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Exam Repts 50-277/88-16OL & 50-278/88-16OL on 880719-21.Exam Results:One Senior Reactor Operator Candidate & Seven Reactor Operator Candidates Passed Exams.One Candidate Failed Written Exam & One Candidate Failed Operating Test
	ML20154A816
Person / Time
Site:	Peach Bottom
Issue date:	09/02/1988
From:	Howe A, Lange D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20154A760	List: IR 05000277/1988016; ML20154A755;
References
	50-277-88-16OL, 50-278-88-16OL, NUDOCS 8809130100
	Download: ML20154A816 (147)
	v • d • e

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U.S. NUCLEAR REGULATORY Com!SSION REGION I I OPERATOR LICENSING EXAMINATION REPORT l t

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EXMi! NATION REPORT NO. 50-277&278/88-16(0L) l

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FACILITY P0CKET NO. 50-277/278 l j FACILITY LICENSE NO. DPR-44 and DPR 56 !

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LICENSEE: Philadelphia Electric Company !

2301 Mar (et Street l l Philadelphia, Pennsylvania 19101 !

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FA0rLITY: Peach Bottom Units 2 and 3 f 3 EX#I! NATION DATES: July 19 to July 21, 1988 '

i CHIEF EXAMINcR: M, ,2),

Allen G. Howe, Senicf Operations Engineer bv @ /-ff Date ;

APPROVED BY: -, bd h .[or - f- 2 W i D6vid J./Lange, Chief, BWR 5ection. Fate ~~ *

I Operations Branch, Division of Reactor Safety !

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i I i SumARY: Written examinations and operating tests were administered te one !

(1) senior reactor operator (SRO) candidate and eight (8) reactor operator (

) (RO) candidates. A ninth R0 candidate took only the written examination. One i

! (1) SRO candidate and seven (7) RO candidates passed these examinations. One l l

candidate failed the written examination and one candidate failed the operating i

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8809130100 8S0902 DR ADOCK0500g7

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TYPE OF EXAMINATIONS: Replacement c EXAMINATION RESULTS:

U l R0 l SR0 I

, l Pass / Fail l Pass / Fail l l

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Written l 8/1 l 1/0 1

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1 Operating l 7/1 l 1/0 I L i I I I I I I I l Overal' l 7/2 1 1/0 l l l l l j l l I l i

1. CHIEF EXAMINER AT SITE: T. Walker, Senior Operations Engineer

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2. OTHER EXAMINERS: D. Lange, Chief, BWR Section T. Fish, Operations Engineer R. Miller, Sonalysts

M. Sullivan, Sonalysts e

3. The following is a summary of generic strergths or deficiencies noted on operating tests. This information is being provided to aid the licensee in upgr ading license and requalification training programs. No licensee response is reouNed.

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STRENGTHS t

For systems discussed while out in the plant: Ability to explain

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operation of system components.

DEFICIENCIES

Ability to give typical, expected values of major plant parameters for 100% power conditions.

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4. The following is a summary of generic strengths or deficit: s noced

from the grading of the RO written examinations. No summary was made on

! the SRO exair, as just one exam was administered. This information is being i provided to aid the licensee in upgrading license and requalification j -

training programs. No licensee response is required.

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-3-STRENGTHS a. Knowledge of thermal hydraulic limits, (Question 1.01). Ability to calculate cooldown rates, (Question 1.02). Knowledge of Shutdown Margin and the ability to predict changes in SDM due to changes in other parameters, (Question 1.09).

b. Knowledge of the plant's automatic response to a loss of a condensate pump at 100% power; condenser over-pressure protection, (Question 2.07). Knowledge of the response of the HPCI system to various component failures and the consequences of continued HPCI operation, (Question 2.08).

c. Knowledge of RPS response to loss of various power supplies, (Question 3.02).

d. Knowledge of the rationale behind a particular operator action in OT-114, "Inadvertent Opening of a Relief Valve", (Question 4.07).

Knowledge of the basis for two Cautions associated with TRIP procedures, (Question 4.10).

DEFICIENCIES a. Knowledge of how RPV level indication is affected by various operating conditions, (Question 3.01). Knowledge of the response of the Off-Gas system to a Hi-Hi Radiation alarm, (Question 3.03).

Knowledge of the conditions which generate an RPIS IN0P condition,

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(Question 3.06).

5. PERSONNEL PRESENT AT EXIT INTERVIEW:

Us Nuclear Regulatory Conunission Personnel D. Lange, Chief BWR Section, DRS T. Walker, Chief Examiner T. Fish, Operations Engineer L. Myers, Resident Inspector J. Linville, Chief Projects Section 2A Philadelphia Electric Company Personnel J. Franz, Manager, Peach Bottom F. Polaski, Assistant Superintendent of Operations K. Andrews, Supervisor, Operations Training J. Lyter, Operations Training R. Scheide, Compliance

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6. SUMMARY OF NRC COMMENTS MADE AT EXIT INTERVIEW:

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The chief examiner thanked the training and operations staffs for their cooperation during the examination. ,

The examiners felt site access was well coordinated and went snoothly.

The written examination review resulted in a few comments requiring resolution. The reviewers stated that the examination was a good test.

The generic strengths and weaknesses noted on the operating examinations were discussed. i

Attachments:

1. Written Examination and Answer Key (RO)

2. Written Examination and Answer Key (SRO)

3. Facility Comments on Written Examinations after Facility Review 4. NRC Response to Facility Comments

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$TT(4cHMGAJT l .

U. S. NUCLEAR REGULATORY COMMISSIGN REACTOR OPERAf0R LICENSE EXAMINATION F ACIL I T Y: _PEAGH_QQlIQM_gk3_.,______

REACTOR TYPE: _QWB-Og4_________________

DATE ADMINISTERED _60fgZ412________________

EXAMINER: _NBQ_ _65G19N_1__________

CANDIDATE: __ h[f_______________

INSIB99IIQUS_I9_G6NQ1Q81g1 Uno aeparate paper for the answers. Write answers on one side oniv.

Stcple question sheet on top of the answer sheets. Points for each quOction are indicated in parentheses after the ouestion. The passing gredo requires at least 70% in each categor y and a final grade of at 1coct S0%. Examination papers will be picked up six (6) hours after tho oxamination starts.

% OF '

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2ATEGORY % OF CANDIDATE'S CATEGORY

._YeLUE_ _I9IOL ___SG9BE___ _YeLUE__ ______________GGIEG9BX_____________

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.2Es22__ _20t99 ___________ ________ 1. PRINCIPLED OF NUCLEAR POWER PLANT OPERATION, THERM YNAMICS.

HEAT TRANSFER AND FLUI . '.OW M8 Mi __ _25t99 ___________ _..______ 2. PLANT DESIGN INCL.UDING 68.FETY AND EMERGENCY SYSTEMS 13.o r

.setww__ _25t99 ___________ ________ 3. INSTRUMENTS AND CONTROLS

.2Dt99__ 2Et9Q ___________ ________ 4. PROCEDURES - NORMAL. ABNORMAL, .

EMERGENCY AND RADIOLOGICAL i CONTROL

}. 0, exz1xx__ ___________ _______.% Totals Final Grade l All work done on this examination is my own. I have neither otven ;

nor e

REGION I l

08/07/19

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. 5. . THEGRY OF NUCLEAR POWER Pt. ANT OPERATION. FLUIDS. AND PAGE 44 THESt!D0Y148t! ICE ANSWERS '- PEACH BOTTOM 2&3 -99/07/19-USNRC - REGION I ANSWCR 5.01 (2.00)

6. Loco than 50% (0.5)

Xenon concentration is higher than just after the power reduction so.5)

8. GrCOter than 50% (0.5)

Xenon concentration is lower than just af ter the power reduction (0.5)

REFERENCE LOT 1510 p. 9 3.2 ...(KA VALUES)

292006K114 ...(KA*S)

ANSWER 5.02 ('.00)

c. 1. Increase (0.50) Since the incomin( water is colder, more heat ;

can be added to the coolant bef or's OTB occurs, therefore I the power at which transition boiling occurs will increase. (0.50)

2. Decrease (0.50) As pressure increases the amount of heat required fc vaporization decreases; theref ore, the bandle power required to cause transition boiling decreases. (0.50)

b. Tha Critical Power Ratio, CP/AP, will decrease (0.50) because an incruase in core flow results in a larger increase in the actual power of a uundle than the increaue in critical power of the bundle. (0.50)

REFERENCE LOT 1370 p. 10 3.3 3.2 ...(KA VALUES)

293OO9K122 293009K124 ...(KA*S) l

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I lu,J UE_ORY OF NUCLEAR POWER PLANT OPERATION._ FLUIDS._AND. PAGE 4]

IBERMOAYlfetilCE

%NSWERS -- PEACH BOTTOM 2b3 -88/07/19-USNRC - REGION I (1,50 WSWER 5.03 +3s00) ;

Avail. Required m. DECREASE REMAIN THE SAME 2 DECREASE RENAIN THE SAME c. INCREASE DECREASE (0.5 pts each)

gg/j,gy REFERENCE LOT 1290 3.7 3.3 3.4 3.6 ...(KA VALUES)

202001K101 202001K103 202001K105 202001K122 ...(KA*S)

ANSWER 5.04 (1.50)

To m31ntain the reactor critical without delayed neutrons, the rcactor would have to be prompt critical (0.5) (critical on prompt neutrons only). The average generation time of prompt neutrons 10 very short (10E-4 seconds) (0.5). Because of the short time between neutron generations, the reactor wculd not be controllable (0.5)

Alternate wording acceptable REFERENCE LOT 1420 3.7 ...fkA VALUES)

292003K106 ...(KA*S)

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5. . Tff0RY OF NUCLEAR POWER PLANT __ OPERATION. FLUIDS. ANQ PAGE 43 THEfm0DlNeti[CS ANSWERS '- PEACH BOTTOM 2L3 -88/07/19-USNRC - REGION I

ANSWER 5.05 (3.00)

Q:sume alpha T = -1 X 10E-04 (+ 5X 10E-05 / -4X 10E-4 ) (0.5)

i d31ta K 1* beta !

