this procedure.

3l '. Confirm all scram valves are open by observation of scram valve position lights ~. If not, de-energize the Reactor Protection System by tripping the RPS breakers.

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Multiple Rod Insert Failure (Two or More) - ONP 2.9, Rev. 3 7/17/80 g

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2.9.2.3 Ier.ediate Operator Actions (continued) 2.

Place the reactor mode switch in the " bypass dump tank" position.

3.

Insert key in HS7048, turn switch to right and hold.

4.

Reset reactor safety system.

5.

Continue to hold 11S704S until ducp tank high level alarm resets, then release and recove key I

from HS7048.

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i Multiple Rod Insert Failure (Tuo or More) - ONP 2.9, Rev 3, 7/17 /8 0 2.9.2.3 Immediate Operator Actions (continued) 6.

Attempt to manually insert those control rods which failed to insert at time of scram, if i

control rods cannot be fully inserted, individual-ly scram them. Consider scram =ing rods in areas of high reactivity first.

7.

If control rod inward motion was observed during step 6 above, but full insertion did not occur, i

return the control rod circuit breakers on power switches for reactor protection system f

channels 1 and 2 to the "on" position. Repeat Step 6 until control rods are fully inserted.

ONP 2.9.3 If two or more control rods fail to insert following an i

.. insertion signal but..'a scram has not_ occurred:

2.9.3.1 Symptoms

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Two or more selected control rods have been given

, j individual insert aignals and the following obser-l

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vations are made:

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- rod (s) position'digiEl display ~ ~

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unit numerical value does not change and/o decrease.

2.

The out-of-core neutron monitoring instrument j

meters do not change and/or decrease in their l

readout values.

3.

The out-of-core neutron monitoring instrument l

recorders do not change and/or ' decrease in their recorded values.

l 2.9.3.2 _ Automatic Actions a

The automatic operations associated with the control' rods are.

1.. Rod drive high temperature alarm at 250 F.

2.

Rod drive accumulator lo pressure' alarm at '

750 psig. Two low accumulators prohibit control rod withdrawal to assure that shutdown criteria is not. exceeded.

L 3.

Control rod drive filter (s) high A p alarm at -

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20 psig.

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Multiple Rod Insert Failure (Two or More) - 0:'? 2.9, Rev 3, 7/17/80 2.9.3.3 l==eciate Operator Action q

1.

Check control rod drive (s) selector switch's (alphabetical and nunerical) positions to assure selected drive nunber. Also observe selected i

drive (s) digital display unit is illuminated.

2.

Check control rod drive system operating pres-sures normal:

1 Accumulator Charging Header PR +400 Rod Drive Header PR +200 4

i Rod Drive Cooling Header PR +30 3.

Check breaker 4CEl closed, located in Section "C" of control console in control roo=.

(Toggle up is closed.)

4.

Check control rod drive system filter (s) for

.high Ap >20 psig.

l 5.

Check rod drive temperature recorder for individ-ual high rod drive (s) tcuperature.

i 6.

Check control rod drive (s) associated valving s

for proper alignment.

7.

Check control rod drive (s) Atkomatic hydraulic 4

system "A" and/or "B" set, whichever is in service, for proper valving alignment. Verify "A" and/or "E" Athomatic set selector switch position inside control console Section "C" Control room.

2.9.3.4 Subsequent Operator Actions 1.

Af ter verifying selected control rod (s) selector switch's position, check ~ control rod drive (s) selector valve (s) for : proper operations, adjust s l

rate set valves and/or clean screens in rate set block assembly.

2.

Readjust control rod drive system operating pressure an6/or switch control drive pumps.

Try increasi$g drive pressure to move control

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2.9.3.4 Subsequent Operator Actions (continued) 3.

If control rod drive filter is > 20 psig, cut in standby filter and valve out high 2p filter.

4.

If the control rod drive (s) associated valving is properly aligned, switch Atkomatir. bydraulic system valving from "A" to "B"' set or vice versa, depending on which set is in service at the time.

5.

If control rod drive (s) temperature is 350 F, and coolant flow is lost to all drives, nanually Scram reactor.

6.

If two or more control rods cannot be inserted and the reactor power is stable: Immediately shut down the reactor per control rod insert procedure sequence cards.

t

The reactor shall be shut down:

2 a.

If it is determined by investigation that the manfunction which has occurred impairs the ability to control the reactor.

b.

If the co e shutdown margin require =ent-cannot be met with the remaini2g operable rods, evalu-ation of this requirement shall be based on

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previous experimental measurements.

c.

If control' rod drive (s) temperature is 350 F and coolant flow is lost to all drives, manually *

(Highroddrivetemperature[may scram reactor.

cause graphitar seals to swell and bind the drive (s), preventing control rod insertion.)

d.

The liquid poison system shall be used c; any time reactor subcritical'ity cannot be assured due to failure of normal ~ shutdown mechanism.

Ref. SOP _4 - Liquid Poison System.

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Single Rod Insert Failure - ONP 2.8, Rev 1

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r CNP 2.8 SI'IGLE ROD I ISERT FAI'URE An f.

ediate investigation shall be made to deter =ine the cause of a cont rol rod failure to insert.

The reactor shall be shut down unless:

a.

It is dete=ined by investigation that any =alfunction which has occurred neither impairs the ability to control the reactor nor indicates the i==inent i=paiment of the perfomance of additional co=ponents of the reactivity control syste=.

b.

The operating hydraulic water to the defective control rod has been tagged and valved out to prevent withdrawal of the control rod after an atte=pt has been =ade to insert the control rod.

c.

The core shutdevn =argin require =ent can be =et with the re=aining operable control rods. Evaluation of this require =ent shall be based on previous experimental =easure=ents.

d.

Per=ission has been received frc= the Plant Superintendent to con-tinue reactor operation with one control rod valved out.

o ONP 2.8.1 SYMPTOM 5 A specific control rod has been selected and an insert sie, al given and the following observations are made:

a.

The control rod position digital display u=it nu=erical value does not change and/or decrease.

b.

The out-of-core neutron =onitoring instrc: ent =eters do not change and/or decrease in their readout values.

c.

The out-of-core neutron monitoring instru=ent recorders do not change and/or decrease in their recorded values.

d.

An auto =atic and/or mual reactor scra= occurs and all control rods insert except one.

02P 2.8.2 AUTOMATIC OPERATIONS The auto =atic operations associated with the control rods are:

a.

Rod drive high tenperature alar = at 250 F.

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b.

Ecd drive accu =ulator low-pressure ala= at 750 psig (two required) (prohibits control rod withdraval to assure that shutdown criteria are not exceeded).

Control rod drives filter (s) high do alar = at 20 psig.

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c.

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d.

Control rods insertion upon initiation of an auto =atic or manual scra= function.

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2 Single Rod Insert Failure - CMP 2.8, Rev 1 1

y-03P 2.8.3 D' MEDIATE OPE"ATOR ACTIONS Check control roi drive selector switches (alphabetical a.

and nu=erical) positiens to assure selected drive nu=ber.

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Also observe selected drive digital display unit is illu=inated.

4 l

b.

Check control rod drive syste= operation pressures nor=al:

1 Accu =ulator Charging Header PR + E00 Rod Drive Header PR + 200 Rod Drive Cooling Header PR + 30 c.

Select another control rod drive and give it an insert signal to dete:=ine if drive insert proble: is cc==on to all drivec.

d.

Check contrt,,1 rod drive syste filter (s) for high op > 20 psig.

e.

Check rod drive te=perature recorder for individusi high l

rod drive te=perature.

f.

Check centrol rod drive associated valving for proper

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align =ent.

g.

Check control rod drive (s) Athc=atic hydraulic syste l

"A" and/or "B" set, whichever is in service, for proper i

valving align =ent.

Verify "A" and/cr ".B" Atic=stic set selector switch position, inside control censole Section "C"

control roc =.

03P 2.8.h SU3SEQUU T OPERATOR ACTIO3 a.