= ------ + ------------

(0.5)

K T*Keff (1+1ambda*T)

rho = insignificant + (0.007)/(1 + 0.1*100)

rho = 6.4 X 10E-04 delta K/K (0.5)

Th3 moverator temperature must increase to add enough negative rccctivity to overcome the 100 secono period. (-6.4 X 10E-04) (0.5)

(-6.4 X 10E-04 delta K/K) / (-1 X 10E-04 delta K/K)

(0.5)

deg. F change in Mod Temp

= 6.4 deg F change in Mod Temp. (0.5)

alter'# ate answers will be acceptable depending on the assumed value of alpha T

' REFERENCE LOT 1430 & 1440 2.0 3.2 ...(KA VALUES) :

,292003K100 292004K101 ...(KA'9)

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'Re.._IlfEORY (E_MJCLEfe POtfR PLANT OPERATION. FLUIDS. ANQ PAGE 47 T1!ERf10DYNAMILE M RS J- PEACH BOTTON 2s3 -89/07/19-USNRC - REGION I

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ANSWER 5.06 (2.50)

O. veid coefficient ,

(0.5)

D. moderator temperature coefficiont (0.5) i

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!c. vrid coefficient (0.5)

ld. doppiar coefficient (0.5)

0. modcrator temperature coefficient (0.5)

REFERENCE LOT 1440 3.2 ...(KA VALUES)

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292OO4K110 ...(KA*S)

ANSWER 5.07 (2.00)

i lo. 295 deg F (+- 15 deg F) (0.5)

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4d. 450 psia (+- 50 psia) (0.5) ,

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3.1 ...(KA VALUES) i

' 293OO3K123 ...tKA'S) '

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"J . , THEORY OF NQCLEAR POWER PidNT OPERATION. FLUIDS. ANQ PAGE 43 i TBERMOQ1NAMICR l .

ANSWERS '- PEACH BOTTOM 2&3 -88/07/19-USNRC - REGIOri I (

ANSWER 5.08 (1.50)

c.(0.75) decreased 2 phase flow resistance, so the system flow PC010tance decreases. (0.75)

REFERENCE LOT 1860 3.7 ...(KA VALUES)

202OO2K103 ...(KA'S)

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ANSWER Y[. 09 (2.00)

c. 295 deg F (+- 35 deg F) (0.5)

b.' Increase (0.5)

c. Increase (0,5)

d. 450 psia (+- 50 psia) (0.3)

REFERENCE LOT 1150,1160 3.6 3.7 3.6 3.7 ...(KA VALUES)

218000A101 218000A302 218000A406 239002K406 ...(KA*S)

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h_ IMQRY DE NUCLEAR POWER PLANT QPE)38 TION. FLUIM._ ANQ PAGE 49 THERMQDlNet1LCS ANSWERS d- PEACH BOTTDM 2&3 -CS/07/19-USNRC - REGION 1 ANSWER 5.10 (2.00)

c. Tru3 (0. 5 )

b. Folce (0.5)

c. Folce (0.5)

d. Trug (0.5)

REFERENCE PBAPS LOT 970 p.8

'LO 970-4 KQ 292OOOK103(4.1) 292OOSK104(3.3)

292OOOK103 292OOOK104 ...(KA*S)

. ANSWER 5.11 (2.50)

93. Decrease (0.50), high pressure fluid will mix with low pressure 4 fluid and lower the temperature dif f erence between the cooling medium and the cooled medium OR decrease in mass flow rate (0.75)

'b. Increase (0.50), as the loads on the RBCCW system increase, more hoct will be added to the RBCCW water in the form of sensible heat (temperature of the RBCCW water will increase). This will increase l tho temperature di f f erenti al across the heat exchanger fluids. (0.75)

l REFERENCE LOT 1240 291006K104 291006K108 293OO7K106 ...(KA*S)

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k.._ PLANT SYSTEMS DESIGN. CONJROL. AND INSTRUMENTATION PAGE 50 ANSWERS -- PEACH BOTTOM 2&3 -08/07/19-USNRC -

REGION !

ANSWER 6.01 (2.00)

c. First Stage Pressure < 30% of rated pressure (126 psig)

b. Mode Switch in Refuel, Shutdown, or Startup also accept Mode Switch not in Run c. Mode Switch in Refuel or Shutdown and Bypass Switch in Bypass d. Mode Switch in Run with APRM 'A' not downscale (0.5) per answer REFERENCE LOT 0300 Fig. 0300-4 LOT 0300 L.O. 8

0.1 ...(KA VALUES)

212OOOK412 ...(KA'S)

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Of. PLANLLYSTEMS DESIGN._CQti1EL AND INSIBUMENTATION PAGE 51 ANSWERJ '- PEACH LOTTOM 2&3 -98/07/19-USNRC - REGION I ANSWER 6.02 (2.50)

o. 0.66W + 41 High 0.66W + 33 Intermediate 0.66W + 25 Low l

0.60W + 60 (cl amped at 107%) Backup l (0.25) each b. 1. yes (0.25)

2. no (0.25)

3. yes (0.25)

c. 1. no (0.25)

2. yes (0.25)

3. no (0.25)

REFERENCE LOT 0200 pp. 5-B L.D. 4,5 2.0 3.5 ...(KA VALUES)

215002A403 215002K401 ...(KA*S)

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4.. PLANT SYSTEMS DESIGN. CONT 80L. AND INSTRUMENTATIQM PAGE 52 ANSWERS s- PEACH BOTTOM 2&3 -88/07/19-USNRC - REGION !

ANSWER 6.03 (1.50)

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Trip the Main Generator Output Breakers Trip the Exciter Field Breaker Trip the Main Turbine Transf ers voltage regulator from Auto to Manual ,

Trip auxi!!ary switchgear main circuit breakers l Fcat transfer of unit auxiliary switchgear l Trip D & E cooling towers Trip reactor recirculation pumps if powered frem unit Trip main and aux transf ormer f ans and oil pumps Trip the stator cool ant pumps (5 required S 0.3 each)

i REFERENCE LOT 0600 p. 23 3.1 ...(KA VALUES) s 262OO1K404 ...(KA'S) '

ANSWER 6.04 (3.00)

c. Tho TCVs will close to 50% flow position (0.5)

Tho TCV low value gate passes a MCF signal of 50%

rcther than the uignal from the pressure controller (0.5) ;

b. 'fbo BPVs will remain closed through the transient (0.5) '

tho MCF summer will send a zero signal to the BPV LVG (0.5)

'c . Racetor power (and pressure) will rapidly increase following (0.5)

tho fault.

Tho reactor will scram on High Flux and/or high pressure (0.5) ;

beccuse of the closure of the TCVs ;

REFERENCE LOT 0590 pp. 12-15 3.7 . . . ( K A V AL' !F 3 )

245000K602 ...(KA*S)

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h. PLANT SYSTEMS DESIGN. CONTROL . AND INRIBW1EMIeIJQN PAGE 53 ANSWER 3 *- PEACli BOTTOM 2L3 -88/07/19-USNRC - REGION I ANSWER 6.05 (3.00)

indiccted level will initially c. incrense (0.5)

3. remain the same (0.5)

3. increase (0.5)

d. increase (0.5)

D. increase (0.5)

f. Cecrease (0.5)

REFERENCE LOT 0050 3.2 3.1 3.5 3.4 3.5 4.1 3.6 3.8 3.3 ...(KA VALUES)

216000A201 216000A203 216000A207 216000A209 216000A210 216000K324 216000K506 216000K507 216000K512 ...(KA*S)

ANSWER 6.06 (2.00)

c. 1. diesel speed (frequency) (0.5)

2. load control (0.5)

b. 1. voltage control (0.5)

2. VAR control (0.5)

REFERENCE LOT 0670 3.6 3.1 3.4 3.9 3.4

...(KA VALUES *,

2SCOOOA201 264000A304 264000A401 264000 GOO 9 264000K505

...(KQ'S)

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4t . PLANT SYSTEMS DESIGN. CDtORQ k AND INSTRUMEt[TATION PAGE 5'S ANSWER 3 '- PEACH BOTTOM 2h3 -08/07/19-UCNRC - REGION I ANSWER 6.07 (2.00)

c. Rod 32-35 must be inserted (1.0)

6. 3 (1.0)

REFERENCE LOT 0090 3.5 3.5 3.5 3.4 ...(KA VALUES)

201006A205 201006K401 201006K402 201006K403 ...(KA*S)

ANSWER 6.08 (1.50)

1. decay of Xenon 2. make the reactor subcritical from 100% power 3. allow for uneven mixing 4. overcome not positive reactivity due to cooling down 5. maintain at least 3% shutdown margin (5 required e 0.2 each)

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REFERENCE LOT 0310 pp. 3,4 L.O. 3 4.1 3.6 3.9 ...(KA VALUES)

211000 GOO 4 211000K105 211000K407 ...(KA*S)

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4.. PLANT SYSTEMS DESIGN. CONTROLe AND INSTRUMENTATION PAGE 55 ANSWERS ~'- PEACil BOTTOM 28,3 -

-88/07/19-USNRC - FEGION I !

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l ANSWER 6.09 (3.00)

76 Cog C Alarm (Stator liquid in/out high temperature) (0.5)

G1 C:q C decreases load limit to 25% (0.5)

Trips 'A' Recirc Pump after 1 second (0.5) e Trips 'B' Recirc Pump after 10 seconds (0.5)

trips turbine if generator amps are not below 26,530 (0.5) !

in two minutes l trips turbine i f generator amps are not below 7,726 (0.5)

in three and one half minutes ;

i REFERENCE l LOT 0460 pp. 6,0 L.O. 6 !

2.9 3.5 :

845000K605 245000A312 ...(KA'S) !

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4.. PLANT PYSTEMS DE N QtGBQL. AND__'iNSTRuttENTATION PAGE 53 ANSWERS * PEACH BOTTOM 2&3 -80/07/19-USNRC - REGION I ANSWER 6.10 (2.50)

a.1. The level control system sees a mismatch between feedwater flow cnd steam flow. (0.25) The system increases feedwater flow in an attempt to match feedwater flow snd steam flow (0.15)

Lovel star ts to increase causing a level error signal (0.10)

Lovel will stabilize several inches above the reference point where the level error will equal the steam flow / feed flow cismatch error (0.25) Power will not change (0.25)

a.2. The Level control system sees a large level deviation (0.25)

cnd increases feedwater flow (0.25) Actual water level increases to the turbine trip level (both main turbine and feedpump turbine)

(0.25). The reactor will scram due to a turbine trip if power is ~

) 30% or due to a low level after the feedpump turbine trip if power is < 30% (0.25)

u. (This indicates that the lockout relay does not have power available)

cnd therefore a lockout cannot occur (0.5)

i REFERENCE I LOT 0550 pp. 15,16 L.O. 10,16 3.6 3.1 3.4 4.3 3.0 I

...(KA VALUES) !