After verifying selected control rod selector switch's position, check control rod drive selector valve for proper operation, adjust rate set valves and/or clean screens in rate set block asse=bly.

b.

Readjust control rod-drive syste= operating pressures and/or switch control red drive pu ps.

Try increasing drive pressure to =ove control rod.

c.

If centrol rod drive filter is > 20 psig, cut in standby filter and valve cut high Ao filter.

. - ~ -

3 Single Rod Insert Failure - 03P 2.8, Rev 1

(

d.

If control red drive te=;erature is > 350 F and coolant flev is lost to all drives, manually scram reactor.

e.

If control rod drive assceinted valving is properly aligned, switch Athc=atic hydraulic systen valving from "A"

to "3" set or vice versa, depending on which set is in service at the time.

f.

Valve cut and tag the stuck control rod, with the excep-tica of the drive ecoling water which must be left "on."

g.

At earliest convenience, shut down reactor and make preparations to renove the stuck drive.

h.

Remove, repair and/or replace stuck drive.

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ww mounrmmm a Appendixes (continued)

ATTACEME:IT L

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APPENDIX K r

BIG ROCK POIUT PLA'IT Three Day Simulator Program GE BWR TC SIMJLATOR I

PROGRAM OBJECTIVES:

The three-day simulator program is structured to provide operating experience during normal, off-normal and ' emergency conditions as vou'.d be experienced at the Big Rock Point Plant. This program meets the objective stated in the requalification program submittal to the NRC Licensing Branch in 1963, and is accepted by NRC approval of the program. Reactor Operation gained at the simulator is credited as experience on NRC License applications and reapplications or renewal.

II PRE-REQUISITES:

Students participating in the program should hold a valid NRC react'or operating or senior reactor operating license.

If a license is not g...

t/

held, the student should have completed the reactor fundamentals course v

and reactor technology course.

He should also have had two years of nuclear plant experience. He should be familiar with control room operation.

He should have completed a three-week preparatory course conducted by the Big Rock Point training staff prior to attending the simulator course.

1 III ENROLIlE;T:

The maximum nunber of studeits attending the three-day simulator program t

i should be limited to four.- This may be exceeded if this is a familiar-i j

ization nession prior to hot licence demonstration course, or a first time visit for information,.orposes.

1 l

l IV COURSE PRESENTATlud.

The three-day simulator program consists of Practical hands-on der.onstration on similar controls as are located at the Big Rock Point Plant.

y 16 April 1980 50 Rev 25

10.4 Operation's Department (continued)

Appendixes (centinued)

IV COURSE PRESENTATION:

(continued)

The simulator control consol being of a screwhat more aCtanced design t

than the BRP control room consol, requires a short familiarization with control locations. Several of the simulator systems are Jocked out-of-service to make the plant operate similar to Big Rock P'oint.

The plant operation and off-normal operation very nearly duplicates the ERP plant operation. The student, when not directly involved in hands-on operation vill be expected to explain plant operation during oral discussions.

Students vill be expected to work on reactor auxiliary panels, conventional plant system panels and reactor operating panels.

One student vill also be assigned to act as senior operator, directing the overall plant operation. All students are expected to take an active part in all plant manipulations as estsblished by the simulator instructor. The simulator is to be treated as an operating nuclear plant.

Appendix "A" to this lesson plan describes the schedule for the three-day program.

l 4

l 16 April 1980

$1 Rev 25

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Appendi5es (continued)

ATTACIDENT I DAY 1 BIG ROCK FOIUT SIMULATOR TRAINIUG 3-day Program 0

A.

Initialise mode 2-165 F Co2d suberitical All rods in.

1.

Check Panel 902-5 as a group (1 system per student) a.

Feedvater and level contro'.

b.

Control rod drive system setup and control c.

Reactor manual control system d.

SRM, IRMs c.

RX recire pumps f.

Cleanup blevdown and flow control 2.

Discuss requirements for startup-GOP 1 3

Discuss requirements for startup-RCP system h.

Discuss requirements for startup-CRD systen B.

Initialize mode 2-165 Oy cold suberitical 1.

First operator pull critical a.

Students 'ob' serve RX response ^and~ discuss the following:

Suberitical multiplication f--

Administrative limits for startup Limitations on rate of power increase Reacter period calculations-cach student b.

Uncoupled rod-stall comp. discuss 0

Establish htng power-increase mod temp to 185 F c.

discuss following:

I' System heat loss-vs heating power Overlap of instrumentation Te=perature coefficients I

Heatup rate limitations d.

Reduce power thru irms until strongly suberitical 2.

Second operator pull criticals from previous condition (Mode 2) l a.

Students observe response and discuss:

i l

Tech Specs concerning reactivity anomalies.

e 16 pril 1980 52 Re'v 25-

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Appendixes (continue'd)

DAY 1 BIG ROCK POINT SIMUIATOR TRAINING 3-day program (continued) 2.

b.

Establish heating power and increase mod temp Discuss following:

Min temp for pressurization, operable nuclear inst channels for startup.

Minimum source range countrate for startup Sources and contribution.

c.

Initialize Mode 15 - Hot Scram Recovery 1.

Third operator take reactor critical and into run mode a.

During startup discuss fission product poisons b.

Discuss plant startup requirements for turbine Eeview procedures for turbine control system during c.

plant heatup and startup and reasons.

(BRP)

d. ' Trip recirculation pump Discuss procedures for re-start of a recire pump and how e.

starting at present te=p is different from a cold condition.

~

f.

Initialize Mode 15 - Hot Scram Recovery 1.

Fourth operator take reactor critical and into Run Mo.

2.

Instructor introduce failure of an"IRM channel (fail '

high) Discuss BRP Log H failure before h% and after h5 on pico's.

3.

Instructor. introduce failure of control rod drive y

hydraulic system. Discuss effects on continued l

operation, operator response h.

Turbine'startup a.

Light off turbine b.

Synchronize T.O. to system c.

Discuss and continue power: ascension -

DAY II A.

Initialize Mode 6~- 850//

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1.

Continue plant heatup transfer mode switch to run l

16 April 1980 53 Rev_25

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DAY II 2.

Align feedvater system for power operation 3.

Prepare turbine for startup and bring turbine to 1800 rpm h.

Synchronize turbine to grid 5

Review turbine trips 6.

Trip turbine B.

Initialize Mode 6 - 850#

1.

Continue plant heatup transfer mode switch to Run 2.

Align feedvater system for pove$ operation 3

Prepare turbine and generator for startup h.

Continue plant startup and power ascension. Discuss the following:

MCHFR Peaking factors Control Rod Worth at Power C.

Instructor introduce the following conditions and discuss each individually:

/

1.

Control rod' drift in V

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2.

Control rod scram

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3 Accumulator failure h.

Stuck rod i

l 5

Total loss of feed pumps l

l 6.

Flov control value failure-closed 7

Timer malfunction.

D.

As tite @ermith,.the following items may be covered:

1.

Bypass valve failure. Open, closed.

2.

Recirculating pump seal failure 2

3 Turbine trip without bypass valves h.

Feedvater regulating valve lockup-what to do?'

5 Relier valve failure without annuciation-accucitic device 6.

Scram-scram recovery.

DAY.III A.

Initialize Mode 8, 50% power, clean-l 1.

Discuss and observe xenon transients NOTE: DAY 3 is a-day of Transient Analysis that requires student participation 16 April 1980 54 Rev 25 i

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Appendixes (continued)

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DAY III I

B.

Initialize Mede 9 shutting devn from 505 power 1.

Perform the following transients a.

Fail feedvater master controller closed.

How possible?

Discuss feedvater centrol systen.

b.

Losses of flux indication c.

Place standby feed penp out of service. Fall a second feed pump d.

Fail bp valve open 6 h55 power Trip one recirculation pump e.

f.

Trip 2nd RCP g.

Turbine trip h.

Manual scram 1.

MSIV closure

,J.

Offgas valves closed k.