259002K104 259002K604 259002A202 259001A309 259002A310

...(KA*S)

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h a.. PLANT SYSTEMS _DE119N d QttTROL 3 AND _INSTRutENTATIQM PAGE 57

%NSWERS -- PEACH BOTTON 2&3 -BW/07/19-USNRC - REGION I l

4NSWER 6.11 (2.00) ;

m. 'A' fan starts, (0.25) both filter train damper sets open (0.25)

3 'C' fan starts, (0.25) both filter train damper sets open (0.25)

c. 'B' fan starts after 20 seconds, (0.25) !

b th fi. iter train damper sets will remain open (0.25)

(thould open from initiation signal to fan 'C')

3. No response (0.5) ;

4EFERENCE LOT 0210 p. 6 L.D. 2 3.0 ...(KA VALUES)

231000K401 ...(KA'S)

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7 PRQCEDURES - NORMAL. ABNQfh_Et!EflQENGL8MQ PAGE 58 B&D19L001 CAL _GQMIBOL ANSWERS -- PEACH BOTTOM 2&3 -99/07/19-USNRC - REGION !

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ANSWER 7.01 (2.50)

i D 35 mrom Gamma 5 mren Beta 10 mead Epithermal neutrons 10 X 3 (QF)= 30 (O.25)

Tctc1 dose 70 mrom (0.251 b Worker "A" limit 2500 mrem / quarter (0.50)

Prccent dose 2100 mrom

2500 - 2100 = 400 mrom (0.25)

400/70 = 5.71 hours8.217593e-4 days 0.0197 hours 1.173942e-4 weeks 2.70155e-5 months or 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 42 minutes (0.25)

Worker "B" limit 1000 mrom/ quarter (0.50)

PrcCont done 800 mrom 1000 - 800 = 200 mrom (0.25)

200/70 = 2.86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months or 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 52 minutes (0.25)

' REFERENCE

.LGT-1730 B1 and B4 LCCrning Ob.jective LOT 1730 -2 3.8 294001K103 ...(KA'S)

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7.-._PROCEDUBEB - NORMAL, ABNG8M8L, EMERGENC( AND

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PAGE 57 88DlDLQGlCAL.CONTRQL ANSWERS -- PEACH SOTTON 2b3 -98/07/19-USNRC - REGION 1 l

ANSWER 7.02 (3.00)

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o T-101 - - C (0.50)

D T-102 (0.50)

,,+.

c T-101 (0.25)' AND T-102 ( 0. L )

d T-101 (0.50)

o T-102 (0.50)

f T-101 ' ~ --

(0.50)

REFERENCE Cyatematic EDP Flow Path T-101 Lcarning Objective LOT 1560-9 4.5 4.5 4.6 4.7 2950250011 295024G011 295031GO11 2950370011 ...(KA*S)

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Z.c PROCEDi.*lEhEQflM6L, ASNQht)6kKMERDENCY AND PAGE CO 880LDLQGIE8L_EQNIBDL

' ANSWERS '- PEACH BOTTOM 2h3 -88/07/19-USNRC ~ REGION :

ANSWE3 7.03 (3.00)

'o Th3 reactor is not shut down when the reactor is not subcritical (0.50)

(Th3 definition does not reference any power to allow the operator to inject boron during any condition in which the reactor is not shutdown

'on rods including power leveth less than three percent.)

b Fo * the worst case ATWS, the Heat Capacity Temperature Limit (HCT'. ) '

,

cf the torus may be exceeded bef ore the reactor is shutdown by SLC.(0.50)

'

The operator will reduce level to reduce power (ard therefore, the cmount of heat rejected to the torus) to maintain the torus temperature b310w the HCTL. (0.50)

c. Th3 maximum injection time produces the specified minimum concentrate.on an approximately 123 minutes, which is substantially 1cos time than the cooldown time which will take several hours. (0.75)

d Th3 minimum injection time allows for sufficient mixing so the boron doesn't recirculate through the core in uneven concentrations which could cause power excursions (0.75)

CREFERENCE

' REFERENCE LOT-0310 p 6,7 TRIP 'i-101,T-102 (BASES)

4.5 4.3 4.5 4.2

295037A104 29bO37 GOO 3 295037K204 211000A208 ...(KA*S)

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P.. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY A G ' PAGE 61 RADIOLQQ1 GAL _ CONTROL

%NSWERS ' PEACH DOTTOM 2&3 -88/07/19-USNRC - REGION I

,

ANSWER 7.04 (2.75)

a 1 Frcsh Air Supply Fans (OAV-79 and OBV-79) trip. (0.25)

2 Air Conditioning Supply Fans (OAV-28 and ObV 28) trip (0.25) s 3 R; turn Air Fans (OAV-29 and OBV-29) trip (0.257 O Emergency Vent Fans (OAV-30 and OBV-30) trip (0.25) ;

5 Tcitet Exhaust Fan (DOV-30) trips (0.25)

(Either the fan numbers or a functional word description of the fans cro ccceptable as answers)

b 1. control room radiation levels exceed 300 mrom/hr (0.75)

2, rcspiratory equipment is required in the control room for more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (0.75)

REFERENCE L Of f Normal Procedure ON-115 LOT-1550 LCcrning Objective LOT 1550 #1 ;

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3.1 3.4 395016K203 295017K207 ...(KA*S) ,

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%,. PROCEl d S - NQB & _J3RNQFf h*Et!EBGENCY_6NQ PAGE 62 i

R803QLQQ1CfR_C.QNIBA

48WER9 4 PEACH BOTTOM 28 3 -98/07/19-USNRC - REGION !

l l

l M49WER 7.05 (2.00)

R pid depressurizatior, of the reactor will result in injection of 1crge amounts of cold unborated water, resulting in an uncontrolled i power increase. (1.00) ,

)

CRD is left on because under these conditions, the operator should

) to trying to insert rods. (1.00)

REFERENCE EB Flow Paths T 101 and T-112 EB T-112 bases LCCrning Objective LOT 1560 #3 4.4 4.1 3.8 3.9 295015G012 295015K102 29501SK104 295012K201 ...(KA*S)

Le PROCEDURES - NQfit$8LuBRNNMERGENCY AND PAGE 63 R$DIQLOGICAL .CQtLTBOL i

WWERG '- PEACH BOTTOM 2&3 -88/07/19-USNRC - REGION I ANSWER 7.06 (3.00)

0 1 Minimizes plant transient on subsequent scram. (0.50)

2 Provides an additional signal to maintain the Recirc Pumps at cini mum speed, should the 30*/. speed limiter circuit fail, following Cardox discharge. (0.50)

D 1 Provents Recirc Pumps from tripping during the automatic 13 KV trcnsfer that normally occurs after a scram and turbine trap.

(0.50)

2 Accures continuous power supply to all house loads without roliance on automatic breaker operation (0.50)

c 1 Minimizes plant transients as systems become uncontrollable fc11owing Cardox discharge. (0.50)

2 Tho Safety Analysis for Automatic Cardou Discharge in the Cable Cpreading Room, assumes the units are shut down prior to diccharge. (0.50)

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REFERENCE SE-2 Cases 4.2 3.7 295016K301 295016K303 ...(KA'S) !

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L NDLBEB - NORMAL, ABNQEMAL. Ef1EBGENCY AND PAGE c4

.58EIDL901C8LGQNIRQL ANSWERS *- PEACH BOTTOM 2k3 -88/07/19-USNRC - REGION I N 7.07 (2.50)

3 O tacond licensed operator, (0.25) with no other duties, (0.25)

shall verify the correct rod is being moved, to the correct position, in the required sequence. (0.25) by using the procomm computer. (O.25)

3 Th3 Reactor Operator, Second Licensed Operator and the Control Room Superintendent (0.25) shall INDEPENDENTLY verify the control rod pcttern (0.25) by comparing the OD 7 Option 2 to the GR-2 Appendix

! core map. (0.25)

e Th3 Reactor Operator, Second 1,1 censed Operator and Control Room Shift Superintendent (0.25) shall sign the OD 7 Option 2 and the GP-2 Appendix ! core map (0.25) and attach it to the appropriate coro map in GP-2-2. (0.25)

REFERENCE GP-2-2' Appendix !

OT 1530 ccrning Objective LOT 1530 #4 3.5 3.8 3.2 201002K105 2010020001 201002G013 ...(KA*S)

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' ANSWER 7.06 (2.00)

o Th3 3 times normal background scram also closes the isolation l volves which removes the heat sink. (0.50) Scramming the reactor grior to the isolation allows the condenser to be used f or cecay l

hoct di ssipation. (0.50)

D Reduce reactor power until the radiation level is below 1.5 time norcal background. (0.50) This reduction below the alarm level occures the release rates will be acceptable. (0.50)

REFERENCE OT-103 BASES LOT-1540 Locrning Objective LOT-1540 B-2 3.7 3.9 223002G014 223002K101 ...(KA*S)

ANSWER 7.09 (3.00)

o Per DN-107 the Scram is required if three or more CLD accumulator low procsure alarms are received. (0.50) The scram is inserted prior to cccumulator depressurization beacause the accumulators are required to assure adequate scram speeds on the rods. (1.00)

b Per DN-107 the Scram is not required. (0.50) With reactor pressure absve 550 psig, the Scram is required only when all HCU accumulators cicrm (1.00)

REFERENCE LOT-70 LCarning Objective LOT-70 #2 3.4 3.9 295022K101 295022K301 ...(KA'S)

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7.. PROCEDURES - NORMAL. ABNOFMAL. EMERGENCY ANQ PAGE 63 RADIOLr&ICAL CONTROL ANSWERS -- PEACH BOTTOM 2&3 -08/07/19-USNRC - REGION I ANSWER 7.10 (1.25)

c. Thace status boards must be updated after EACH FUEL ASSEMBLY RELOCATION (0.2"$

D. Surpend 4uel handling operations (0.50)

Evccuate the refuel floor (0.50)