Cond p trips closed 1.

Loss of condenser vac f

a m.

Loss of coolant f

n.

ATWS o.

Loss of coolant, yarway fails Hi p.

Turbine gov closed.

IFR failure q.

Loss of cond vacuum r.

EVS 21 overcurrent trip-RCP mg's and RX fd Pps i

16 April 1980 55 ReV 25 i

\\

6P6 Appendixes (continued)

C0:iSliEls PO.ER C0'P! sly

,f BIG ROCK Poliff PLA'lT RE0!!ALIFICATIO:! P;UGPW1 EVAll'ATIO:1 SIIRATOR DPERIE! ICE LICEtlSEE'S I!A!'E:

DATE:

1. DATE N!D IYPE OF l1RC LICENSE:

2.

PRESarr LICa:SE RESP 0:lSIBILITIES:

3. EVALUATIort I!!STRUCTIONS.

A.

CHECK BOX MOST CLOSELY C0i: FOP 1111 G TO LICEl:SEE'S PERFORiWICE.

B.

ADDITIO:!AL REf 7RKS PAY BE ADDED AT THE li!STRUCTOR'S DISCRETI0:1 AT THE Et4D OF EACH SECTIO;1 UilDER CCliETTS.

C.

COITBITS ARE REOUIRED FOR A GRADE OF L (POOR) OR 2 m

o y l (UilSATISFACTORY).

b h

f' PERF0PJW!CE DURIllG REACTOR STARTUP.

hg g

e m

s

<.B s

d5 cJPECIFIC AREAS OR EVOLUTIOflS EV?' UATED:

1 2

3 Il ti E

A.

USEANDVi40WLEDGEOFI!UCLEARINSTRUiElTATIO:1.

l

-l B.

'[

L l.g

[

C.

D.

C0f7'E11TS:

16 April 1980 56 Rev 25

1TT.~4 UperaY,lon's Depart,nent (continued)

. Appendixes 9

'$ e =

a E

G f

g<

c-a u

u>

bb

5. PERFORiWiCE DURIf;G fl0PJML OPERATIOils 1

2 3

4 5

6 SPECIFIC AREAS OR EVOLUTIONS EVALUATED:

l i

A.

4

~

p B.

I C.

D.

CC"NBITS:

1 t

6. PERFOPJWICE DURI!!G AB:10PJ%L SITUATIO :S.

I 1

2 3

4 5

6 SPECIFIC SITUATIONS EVALUATED.

I A.

-l l

I B.

C.

D.

COT 749TTS1.

lo

7. GEf!ERAL C0tNENTS:

o 8.

IrlSTRUCTOR:

9. SUPERVISOR:

See attached sh' et for additional cc=ents e

i i

16 April 1980 57 Rev 25 O

18.h Operation's Department (continued)

Appendixes (continued)

GEIEdAL ELECTRIC E01LII;G WATER REACTCR TRAIIll"G CE!!TER

\\

Sinulator Retraining Record for Big Rock. Point I uclear Plant Clas: Dates:

to Student:

TYPE OF REACTIVITY CHA!!GE number perforned 1.

Reactor startup to point of adding heat 2.

Heatup of 50% or nore 3

Power Change with control rod 105 h.

Starting Rep with reactor critical 5

Reactor operations involving energency or

.{

special procedures where reactivity is changing

' I.

l 6.

CRD Scran (single) and Recovery 7

T.

l 8.

x.

9 10.

TOTAL:

Date Instructor:

il l

I i

l Rev 25

(

16 April 1980 58 l

i

~

SPECIAL PROCEDURES OBSERVED OR PERFORMED

/"

/

1.

Loss of Coolant 2.

Feedvater Malfunction, Decreasing Flov 3.

Feedvater Malfunction, Increasing Flow h.

Turbine Trip with Bypass 5

Turbine Trip v/o Bypass 6.

Bypass Valve Fails Open 7

Relier Valves Fail Open at Power 8.

Condensate Pump Failure 9

Malfunction of Control Rod Drive System a.

Control Rod Drive Stuck h.