REFERENCE FN-6C 3.0 3.9 23t,0000001 234000G002 ...(KA*S)

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0,- AQ&lNISTRATIVE PRQCEQUBES. CONDITIQNEdND LIMLIAUQNR PAGE 67 ANSWER 3 k PEACH BOTTOM 2&3 -88/07/19-USNRC - REGION I ANSWER 8.01 (2.75)

3 Any area, accessible to personnel, where there exists radiation at Cuch levels that a major portion of the whole body could receive in ono hour, a dose in excess of 1000 mREN. (0.50)

b A LEVEL !! Locked High Radiation Area has a dose rate in Cv' 'as of 30 Rem per hour, or has the potential for an Cxtre_.ly high dose rate (0.50)

O LEVEL I Locked High Radiation Area is a Locked High Radiation ARCO not considered a LEVEL !! Locked High Radiation Area (0.50)

c 1 Tha high radiation Master Keys are kept in the custody of; Cocurity Supervision (0.25)

Shif t Supervision (0.25)

Health Physics Supervision (0.25)

2 M20ter Keys are used for emergencies only. (0.50)

REFERENCE h0T1570 Ad2inictrative Procedure A-84 LCarning Objective LOT 1570-5-m 3.8 3.7 2?C001K103 294001K105 ...(KA*S)

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8.. ADMINISTRATIVE PROCEDURES._ CONDITIONS. AND LIMITATIQNE PAGE 68 ANSWERS *- PEACH BOTTOM 2&3 -98/07/19-USNRC - REGION 1 ANSWER 8.02 (2.50)

3 Th3 operator can continue to work for the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (0.50)

If he worked more than four (4) hours he would exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in Coven (7) days (0.50) and, he would exceed working more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . (0.50)

b 3 Q Personnel Staffing Deviation Form shall be filled out. (0.50)

2 ThD overtime shall be authorized by the Stat 5 -9 perEnkendent, this alternate, or higher level of management.) (0.50)

REFERENCE Adminictrative Procedure A-40 botrning Objective LOT 1570 3-h 4.5 290001A011 ...(KA'S)

ANSWER B.03 (1.00)

3 Ancign a Utility Shift Supervisor, or Utility Shift Superintendent to be the team leader. (0.50)

@ R311 eve the Control Room Supervisor so he can become the team loofer. (0.50)

REFERENCE Abnorc31 Procedure ON-114 Ad inictrative Procedure A-7-6 bCcening Ohjuctive 1550-1 4.2 290001Q112 ...(KA'S)

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h_ ADM1R121BMIVE PJQCFAQf1ERuqQNQllLQNS. AND _L1MITATIDNS PAGE 69 ANSWER 3 ' PEACH BOTTOM 2&3 -8F./07/19-USNRC - REGION I ANSWER 8.04 (3.00)

O Th3 procedure change must not alter the intent of the procedure. 1 (0.50)

) Tho change must be approved by two (2) members of the Plant t M:ntgoment Staff (0.50) one (1) of whom holds an SRO license (0.50)

c Tho change must be approved by PORC (0.50) and approved by the M:ncqer Nuclear Plant (0.50) within 14 days. (0.50) l REFERENCE 7echnical Specification 6.8.3 LOT-1570 !!.C LCcrning Objective LOT 1570-1, 3a 4.2 3.7 294001A101 294001A103 ...(KA*S)

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8. ADMINISTRATIVE PROCEDURES.. CONDITIONS, AND LIMITATIONE PAGE 70 ANSWERS J- PEACH BOTTOM 2&3 -88/07/19-USNRC - REGION I ANSWER 8.05 (3.00)

' c. Licit applies any time the pl ant is in STARTUP, HOT STANDBY, OR RUN mode. (0.50)

b. 105 deg F. (0.50)

c. Stop testing (0.125) reduce temperature to less than 95 deg F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (0.125) OR be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (0.125) and COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (0.125)

d. Licit applies during reactor isolation conditions. (0.50)

O. 110 deg. F (0.50)

f. Scram the reactor. (0.50)

REFERENCE

Technical Specification 3.7.A 3.6 3.8 4.0 4.1 4.0 4.3 i223001A109 223001A212 223001G001 223001G005 223001G006 295026G003 ...(KA'S)

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d.* eDU tlphDM UR thut EDU%s u_ pi lt> t a l.up a ,_ ruo L tiju G1 totw rnes ' t

- MiGWERG -- PEACH BOTTOM 2&3 -88/07/19-03NRC - REGION I ANBRER R.06 (2.50)

a 1 Single loop operation not permitted in Region I to prevent thermal hydraulic instabilities (0.50) which could cause neutron !

flux oscillations (0.50)

i 2 Reduce reactor power to less than the power described by line "A" on figure 3.6.5 (0.125) OR increase flow to 39% or greater (0.125)

within four (4) hours. (0.25)

b 1 After entering Region I. determine APRM and LPRM noise levels (0.25) within one (1) hour (0.25) and at least every eight (8)

, hours thereafter. (0.25)

2 After increasing power by at least 5% (0.25) determine APRM and LPRM noise leveis (0.25) within one (1) hour. (0.25)

REFERENCE

. Technical Specification 3.6.F f.ot 0030

Learning Objectiv.* LOT 0030 SRO 2 l 1 3.7 4.2 202001A203 202001GG11 ..(KA*S)

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8. - ADMINISTRATIVE PROCEDURCS,. CONDITIONS. AND__W NITAT!QME PAGE 72

l ANSWERS *- PEACH BOTTOM 2&3 -88/07/19-USNRC - REGION I ANSWER 8.07 (2.50)

i

'o Control rod #1 is INOP (0.25) and must be inserted and disarmed. '

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(0.25)

Control Rod #2 is INOP (0.25) but not required to be disarmed or incerted. (0.25) i i

b Control rod #3 by Technical specification, is INOP (0.25) and requires a reactor shutdown wi thin 48 hou. s. (0.50)

(or investigation must demonstrate that the failure is not due to o failett control rod drive mechanism collet housing) '

Control rod #3 by Technical dpecification, does not meet the 5 X 5 ,

crrcy (4 operable rods between inoperable rods) with control rod #2 '

(0.25) and requires a reactor shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (0.50)

REFERENCE Technical Specification 3.3.A.F LOT-OO70 Lccrning Objective LOT-0070 #1 and #2 4.2 3.9 !

201001G005 201001G011 ...(KA*S)

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a 1 Tha plant conditions and degree of accident control.

2 Emergency classification.

3 Tho potential for emergency escalation.

4 Stcff augmentation and technical assistance requirements.

5 Tho need f or long term support.

(Three required at 0.50 ea)

b ALERT Invel. (0.50)

c OITE EMERGENCY level. (0.50)

REFERENCE

'ERP-2OO Emergency Director SEC 2.1.5, 2.1.6 and 2.1.7 4.7 290001A116 ...(KA*S)

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8.- ADMINISTRATIVE PROCEQQBES; CONDITIONS. AND L IMITATIONE PAGE 74 ANSWER 3 '- PEACH BOTTON 2&3 -88/07/19-USNRC - REGION I

ANSWER 8.09 3&p (2.25) -

FG11uro of APRM "A" results in less than two (2) instrument channels for trip channel "A". (0.25) The Technical Specification requires two *

(2) inctrument channels for each trip channel. APRH "E" is already bypccccd, so APRM "A" cannot be bypassed.(0.25) Trip system "A" shall b3 picced in the tripped condition. (0.50)

APRM."A"_is_ also the normal reference APRM for RBM "B" (0.25L. SinceTjpg APRM "E" i a tWe' alternate _ref erence APRM_f or-ROM-"B"', RBM "B" ia INOP.

(0.50) A seven (7_Lday-LCO-wasTM f ect-due, to RBM "B" bei ng INOP.

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RSM *C"~1m'E6 M NOP and shall be placed in a tripised' condition d o.50) '

Me L ee's 4ec Ae. R. B M, (,!.e) ~~

REFERENCE Technical Specifications 3.1 and 3.2 3.0 3.0 3.3 4.3 3.5 3.3 215002K101 215002K604 215002A203 215002G011 215005K103 215005K307 ...(KA*S)

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G.- ADMIN 1BTRATIVE PROCEDUBESdQNDITIONS. AND LIMITATJDtM1 PAGE 75 ANSWER 3 -- PEACH BOTTOM 2&3 -88/07/19-USNRC - REGION I ANSWER 8.10 (3.00)

o 1 A Deficiency Tag (or sticker) is used on nuclear safety related cquipment, to indicate a compon**nt deficiency has been identified.

(0.50)

2 Tho tag is placed when an MRF has been written BUT the component hco not been blocked and the deficiency is not obvious. (0.25)

3 Th3 tag chall be removed when the condition is known to have been c orr ec t ed . (0.25)

b 1 A Manila Information Tag may be used to note useful inf ormation in tho operation of the plant (0.50). (It may also be used in place of o Deficiency Tag, if there is not enough room to write all the inf ormation on the Deficiency T(.J. )

2 It is placed when the deficiency is identified, or a need f or more inf ormation is required. (0.25)

3 It shall be removed when the condition is known to have been c or r ec t ed . (0.25)

c 1 Tho Operation Verification Tag, is hung as a reminder that the tocting requirements which prove operability (required by section 7 of the NRF), must be def erred due to plant conditions. (0.50)

2 Tho tag is hung af ter the MRF has been completed, (except for cec t i on 7, in order to clear the blocking permit.) (0.25)

3 Tho tag is removed when the operational verification has been completed as documented on the MRF. (0.25)

REFERENCE Ad2inictrative procedure A-26, Procedure f or Corrective Maintenance p;c 7 cnd 8 4.5 294001K102 ...(KA*S)

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RTTecMMNT 3 i e PHILADELPHIA ELECTRIC COMPANY PEAcil BOT 10M ATOMIC POWER STATION R.D.7.Scx20N E w Delta, Penm)lvania 17314 esua sonois-rwa roe sn oe suntiva (717) 45G7014 D. M. Smith l Vice Prtudent i I July 25, 1988 l

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Mr. Robert M. Callo, Chief

Operations Branch Division of Reactor Safety !

U.S. Nuclear Regulatory Commission d

Region I i i 475 Allendale Road King of Prussia, PA 10406 j

Dear Mr. Callo:

[

SUBJICT: Facility Comment on License Examination (Report #88 16) !