CRD Drift in-leak b.

Control Rod Uncoupled

{

m r ma une n.

c.

Failure of Flow Control Valve d.

Acewnulator Trouble (1 & 2)

Rod Drive Pump Trip (1 & 2 - cold and hot) e.

f.

Rod Drive Pump Trip-2nd Pump unavailable, g.

Control Rod Drive Scran. Single 10.

Reactor Scram

('.

11.

Loss of Condenser Vacuum i

12.

Loss of Reactor RCP l

13.

Loss of Both Reactor RCPs.

lb.

Loss of Flux Indication l

a.

SRM Downscale l

b.

SRM Upscale

~

~

SRM Detector Drive Failure c.

d.

IRM Upscale

..c.

IFII _Downacale _._

f.

Two IRMs Downscale t$

16 April 1980 59 Rev 25

.e.~

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18.h Oy.ation's Depart:nent (continued)

Append'Aes (continued)

SPECIAL FROCEDURES OESERVED OR PERFORMED f

15 RPs-MG set failure

> 600 psig 16.

RPs failure-AT'.G 17 Feedwater Valve Lockup 18.

Offgas Hi Activity 19 Turb cov failed closed v/EP valve capability 4k5% power 20.

Loss of cond vacuun - recevery 21.

MSIV closes 22.

ISOL condenser tube leak 23 ISOL cond valve leak 2h.

ISOL cond Lo Level \\ Sources for nakeup) 25 Rupture in Dryvell 26.

St. leak pipe tunnel 27 Tu.rb b1dg. Hi Rad (Vaat to check) 28.

Rx bldg Hi Rad Rx Deck 29 r

"0" IS OBSERVED.

"P". IS PERFORMED.

INSTRUCTOR:

Date:

i l

~

1 l

r l

l l

16 April 1980 60 Rev 25 l

l l

i I

AWACF2ENT 5 i

j j

A.

TITLE:

Course

Title:

19 80 Operator Requalification Topic

Title:

Requalification Classroom Review 4

Lesson

Title:

5-80 3.

LESSON OBJECTIVES:

Classroom review of changes vill be made to:

Administrative Procedures pertaining to the Operation's Department, Technical Specifications, Operating Procedures, Operation's Mez::o's, License Event Reports and othe,r docu=ents affecting the trainee in performing his licensed duties, i

C.

RELATIONSHIP TO COURSE OBJECTIVES:

l This classroom review is one of the six (6) required by the Federal Regulations and the Master Training Plan j

D.

MATERIALS REQUIRED:

i Those documents listed under lesson objectives.

E.

INTRODUCTION:

1.

List the lesson objectives and relationship to the course objectives 2.

Class rules: Informal

.L e

l i

i I

1 f

l i

6

Page 2 F.

LESSON B0DY:

OPERATION'S MEMOS 1.

17-80 Radvaste Tank Room 2.

18-80 Scram Dump Tank Surveillance 3

19-80 NRC Reporting for IE Bulletin 80-17 h.

21-80 Reactor Recirculating Pu=p Restart 5

22-60 Sabotage Surveillance 6.

2h-80 Scram Dump Tank Sensors PRC minutes 6-80 and 8-30 VOLUME 1 Ad-in' Procedures 1.

Rev 201 When procedures are revised only the affected nares vill be issued 2.

202 Chapter 15 reporting requirements.revicef

• 3 203 Chapter b Shift turnover is to te done by procedure T1-08 h.

20L Chapter 17 Refers to Tech Spec Tests that are not conpleted on time 5

203 Chapter 10 References to SARB ar.d VP of Huelear operations is deleted 6.

206 Chapter 5 Minor word changes

• T.

207 Chapter h The periodic Test Eoard is issued and approved in accordanet with A.T.3 Methed, of annual Transfer is included

• 8 203 Severai Chapters. The major impact is that QA vill review all safety related procedures whether they are work activity type or administrative type prior to implementation.

Affects the operation's department

[

l VOLUME 2 Tech Specs 1.