Administered at Peach Bottom Atomic Power Station July 19, 1988

) The attachment to this letter documents the complete formal comment [

,

summary of the Senior and Reactor Operator License Examinations administered !

on July 19, 1988. [

. In the vast majority, comments have been limited to those questions i and/or answers which were specifically addresaed during the post examination e review session conducted on July 19

.l In the majority of cases, the referenced supporting documentation can t i be found in the materials forwarded to your office for exam preparation. In j l several cases specific references are not appropriate due to the nature of l

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! the reviewer's comments.

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Sincerely, f

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! DMS/RGA:bgh (

j Attachments

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QUESTION 1.07 (3.00)

The Residual Heat Removal pumps are being used in the Shutdown Cooling Mode. HOW will AVAILABLE and REQUIRED Net Positive Suction Head (NPSH) for the RRR pumps be affected (INCREASE, DECREASE, or NOT AFFECTED) by each of the following changes?

NPSH NPSH AVAILABLE REQUIRED a. Reactor water temperature increases

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b. Reactor water level decreases j

c. RRR system flowrate decreases

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ANSWER 1.07 (3.00)

i AVAILABLE REQUIRED a. decrease not affected b. decrease not affected

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c. increase decrease

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(0.5 pts each)

., REFERENCE

PBAPS LOT 1290 t

i KA 202001K101(3.6) 202001K103(3.2) 202001K105(3.4) 202001K122(3.4)

j 202001K101 202001K103 202001K105 202001K122 ...(KA'S)

FACILITY COM2 TENT:

Required NPSH is not taught within lesson plan subject matter or listed as an Objective. Many candidates may assume "required" varies inversely as

"available" and answer accordin5 17-

REFERENCE: LOT 1290 I

l RESOLUTION: Reduce weight of "Required NPSH" portion to 0.25 pts each, therefore j reducing total question value to 2.25.

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QUESTION 2.07 (2.25)

a. While operating at 1004 power, a condensate pump trips. STATE the two (2)

automatic actions which occur directly because of this trip and STATE the purpose for these actions.

b. h0V is the main condenser protected from an over pressure condition?

Include applicable setpoints.

.

ANSVER 2.07 (2.25)

a. Recirculation pumps runback to 604 speed (75% flow) (0,5) and a 90t maximun speed signal to the feed pumps is inserted. (0,5) This action reduces the feedvater requirements e.> a point that can be handled by two condensate pumps. (0,5)

b. Two rupture diaphragas are provided on rach LP turbine exnaust shroud, (0,5)

set at 5 psig. (0.25)

REFERENCE PBAPS LOT 520 pgs 5, 17 LO 520 10, 11 256000K304(3.6) 256000G007(3,4)

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60000007 256000K304 ...(KA'S)

FACILITY COMMENT AND RESOLUTION 2,07 b. Other means are available for main condenser o'.or pressi.re protection and should be accepted. Such items could include automatic acticus as condenser vacuum decreases. For example, reactor scram, esin turbine trip, reactor feed pump turbine trip and bypass 'rcive open permissive removed.

REFERENCE 1. PBAPS LOT 0300 page 12 2. PBAPS Technical Specifications page 23 3. PBAPS LOT 0590 page 15 '

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QUESTION 2.09 (2.50)

For each ot the following plant conditions, STATE (YES/NO) if ALL RUNNING dryvell ventilation fans will trip off: (Consider each condition separately.)

a. 1004 power, D/V temperature is 138 F.

b. Rx is shutdown. Rx water level is 135 inches, c. DV pressure is 2.1 psig, Rx vater level is +38 inches, d. 504 power, torus bulk water temperature is 107 F.

e. 1004 power, m2 13.2KV bus is lost.

ANSVER 2.09 (2.50)

a. no b. n3 C. ye5 d. no e. no REFERENCE PBAPS LOT 140 P.C LO 140 3 KA 223001K403(3.7) 223001K611(3.0)

223001K403 223001K61.1 ...(KA'S)

FACILITY COMMENT AND RESOLUTION 2.09 b. Esther yes or no or both should be acceptable ansvers since the LOCA signal of triple low reactor vessel water level is 130" for Unit 3 and-160" for Unit 2.

REFERENCE PBAPS LOT 0140 page 6

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QUESTION 3.01 (2.50)

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a. For each of the following parameter changes and operations 1 conditions, l STATE if the INDICATED LEVEL will INCREASE, DECREASE, or REMAIN THE SAME for '

the cpecified level instrument if ACTUAL level REMAINS THE SAME.

.

1. The 1,/W temperature increases about 200 degrees. How will the NARROW RANGE level instrumentation respond? )

2. The D/V temperature increases about 45 degrees. How will the VIDE

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RANCE level instrumentation respond? ;

3. A reactor startup is in progress. The head vent is closed. Vessel ,

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, temperature and pressure are increased from atmospheric and 220 degrees to 800 psig and 518 degrees. How will the NARROW RANGE level instrumentation respond? :

i 4. A reactor startup is in progress. The head vent is closed. Vessel I

temperature and pressure are increased from atmospheric and 220 degrees

! to 800 psig and 518 degrees. How will the VIDE RANGE level

!.nstrumentation respond? !

l b. VHAT level instrument (s) is/are responsible for initiating the main turbine -

and RFPT trips at +45 inches?

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I ANSWER 3.01 (2,50)

s. 1. Increase (0.5) l 2. Increase (0,5) i I 3. Remains the same (0,5) {

4. Decease (0,5)

l j b. Feedwater Control (narrow range) (0.25) and [

Yarways (vide range) (0.25)

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- REFERENCE ;

! PBAPS IDT 50 pgs.14,17 18, 26 l 1 :

1 14 50 4, 6 !

l l KA 216000K501(3.1) 216000K507(3.6) 216000K510(3.1) 216000K113(3,4) ,

j 216000K116(3.0) ;

216000K113 216000K116 216000K501 216000K507 216000K510 l 3 ...(KA'S) l

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3 FACILITY COMMENT AND RESOLUTION (

l 3,01 a.2. P,emain the same is also an acceptable answer. The students were taught [

l MOD 1457 Yarway Level Modification which is installed in Unit 2. !

Changes in indicated level due to changes in drywell temperature are !

,

essentially eliminated. I i

)

REFERENCE PBAPS 14R-8704 page 6 3)a? f

!

l l

2 !

a f

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LOR 8704 R:v. 1 RJA/cjo Page 1 of 45

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PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POVER STATION 1987 PBAPS OPERATOR REQUALIFICATION LECTURE SERIES LECSON PLAN IIIL'*

Unit 2 Modifications PURPOSE:

To familiarize RO/SRO uith modifications performed on Unit 2 prior to the hydrostatic test, OBJECTIVES:

Upon successful completion of this lesson, the trainee vill be able to:

1. Discuss the importance of the Yarvay level modification, 2. State whether a given instrument in the control roou is fed from the vide range compensated instrument, 3. State the conditions that vill cause the 18" vent and purge lines to

_) isolate.

4 List the alterncce power sources available to the Un'.t 2 RPS BUS.

5. List the conditions that vill cause an ADS blevdown on Unit 2.

6, State why the addition of the 9 minute lov level timer is important.

7. List the indications available on the HPCI ACS for /.DS.

8. Discuss why the change to the Unit 2 RHR minimum flow valves is important.

REFERENCES:

Associated modification packages and Shif t Training bulletins, MATERIALS:

Instructor:

1. Whiteboard, markers, and erasers 2. "ransparencies and Overhead '.'roj ec'.cr DURATION:

_

}

7/50 Minute Sessions s

_ _ _ _ _ _ _ _ _ _ _ _ _ _ ,

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LOR.8704 R:v. 1 RJ A/cj o Page 2 of 45 SUBJECT MATTER OUTLINE SUPPORT INFORMATION 1. Yarvay Level Modifications MOD 1457 A. purpose 1. To ieprove the accuracy and reliability of the water level mearv.<ement under accident, transient and nornal operating conditions.

2. Decreases the neen for operator diagnosis due to instrument inaccuracies.

B. Components Affected/ Changed 1. Removed the two Yarvay temperature compensated reference columns and the associated reference leg piping from the reference colus.ns to dryvell penetrations N 28 and N.29.

a. The piping through penetrations N 28 6 N 29 is capped on both sides of the penetration, g b. The variable leg piping is sloped continu.

7 ously down from the vessel nozzle to the remainder of the variable les piping.

2. Installed two new condensing chambers which are not temperature compensated, a. The piping is routed through dryvell penetration N 26.

b. The new condensing chambers are located *

above their respective reactor nozzles to allow proper sloping of the reference column piping to their penetrations.

3. Moved the two fuel zone water level reference legs from the curront CEMAC cold reference legs to the new cold reference legs.

4. Return the level 1 (Triple low) trip point to 160 inches.

5. Recalibrated the instruments connected te the modified refere::e legs to compensate for the revised elevstion of the condensing chamber and the new reference leg ambient conditions.

-.

.

LOR.8704 R;v. 1 RJA/cj e Page 3 of 45 SUBJECT MATTER OUTLINE SUPPORT INFORMATION 6. Four independent micro.processot coepensation instruments are installed in Unit 2 cable spreading room as follows:

a. Differential pressure signals from LT.2 3 72A,B,C.D HFCI, RCIC, ADS, C.S. 6 RRR initiation, and indication on the C05 panel 2185 LT 2 3 73A,5 -

Containment spray permis.

sive, LI 91A&B LT.2 3 73C.D* LR 110A 6 110B

LT.2 3 111A,B have been retagged LT.2 3 73C and D are connected free the instrument racks to the new compensation instruments. The Rosemount trip units presently associated

..

) with these transmitters are removed, b. Reactor pressure signals from PT.2 3 404A,5 PT.2 3 404C,D*

PT 2 3 52C.D have been retagged to these PT 2 3 404C D c. Compensated level indication signals from T. LOR.87 04 1 the compensation instruments are co.:cected to the existing indicators LI 2 3 85A and B LI.2 3 915 is ratagged to (C05 panel) LI.2 3 85 AX and BX (Energency LI 2 3 91 and LI.2 3 91A Shutdown panel) and LI.2 3 91 (CO3 panel), is retagged LI.2 3 113 The spare red pen of recorder PR.2 3 404A (C04C RCIC) is used for indication of fuel zone level. LR.2 3 110A and 5 both receive inputs from the compensated unit, d. Pressurs and level contact outputs from the Rosemount trip units that input the ECCS systens have been replaced by contact cutputs from the new level co=pensation instruments.