An=endnent 33 Containment spray surveillance requirements extended to each refueling outage not to exceed 18 ronths for power operated valves and Enclosure High Pressure sensors and Time Delay l

VOLUME 3 Operating Procedures l

1.

GOP 1 Rev 7 Test nudber was changed i

Startup l

2.

SOP 11 Rev 13 Radvaste tank floor not to be used for water storage.

3 ONP 2 7 Rev h "A continuous rod drift may occur when foreign naterial Mis Positioned Rod becomes lodged in the collet area which prevents relatching" l

t k.

ALP 1.h Rev 11 DrT2 Safety Valve Leak alarm corrective actions revised to include checking acoustical monitor for indication that a relief valve has opened.

Page 3 VOLTE 3 operating Procedures (continued) 5 ALP 1.16 Rev 1 Relief Valve Monitor alarm corrective action revised to include checking sphere pressure for signs or increase 6.

SOP 26 Rev 16 A fire brigade of at least 5 rembers shall be raint-Pire Protection ained en site at all times. The brigade nny be less than mini== for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if immediate action is taken to restore the brigade to mini =um requirements.

7 SCP h Rev 3 Valve nudber changes Liquid Poisen 8.

SOP 17 Rev 3 Equipment tagging procedures added for the Turbine Main Steam Byrass Rydraulic Unit 9

ALP 1.6 Rev 12 Corrective actions revised so that the reactor is scrazed and the main steam valve is closed if there is any indication of a cable tray fire 10.

SCP 10 Rev 7 To insure closure,0the sphere supply and exhaust must be throttles to 75 open position (900 Cont ainment full eren) 11.

SOP 19 Rev 7 Procedure added to receive liqcid caustic Makeup 12.

SOP 30 Rev 1h Surveillance of Scram Dump Tank vent and drain valves is added 13.

OXP 2.9 Rev 3 Multiple Rod Insert Pailure Procedure has been revritten Ih.

ONP 2.31 Rev 6 I==ediate operator actions refer to ORP 2.9 if 2 or more rods fail to insert fully 15 ALP 1 3 Rev 8 Caution: only one reactor recire purp =ay be isolated during power operation 16.

ALP 1.2 Rev 8 Corrective action revised to include dump tank valve

-surveillance 17 E!P 3.5A Rev 2 Reference to SCP h Liquid Poison System added and a procedure added for scraning individual rods

~

which did not insert after a scram signal.

Page h VOLU'E 18 Training Manual 1.

Rev 26 Chapter h for the Operation's Department has teen conpletely rewritten 2.

Rev 27 Chapters 1,2,3,5,6,7,8,9 Appendix A, G and J have nu=erous ninor revisions VOLU'E 21 Fire Protection 1.

Rev 7 Each Bio-Pak is equipped with a Bio-shield TM Hood 2.

Rev. 8 Appendix H revised to include fires in electrical equipment room or cable penetration areas.

3.

Rev.

Fire training responsibility transferred to Property Protection Department L.

Rev 10 Chapter 2 Pernanent fire protection equipdent shall not be used for purpese other than fire protection without the approval of the shift supervisers SS card 3 revised to =ake the SS responsible for fire prevention and protection requirements during eetivities which put a portion of the fire system out of service or increase the chance of fire.

Fire Brigade Leader is to follow the advice of the Property Protection Supervisors if on-site.(Card 5).

CPC0 Letters (Internal) 1.

May 30, 1980 Sphere Tag Boards 2.

Aug 12, 1980 Telephone Reporting Requirements to NRC f

CPCO Letters 1.

June 6,1980 Annual Facility Change Report l

l 2.

June 20, 1980 Quantitiive Thernal and Stress Analysis of Spent i

Fuel Fool Structure as a Result of Fool Boiling l

l 3

June 27, 1980 Proposed Tech Spec Change Request-Fire Protection 4.

July 7,1980 Tech Spec Change Request-Containment Spray Systen Surveillance.

(Approved) 5 July 31, 1980 Response to IE Bulletin 80-17 6.