.

LOR.8704 R v. 1 RJ A/cj o Page 4 of 45 SUEJECT KATTER OUTLINE SUPPORT INFORMATION e. Transmitters LT.2 3 110A and B and PT 2 3 52A and B have been removed.

f. Contacts have been provided from the Not installed at this compensated instrument for the Alternate Rod time.

Insertion Modification.

l g. Four meters have been installed in cabinets These do not provide level

'

20C818 and 20C819 to provide indication of indicatioa actual differential pressure of the fuel zone transmitters. This indication can be used to determine if the equalizing valves for the LT.2 373A.D are open or closed.

1) If the equalizing valves are closed (normal operation) the indicator vill have an upscale reading maximum differen.

l tial pressure.

2) If the equalizing valves are open the indicators will indicate zero.

~ ) h. Vide range level indication is rescaled for

+60 to .165 inches indication (LI.2 3 85A

,

and B, LI 2 3 85AX and BX). Fuel zone range l 1evel indicators LI 2 3 91 and the red pen of FR.2 3 404A are rescaled for +60 to .325 inches indication. The LR 2 3 110A and B will be rescaled for both vide and fuel zone range 1. The accuracy of the fuel. zone range has been SiWee the fuel zone range taproved by taking into account when the variable leg uses the jet recire pumps are running, pump diffurer tap, this measurement is inaccurate whenever jet punp flow is present i

1) Outputs for fuel. zone level indication This signal vill not be vill use the compensated vide range af fected by jet pump instrument whenever level is above 162.5 flow, inches.

2) Outputs for the fuel zone level indica. Reetre puaps vill not be tion vill use the fuel. zone range operating belov .48" transmitter signal whenever level is below 162.5 and vill provide accurate

) level indication to levels of .325

./ inches.

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LOR.8704 Rsv. 1 RJ A/cj e Page 5 of 45 SUBJECT MATTER OUTLINE SUPPORT INFORMATION j. The power supplies to the compensation system will be a 125 Vac safeguard feed and a 125Vdc feed.

1) The outputs from the A 6 C supplies are Either power supply can bused together to form a redundant feed both channels system.

2) The outputs from the B 6 D supplies are Either poves supply can bused together to form a redundant feed both channels system.

'

3) Channel 120VAC 125VDC A,C 20Y35 BKRa2 20D23 BKRa1 B,D 00YO3 BKR=2 20D24 BXRe16 4) There is a small battery backup that vill supply the micro. processor memory in case of a complete loss of power, a) This battery has approx a 2 year life.

) b) Indication of battery failure is on the microprocessor panel.

C. Discussion 1. Because of concerns over inaccuracies of the Yarvay instruments during high dryvell tempera.

ture conditions two major changes were made, s. The Yarvay temperature compensating refer.

ence columns are removed and replaced with cold reference columns wit h piping that has a minimum elevation drop in the dryvell, b. Electronic reactor pressure compensation is installed to replace the Yarvay self.

compensatisn.

1) This also improves the accuracy of the instrument over its entire operating pressure range.

2) The new pressure compensation instruments co=pensate the level ceasurement by approximating the steam table values for

) the density of water and steam as a

_/ function of reactor pressure.

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LOR.8704 Rov. 1 RJ A/cj o Page 6 of 45 SUBJECT MATTER OUTLINE SUPPORT INFORMATION

'a) Errors caused by variations in dryvell and reactor building temperatures are not compensated.

a) Changes in indicated level due to The change in water changes in dryvell tesperature are density will af fect both essentially eliminated by the repiping lines equally, that makes the variable and reference legs have similar elevation drops inside the dryvell.

.

b) Calculations indicate errors due to EE 1457 1 l reactor building temperature changes this error was based on a '

vill be about .35 inches per 10 F. change in temperature from temperature change. 700F, 900F, 2. Reference Leg Boil Off is a condition that can occur at high dryvell temperature and icv reactor pressure, a. The reference leg elevation drop in the g dryvell vill be 30 inches.

i

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b. Under these conditions level indication This amount of error unets could be as much as 40 inches high, the acceptance criteria

,

established by the NRC in

, 1) Since the bottom tap (variable leg) is at Generic letter No. 84 23.

approv.imately .172 inches it is possible that there vill be no LFCI/ Core Spray initiation signals.

The cid Yarvay system

'

2) FRC Generic letter states that, "under .

all reactor and Dryvell conditions. would not do this, reactor level censuring systems shali ,

'

provide, at a minimum, an Operator I

alert".

1 3) The Compensated Reactor Level system will produce a Reactor double lov (.48 inch)

alarm at an actual level of 90 inches.

4) The operator needs to recognize that If initiation is not done under these conditions, LPC: and Core by D.V. pressure

'

Spray should be manually initiated.

5) The operator needs to remember that under these conditions level indication of the vide range and fuel zone vill indicate up

}

- to 40 inches high.

,

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LOR.8704 R v. 1 RJA/cj o Page 7 of 45 SUfJECT MATTER OUTLINE SUPPORT INF0FJ'ATION 3. ECCS Activations a. The level 8 (+45 inch) trips off HPCI and T. LOR.87 04 2 RCIC turbines, the main turbine stop valves and the feedwater turbines will occur at a higher and more conservative reactor water level with the compensation than with the Yarvay messarement for reactor pressures less than 1000 psig.

1) Since these trips occur on increasing level, the trip at a higher level is conservative as considered from an ECCS aspect and will result in maintaining a l

]

greater volume of water in the reactor. L 2) This level vill be close to 45 inches for all reactor pressures.

3) The operator should remember that if HPCI

'l trips on high level it will auto reset.

The reset set point is set at 29 inches.

i b. The level 2 (.48 inch) initiatier of HPCI 6 T. LOR.87 04 3 ,

RCIC and tripping of the recirculation pumps "

will occur at a higher and more conserrative water level with pressure compensation.

'

This is conservative from an ECCS perspec.

tive in that these systems will maintain a :

!

greater volume of water in the reactor.

c. The level 1 ( 160 inch) which had previcasly T. LOR 87 04 4 I been set at 130 is now being moved to .160. T. LOR.87 04 5 Postulated high temp. is 1) Removal of the Yarway columns reduced the 3400F, ,

I error associated with high dryvell temperature to 2.7 inches. Initiation assumes no change in reactor pres. j sure. -

i 2) For nessurements at the desi ,

operatingtemperatureof135gnbasesT., the 169 i inch activation point vill assure i operation above the Technical Specifica. l tion limit.

(

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)

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QUESTION 3.05 (2.00)

The reactor is being started up. All IMs except for IRM D, are on range 7. An I&C technician is about to troubleshoot IRM D and in prepatstion for this, the joystick on panel C05 has been position 5d to bypass IM D.

a. IM A reads 11 when the operator inadvertently ranges DOW to range 6. OtAT

vill it read on Range 67 WAT TRIP, if any, vill occur?

I i b. IM A has been ranged back to range 7. The I&C technician begins to troubleshoot IM D but instead of opening the drawer for IM D opens the drawer to IM B. WAT TRIP, if any, will occur? JUSTIFY your answer.

'

AN S'='ER 3.05 (2.00)

a. 110 (0,5) A rod block will occur (0.5)

b. A half scram vill occur (0.5) since IM B vill generate an inop conaltion.

'

(0,5)

i i REFERENCE PBAPS 1.OT 250 p.9 I LD 250 7 i

{ PA 215003K401(3.7) 215003K402(4.0)

,

215003%401 215003X402 ...(KA'S)

.

FACII.!TY COMMENT AND RESOLUTION

l 3.05 b. If opening the IM drawer results in High voltage supply less than 125VDC or a module being unplugged or if the drawer select switch is l placed out of operate in preparation for troubleshooting, an inop

'

I condition results along with a half scram.

!

RETERENCE PBAPS 14T.0250 pages 9, 10, 11.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. *

l QUESTION 3.08 (2.00)

You assume the shift, with the mode switch in STARTUP, and with the fo11ovin5 rod position distribution:

All Tods in RkH Croups 1 3 have been withdrawn to position 48 except for one

,

rod in each Group: 22 27 in Group 1, 46 35 in Croup 2, and 18 03 in Group 3 these three rods are still fully inserted. All rods in Croups 4 10 are ;

fully inserted (position 00) except for rod 34 27 (Croup 4) which has just

,

been withdrawn two even notches past its insert limit. RSCS is bypassed, ;

"

!

, STATE the Rod / Rod Group number you would see displayed in each RkH window. If 1 nothing vill appear in a vindow, write "Blank".

'

i 1. Rod Group 2. Insert Error , _ . .

. 3. Insert Error '

'

4. Vithdrav Error

,

ANSVEA 3.08 (2.00)

a 1. 03 (0,5)

2. 22 27 (0.5)

i 3. 46 35 (0,5)

4, 34 27 (0.5) ;

) REFERENCE

<

PBAPS 141 90 pgs. 4 6, 10, 13

!

ID 90 2, 3 i FA 201006K401(3.4) 201006K402(3.5) l l 201006K401 201006K402 . . . ( FN S )

i j FACILITY COMMENT AND RES01.UTION i

! i

) 3.08 The stated scenario is not possible, violates procedure and Technical !

- Specifications and therefore should be removed from the exam. Since the mode switch is in start up, power must be less than 216, under these

,

conditions RSCS is required to be operable or the reactor shall be brought l l to a shutdown condition immediately. .

I RETERENCE r 1. PBAPS LDT.0100 pages 8, 11 !

1 2. Procedure S.4.3.1. Revision 6 page 2 I

'

l 3. PBAPS Technical Specifiuations pages 102, 102a I

I

$ e 4 ,

f

'j i l

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. o QUESTION 3.09 (3.00)

An ADS blowdown is in progress on Peach Botton Unit 2.

a. For each of the following conditions, will the ADS valves CLOSE or REMAIN OPEN7 1. Reactor pressure decays to 40 psig.

2. An operator shuts down all Core Spray pumps, 3. Rx water level recovers to +10 inches, b. During the ADS blowdown, the operator depresses the ADS A and B reset buttons. BRIEFLY DESCRIBE how the ADS valves and logic will respond ASSUMING the initiating signals att11 exist. Include applicable setpoints and time delays.

ANSVER 3,09 (3.00)

a. 1. close 2. close 3. remain open

! b. The valves vill close (0,5), the timer will restart (0.25), and at the end of the timer cycle, 105 see (0.25), blowdown vill recommence. (0.5)

REFERENCE PBAPS LOT.330 pgs. 7 8 to 330 2.c. 5.b KA 218000K403(3.8) 21800CK501(3.8)

218000K403 218000K501 ...(KA'S)

TACILITY COMMUTT AND RESOLUTION 3.09 a.2. The ADS valves will remain open if any RRR pump is running. Accept either answer based on status of RRR pump.

RETERENCE PBAPS LOT.0330 page 8

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QUE$ TION 3.10 (2,50)

For each of the conditions below, STATE what AUTOMATIC action vill occur (SCRAM, RALT. SCRAM, ROD BLOCK). If none occurs, state NONE. It more than one action occurs, S' ATE the most severe action, i.e., half scram is more severe than a rod block, a. At 354 power, loss of voltage to "A" AFKM occurs, b. At 60% power, the AFRM flow converter fails downscale, c. UNIT 2 is operating with 50% recire flow, rods are pulled to increase power to 754.

d. UNIT 3 is operating with 50% recire flow, rods are pulled to increase power to 754, e. At 334 power, all four turbine stop valves trip shut.

ANSVER 3.10

a. half scram (0 b. half scran (( -

'

c. none ((

d, rod block (C. .

e, scram (0,5)

E EFERENCE FBAPS LOT 270 pgs. 67 LO 270 2 KA 212000K101(3.7) 212000K110(3,2) 212000K602(3.7)

212000K101 212000K110 212000K602 ...(KA'S)

\

FACILITY COMMENT AND RESOLUTION 3.10 b. Due to a change in the Unit 2 AFRM Flux 5: ram Trip Setting (Run Mode)

to:

$$ 0.58V + 624 0.55 AV, when an AFRM flow tonverter fails devnscale, V goes to zera. Therefore the scram setpoint is equal to 624 which is greater than the current 604 power. The most severe action for Unit 2 is a rod block, not half.

scras.

REFERENCE FEAPS Technical Specifications page 9 (Uatt 2)

Amendment 123, 12/31/87

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Unit 2 PSAPS i 5ArtTY LIMIT 1.1 TUEL CLADDING INTEGRITY LIMITING SArt?Y SYsttM sr?TINc

_ ADD 11 ca bi li t y: 2.1 TUEL CLADDING INTEGRITY

"

Applicability:

The safety Limits established to preserve the fuel cladding The Limitinq Safety System settings integrity apply to those apply to trap setting of the ~

variables which monitor the instruments and devices which are fuel thermal behavior. provided to prevent the fuel cladding integrity Safety Limits

,

.

from being exceeded.

0.b,17tet i ve s i obieetives -

l The objective of the satety

'

Licits is to establish limits The objective of the Limiting safety !

which assure the integrity of system Settings is to define the ;

i the fuel cladding. lavel of the process variables at i

which automatAc protective action is

! initiated to prevent the fuel cladding integrity Safety Limits frem being ,

exceeded. *

!

_Specificatient i seecification: '!

A. Resetor Pressure 1 800 esta The limiting safety system sattings !

l , and core Tlev 1 lot et Ratid shall be as specitaed below:

'

s0 A. Neutron riux scram

The.etistence of a minimum 1. APPy riur seram Tri; settine i

criticti power ratio McFK inun_hedei

!

less than 1.07 fer two I recirculation loep cperation, 1 cr 1.0! forshal. single loop when the Mode Switch is in the !

cperation, constitute RUN position, the APRM flux !

' scram t:1p setting shall be:

violation cf the fuel cladding

'

-

integ !:y safety limit. S 1 C.58W - !21 - C.554W

'

l To ensure that this safety where:

limit is not exceeded, neutron [

flux shall not be above the sa setting in percstat of rated scram setting established in thermal power (3293 MWt)

speci!! cation 2.1.A for longer - ;

than 1.15 seconds as indicated W * Loop recirculating '

j by the process computer. When flow rate in percent the process computer is out of of design. W is 100 for l service this safety limit shall core flow of 102.5 '

be assumed to be exceeded if million ib/hr er greater.

the neutron flux exceeds its {

scram setting and a contr '

red scram does not occur. I

~

.

,

Acen0 ment No. 25, J4, 42, (f. 7f,-S-

12/3U67 i

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. .

i Ql'ESTION 4.05 (2.50)

In accordance with CP.2 Appendix I, "Startup Rod Vithdrawal Sequence Instructions", individual control rod movements are to be verified, if the Rod Vorth Minimizer is INOP bclov 254 power.

a. Wlo must t. tfy rod movement and HOW is this verification performed?

l b. VHAT ADDITIONAI. VERITICATION is required at t.he end of each group pull? IIOW is this performed? i c. VHAT DOCIMUiTAT10N is required for the verification irade at the end of a group pull? ,

I ANSVER 4.05 (2.50)

!

a. A second licensed operator, (0.25) with no other duties, shall verify the correct red is being noved to the correct position, in the required l sequence, (0.25) by using the process computer. (0.25) !

i b. The Reac;or Operator, Second licensed operator, and Control Room SRO (0.25) !

shall INDEPDit , Y verify the rod pattern (0.25) by comparing the 0D.7

'

i Option 2 to t!t . ' 1 Appendix I core map. (0.25) I

c. The Reactor Operator, Second Licensed Operator, and Control Room SRO (0.25) l shall sign the 00 7 Option 2 and the CP.2 Appendix I core map (0.25) and attach it to the appropriate core map in CP.2 2. (0.25)

REFERENCE CP.2 Appendix I p.1 1.0 1530 4 201002K105(3.4) 201002c001(3.8) 201002c013(3.4) i 2010020001 2010020013 201002K105 . . ( rN S ) I

,

TACIT.ITY COMMUTT AND RES01.UTION 4.05 b,c The Shif t Supervisor is the same as the Control Room SRO and either i i should be accepted.

I !

RETERUTCE Procedure S.5.5.D Manual Bypass of RVN ,

i Revision 5 page 2 '

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, e l l

QUESTION 4.06 (3.00) ,

'

Regarding ON-113, "Loss of RBCCW":

a. STATE ALL the actions the operator is required to verify when RWCU non-regenerative heat exchanger outlet temperature reaches 200 degrees.

b. If RBCCW cannot be restored within 5 minutes, ON 113 specifies how to !

shutdown the recire pumps. WHAT IS THE BASIS FOR:

1. Removing the recire pumps from service?

2. First reducing recire flow to minimum and then tripping the pumps 10 seconds apart?

3. Shutting their discharge valves?

ANSWER 4.06 (3.00)

a. 1. MO 12 15 (0.166), MO 12 18 (0.166), and MO 12-68 close, or Croup 2a i isolation (0 5) l 2. RWCU pump (s) trip (0.5)

3. FSCU demin hold pumps start (0,5)

,

b. 1. to protect the pump seals (0.5)

2. to minimize the transient on the reactor (0.5)

3. to prevent the flow through the recire lines from turning the pump after the trip (0.5)

REFERENCE PBAPS LOT-1550 pgs. 17-18

o o 1550 2 KA 295018K101(3.5) 295018K303(3.1) 295018C007(3.2)

295018C007 295018K101 295018K303 ...(KA'S)

FACILITY COMMENT AND RESOLUTION 4.06 b.1. The candidates should not be penalized if they do not use the term seals since the lesson plan Subject Outline and ON 113 do not.

REFERENCE 1. PBAPS LOT 1550 page 18 2. OFF NORMAL BASES ON 113 Loss of RBCCW, page 1.

. .

I QUESTION 4.08 (2.00)

n In accordance with OI-110, "Reactor High Level":

a. While operating at 804 power, reactor water level unexpectedly begins

- increasing. STATE the three methods that can be used to regain control cf 4 water level, b. STATE the concern with reactor water level exceeding 90 inches.

i ANSWER 4.08 (2.00)

a. 1. Lower the water level setpoint. (0.5)

2. Swap the FWCS from auto to manual. (0.5)

3. Remove a RFP from service. (0,5)

b. There is potential for flooding the main srcam lines (and introducing water into the turbine).

l REFERENCE PBAPS LOT 1540 p.9 Bases, OT 110 p.2 l LO 1540 2 KA 295008C010(3.8) 2950080007(3.2)

' 2950080007 295008C010 ...(KA*S)

FACILITY COMMENT AND RESOLUTION I t

' l 4.08 a. Other methods of regaining control of water level should be accepted >

'

l since the decision is left up to the operator according to the Bases of OT-110. For example, use of MSC, swap to alternate loval instrument, ,

trip HPCI and/or RCIC if running, etc.

f b. Besides introducing water into the turbine, there are other concerns with reactor water level exceeding 90 inches and these answers should

,~

,

'

also be accepted. For example, the weight of water in the main steam l

<

lines and flooding out HPCI and RCIC supply lines, i l

l

'

REFERENCE ,

PBAPS OT 110 Reactor High Level Bases pages 1 & 2 l t

l I

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I t

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l

QUESTION 5.03 (3.00)

The Residual Heat Removal pumps are being used in Shutdown Cooling Mode. HOW will AVAILABLE and REQUIRED Net Positive Suction Head for the Residual Heat Removal pumps be affected by each of the folicwing changes (INCREASE, DECREASE, or NOT AFFECTED)?

NPSH NPSH AVAILABLE REQUIRED a. Reactor Water temperature increases (1.00)

b. Reactor Water level decreases (1.00)

c. RHR System flowrate decreases (1.00)

ANSWER 5.03 (3.00)

Avail. Required a. DECREASE REMAIN THE SAME b. DECREASE REMAIN THE SAME c. INCREASE DECREASE (0.5 pts each)

REFERENCE LOT 1290 3.7 3.3 3.4 3.6 ...(KA VALUES)

202001K101 202001K103 202001K105 202001K122 ...(KA'S)

FACILITY COMMENT:

Required NPSH is not taught within Lesson Plan subject matter or listed as an Objective. Many candidates may assume ' required' varies inversely as 'available'

and answer accordingly.