August 8, 1980 Response to Supplement 1 of 80-17 T.

August 11, 1980 Response to Staff Questions on Spent Fuel Pool Area.

l l

Page 5 NRC Letters 1.

May 28, 1980 IE Information Notice 80-22 2.

June 12, 1980 IE Bulletin 80-lh 3.

June 18, 1980 IE Bulletin 80-15 h.

June 20, 1980 IE Circular 80-15 5

June 24, 1980 IE Circuler 80-lh 6.

Junc 26, 1980 NEC response to Christa-Maria G.

SUMMARY

Review the changes pertiining to operations. Minor word changes or typo errors may be designated as such and little or no tine for review of the material should be required except to note it has been updated for correction.

H.

EXAMIUATION:

There is no examination for this lesson I.

INSTRUCTOR's GUIDELIHES:

Docunent attendance on RQ-h forms and enter dates on the attendance shaet in the front of this manual.

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Big Roch Point Nuclear Plant A Acy,.r;; g RECOPD OF QA REVIE'I FOR LICENSED RO/SRO DUPUCATE NOT FOR Ci. Ass SEss'oN and t

PROSPECTIVE LICENSEES 6

"acility Changes Date Issued [ 7 / 8 O O h

JA Procedures Manual Revision Originator c

Technical Specification Change D

g Operating Incidents 3

- (th

[

Site Emergency Plant Special Operation (s) Test (s) Procedures AO's, UE's, Deviation Reportu Plant Revicu Co :mittec Meeting Minutes Operations Memo O Comments and/or Descriptions ((F)[ d.d k k W 2

_ Others l

NAME

' INITIALS DATE RP.AD NAME INITIALS DATE READ

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R*.'Doan (last)

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1.

Revicu of the attached material constitutes your initial exposure. A classroom 4

revicu vill also be hold for license holders as per amendment to 10CFR50.54.

2.

Uhen completed, fil'c in Training Coordinator Office files.

(Operator, Licensing Requirement) 3.

Use this form only for revicu of naterial rado necessary by license commitment.

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M Charlevoix, MI 49; i

RECORDS CONTROL FORM Attached, please find revisions / additions to t1: Big Rock Point llant Operating Procedures, Volume 3A/B.

Procedure Number R'evision Number / Date PRC Log Number EMP 3.5A 2

07/18/80 835-80 I have received the above mentioned documents, inserted in manual, and destroyed all previous issues.

CORRECTION: Page la should be included in the

" Insert Pages" section of ONP 2.9.

Please remove Signature the current LEP 'I aid replace it with the attached corrected LEP I.

~

Date CONTROLLED COPY HOLDERS 1 Docunent Control, BRP 21 Engineering 2 CJHartman 22 REBarnhart 3 CRAbel 23 RWDoan 4 Document Control, GO 24 RWDoan 5 Science Applications, Inc., PA 25 RWDoan 6 ACSevener 26 RWDoan 7 Shift Supervisor 27 RWDoan 8 Control Room /A 28 RUDoan 9 Control Room /B 29 RWDoan 10 RWDoan 30 RUDoan I1 DEDeMoor 31 RUDoan/VAAvery 12 DPBlanchard 32 RWDoan 13 JSnang 33 RWDoan 14 RESchrader 34 RUDoan 15 CEAxtell 35 RUDoan 16 JAJohnson 36 RWDoan 17' TCBordine 37 RWDoan 18 Science Applications, Inc., CA 38 Quality Control' 19 RBDeWitt, c/o P21-ll6 39 Relief Shift Supervisor 20 ARAbbs, P11-420A 40 Shift Technical Advisor-INCDRMATION COPY HOLDERS-m-

1 GCWright, URC 2

JLRucmin, P2 -604 3

JGrif fin, GE-BWR Training Center Route 2 on QA-05A

.QA-05A RUDoan QA-05A VAAvery QA-05A 'EMcNamara.

j

' SOP 26-6 ONP 2.16 only, Property '

-QA-05A - FJValade gA_o3 i

Proter. tion Supervisor 07/18/80 d

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