REFERENCE: LOT 1290 RESOLUTION:

Reduce weight of ' Required NPSH' portion to 0.25 pts each therefore reducing total question vale.e to 2.25.

E~

. .

QUESTION 6.04 (3.00)

With the Unit operating at 75% rawer, an electrical fault causes the Maximum Combined Flow Setpoint of the EHC system to drop to minimum.

HOW WILL EACH OF the following RESPOND after the fault? WHY7 (Consider response through ONE MINUTE after the fault. Assume NO OPERATOR ACTION.)

ATTACHED FICURE, EHC LOGIC, IS PROVIDED FOR REFERENCE a. Turbine control valve position (1.00)

b. Bypass valve position (1.00)

c. Reactor power (1.00)

ANSWER 6.04 (3.00)

a. The TCVs will close to 50n flow position (0,5)

The TCV low value gate passes a HCF signal of 50% rather than the signal from the pressure controller. (0.5)

b. The BPVs will remain closed through the transient (0.5)

the MCF summer will send a zero signal to the BPV LVG (0.5)

c. Reactor power (and pressure) will rapidly increase following (0.5)

the fault.

The reactor will scram on High Flux and/or high pressure (0.5)

because of the closure of the TCVs

.

REFERENCE i LOT 0590 pp. 12 15 3.7 ...(KA VALUES)

245000K602 ...(KA'S)

i I FACILITY COMMENT:

Although minimum value for Max. Combined Flow Potentiometer is 50% in Manual, wording of question "Electrical Fault" was interpreted as failed to Zero.

This would be interpreted as a full closure of che Turbine Control Valves.

REFERENCE: LOT 0590 j RESOLUTION:

Accept either interpretation (faulted to ::ero or 506) as correct.

i

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. . .

QUESTION 7.02 (3.00)

For each of the following conditions, STATE which Emergency Procedure is entered (If more than one. procedure is entered, state all Emergency Procedures that are entered. If none are entered, state NONE).

a. RPV LEVEL BELOW -48" or unknown (0.50)

b. Drywell TEMPERATURE ABOVE 145 deg. F. (0.50) i c. Drywell PRESSURE ABOVE 2 psig. (0.50)

d. Conditions requiring a GROUP I ISOLATION. (0.50) :

e. Torus LEVEL OUTSIDE the 14.6' to 14.9' band. (0.50)

f. SCRAM CONDITIONS with POWER LEVEL above 34 or unknown. (0.50) ,

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ANSWER 7.02 (3.00)

a. T 101 (0.50)

b. T 102 (0.50)

c. T-101 (0.25) AND T 102 (0.25)

i d. T 101 (0.50)

e. T-102 (0.50)

f. T 101 (0.50)

REFERENCE Systematic E0P Flow Path T-101 Learning Objective LOT 1560 9 4.5 4.5 4.6 4.7 2950250011 2950240011 2950310011 2950370011 ...(KA'S)

FACILITY COMMENT:

For parts a, c, and f the operator actually enters Procedure T 100 briefly, THEN proceeds to T 101 and/or T 102. Scee candidates omit this due to the short amount of time involved, realizing they are quickly into a more involved procedure.

REFERENCE: Trip Charts RESOLUTION:

Accept answers which include brief entries into T 100 in addition to T 101 and T 102.

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. . . -

QUESTION 7.03 (3.00)

The reactor is at full power when an ATWS occurs with no control rod motion.

As part of T 101 you are directed:

IF NOT SHUT DOWN WITH RODS THEN INJECT SLC ,

'

BEFORE TORUS TEMP.

REACHES 110 DEG. F. ;

,

a. DEFINE NOT SHUT DOWN. (0.50)

b. Once boron injection is started, the operator is directed to perform T 117 (REACTOR LEVEL / POWER CONTROL) concurrently with '

4 T-101. What is the basis for using reactor water level to control reactor power while boron is being injected? (1.00)

c. What is the basis for the Maximum Standby Liquid injection time? (0.75)

d. What is the basis for the Minimum Standby Liquid injection time? (0.75)

ANSWER 7.03 (3.00)

a. The reactor is not shut down when the reactor is not subcritical (0.50)

(The definition does not reference any power to allow the operator to inject boron during any condition in which the reactor i is not shutdown on rods including power levels less than three percent.)

RSFERENCE

Lor-0310 p 6. 7 i TRIP T 101, T-102 (BASES)

4.5 4.3 4.5 4.2 l

295037A104 295037G003 295037K204 211000A208 ...(KA'S)

FACILITY COMMENT: '

In part a. the candidate can use various methods to define or determine "NOT

. SHUTDOWN" other than key. t REFERENCE: N/A RESOLUTION: Alternate methods used to determine shutdown could be accepted ..

APRMs not downscale, reactor period, steam flow not consistent with decay heat, i

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

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QUESTION 8.02 (2.50)

,

, During a refueling outage, a Control Room Operator has worked twelve (12) hour days for the last five (5) days. On the sixth (6) day ha is scheduled to work only eight (8) hours. His relief called in sick and no other personnel are i available to fill the operator position.

a. HOW LONG can the operator continue to work and STATE the two (2)

limits restricting his work hours in accordance with Administrative l Procedure A 407 (1.50) t-b.1, What documentation is required for an individual to exceed the work l hour criteria of A 407 (0.50) ,

2. Who, by title, may authorize exceeding the overtime guidelines? (0.50) ,

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ANSWER 8.02 (2.50) ,

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,

a. The operator can continue to work for the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

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I (0.50)

) If he worked more than four (4) hours he would exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in seven (7)

days (0.50) and, he would exceed working more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

(0.50)

i b.1. A Personnel Staffing Deviation Form shall be filled out. (0.50)

i 2. The overtime shall be authorized by the Station Superintendent, (his

'

alternate, or higher level of management.) (0.50) l I

.

REFERENCE Administrative Procedure A 40 *

Learning Objective LOT 1570 3 h 4.5 294001A011 ...(KA'S) :

! FACILITY COMMENT:

Part b.2. Title of Authorization has changed to Plant Manager vs. Station I

Superintendent; but the Procedure A 40 has not been changed, t

! REFERENCE: Admin. Procedure A 40 (See attached Station organization Chart) {

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RESOLUTION: Accept Plant Manager vs. Station Superintendent.

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QUESTION 8.09 (2.25)

During plant operations at 804 power, RBM "B" has fai?id and is bypassed. APRM

"E" har also failed and is bypassed.

APRM "A" fails downscale.

Utilizing the Technical Specification provided in the attachments, DETEhMINE any LCO's entered by the failure of APRM "A". Iadicate the DURATION of the LCO's and the ACTION required if the specified LCO time elapses. (2.25)

ANSWER 8.09 (2.25)

Failure of APRM "A" results in less than two (2) instrument channels for trip channel "A". (0.25) The Technical Specification requires two (2) instrument channels for each trip channel. APRM "E" is already bypassed, so APRM "A" cannot be bypassed. (0.25) Trip system "A" shall be placed in the tripped condition.

(0.50)

APRM "A" is also the normal reference APRM for RBM "B" (0.25). Since APRM "E" is the alternate reference APRM for RBM "B", RMB "B" is INOP. (0.50) A seven (7)

day LCO was in effect due to RBM "B" being INOP. RBM "A" is not INOP and shall be placed in a tripped condition. (0.50)

REFERENCE Technical Specifications 3.1 and 3.2 3.0 3.0 3.3 4.3 3.5 3.3 205002K101 215002K604 215002A203 215002C011 215005K103 21500$K307 ...(KA*S)

FACILITY COMMENT:

The second part of the answer is wrong. The reference APRM for the A Rod Block Monitor is the "E" APRM, "C" is the backup. The reference for the "B" RBM is the "B" APRM and "D" is the backup. When the "E" APRM becomes INOP, the reference APRM for "A" RBM becomes the "C" APRM. This leaves the "A" RBM still operable. No LCO's are in effect for the RBM.

REFERENCE: LOT 0280, page 4 Lower right.

RESOLUTION: Eliminate second part of question.

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l ATTACHMENT 4

HRC RESPONSE TO FACILITY COMMENTS l 1.07 Coment accepted. The "NPSH REQUIRED" portion will be deleted and

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the question value lowered to 1.5.

L 2.07 Coment partially accepted. Credit will only be given for rupture

.

diaphragms and their correct setpoint.

2.09(b) Coment partially accepted. Question (b) is to be deleted since it does not distinguish between kncwledge of the LOCA signal and just guessing. Point value of the ouestion is 2.0; Sectior .alue 24.5.

3.01 Coment accepted. This material was not provided in what was submitted to the region for exam preparation.

l 3.05 Coment partially accepted. Placing the drawer select switch to out of operate is not part of the information given in the question and therefure requires the candidate to rnake assumptions in order to answer the question. Taking the drawer select switch to out of g operate will not be allowed as an acceptable answer.

3.08 Coment accepted.

3.09 Coment partially accepted. "Remain Open" is to be considered correct only if the candidate states that RHR is presured to be running; otherwise "Close" is the correct answer.

3.10 Coment accepted. Unit 2 receives a Red Block.

Unit 3 a half scram.

l l 4.05 Coment accepted. No change to answer key is required.

4.06 Coment accepted.

, 4.08 Coment noted. However, since the question tests knowledge of l imediate operator actions, the answer key remains unchanged.

5.03 Same response as for question 1.07.

6.04 Coment not accepted. The question explicitly states that the I Maximum Combined Flow setpoint drops to minimum, and not to zero.

7.02 Coment accepted. Answer key has been modified accordingly.

7.03 Coment not accepted. The question asks for a definition of NOT SHUT DOWN, not how to determine if the reactor is not shut down.

8.02 Coment accepted. Answer key has been modified accordingly.

8.09 Coment partially accepted. Answer key for part (b) has been modified as follows: No LCOs for the RBM. Additionally, the point value of the question was reduced to 2.0.

I

v t e Inspection Report - Peach Bottom - 1988016
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1988 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 11(0) 12(0) 13(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 29(0) 31(0) 32(0) 33(0) 34(0) 35(0) 36(0) 37(0) 38(0) 39(0) 41(0) 99(0) 14(0) Security Operator Exams Emergency Planning Other

IR 05000277/1988016

Text

Attachments:

Dear Mr. Callo:

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