ML20214P376

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Operator Requalification Exam Rept 50-445/OL 86-02 on 860822 & 29.Exam results:4 of 14 Senior Reactor Operators & 4 of 7 Operators Failed at Least One Portion of Exam. Requalification Evaluation Program Rept & Exam Encl
ML20214P376
Person / Time
Site: Comanche Peak Luminant icon.png
Issue date: 11/22/1986
From: Cooley R, Whittemore J
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20214P321 List:
References
50-445-OL-86-02, 50-445-OL-86-2, NUDOCS 8612040177
Download: ML20214P376 (66)
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OPERATOR LICENSE EXAMINATION REPORT No. 50-445/0L 86-02 j Docket: 50-445 1, ,

Licensee: Texas Utilities Generating Company . s

  • 400 North Olive Street Lock Box 81 Dallas, Texas ,

Operator Requalification Examinations administered to the Comanche Peak Station (CPSES) <' '

Chief Examiner: t/ 8 *!8 l J6hn E. Whittemore Date Approved By: .k. h ll/Jo/%

Date R. A. Cooley ~ g Srmmary Written and operating examinations were administered to 14 Senior Reactor Operators and seven Reactor Operators at the Comanche Peak facility during the weeks of September 22 and 29, 1986. Four Senior Reactor Operators and four Reactor Operators failed at least one portion of the examination.

8612040177 86112%I PDR ADOCK 05000445 V PDR

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CPSES OPERATOR LICENSE EXAMINATION REPORT s

Report Details

1. Examination Results Written and operating examinations were administered to 14 Senior Reactor Operators'and 7 Reactor Operators. Ten of the Senior Reactor Operators passed all examinations with one failing the operating examination and three failing the written examination. Three Reactor Operators passed all examinations and four failed the written examination.
2. Examiners J..E. Whittemore (Chief. Examiner)

D. N. Graves F. S. Jagger W. C. Hemming T. P. Guilfoil N. C. Jensen

.I. G. Kingsley S. L. McCrory B. A. Picker A. J. Vinnola

3. Examination Report Performance results for individual candidates are not included in this report as it will be.placed in the NRC Public Documnt Room. This Examination Report is composed of the sections ~ listed below,
a. ' EXAMINATION REVIEW COMMENT RESOLUTION In generai, editorial comments or changes made during the examination or during subsequent grading reviews are not addressed by this resolution section. This section reflects coments and ' recommended changes to examination answer keys by the licensee. Examination key modifications resulting from these comments and recomendations are included in the master examination keys, which are provided elsewhere in this report. Comments and resolutions are listed by examination section question number. It should be noted that the grading examiners were under no obligation to incorporate any of these comments due to their arrival 2 days past the normal 5-working-day period allowed for accepting comments.

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. Comments and Resoluticns Comment: 1.01.b.2 Correct answer should be "1065 + or - 100 pcm Resolution:. Accepted. Key modified.

I Comment: 1.03.h Depending on the reference cited, differential boron worth (DBW) may become.more or less negative. Westinghouse Reactor Core Control shows DBW becoming less negative while the Comanche Peak Technical Data manual depicts it as becoming more negative. In either case the change is small and for all practical applications 1the change is negligible. Therefore, it is recommended that either of the above answers be accepted, or the question be removed.

Resolution: Partially accepted. DBW becomes less negative from BOL to E0L assuming a constant boron concentration. If concentration decreases, then DBW becomes more negative. The answer key has been modified to reflect the above.

Comment: 2.02 Question asked for major components in the flow path. The isolation valves should not be considered major. Also " charging flow control valve," FCV121 is not listed but HCV182, which creates back pressure to ensure RCP seal injection flow, is included. Inconsistency in choosing major components should warrant maximum grader discretion.

Resolution: Noted Comment: 2.06.h CCW isolation valve from RHR heat exchanger. throttles open upon receipt of an "S" signal.

Resolution: Accepted. Key modified Comment: 2.07.4 Pressurizer PORVs are nitrogen operat'de at all' times, therefore a loss of control air has no effect on them.

Resolution: Accepted. Key modified r

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Coment: ~3.01.a Key should allow any response between 1960 and 1829 psig.

- Resolution: Accepted. Key modified Coment: 3.03.c The answer key breaks up FRV and FRBV as separate answers from FW isolation. FRV and FRBV close as part of FW system isolation. Point value should be changed to reflect a 3 part answer.

Resolution: Accepted, key modified.

Coment: 3.07.a Table RPS4, under containment ventilation isolation does not show that a high rad level on the plant vent stack monitors will generate a containment ventilation isolation. However on page IIIl0.30, Safeguards Actuation Logic, fig. ICD 8, in the upper right hand corner of- this drawing, it shows that a containment isolation signal is generated by this monitor.' Due to this conflicting information it is recommended that containment ventilation isolation be accepted but not required for full credit.

Resolution: Accepted. Key modified.

Comment: 3.08 ITT3 parts of this question ask what interlocks prevent opening of a valve. The key lists violations of the interlocks, i.e, conditions that would prevent opening of the in part a, but listed the actual interlocks for parts b. and c. This same confusion was experienced by some candidates. It is therefore recommended that maximum grader discretion be utilized.

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Resolution: This comment is ignored as there is no reason the key should have confused the candidate while writing the examination, and a proper description of the interlock was adequate for a full credit answer.

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Comment: 4.01 -

No clarification was given for total quarterly dose. -It.is .

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uncertain if the 1100 mrem given asithe quarterly dose was ' , -

inclusive of the 200 mrem weekly dose. Answers in parts a. and: ' .

b. should accept answers based on 1100 and 1300 mrem quarterly -t exposure. .,-

Resolution: Accepted. The answer key has been modified to provide full credit no matter what assumption the candidate makes regarding current quarterly exposure.

Comment: 4.04-The main question references FRS0.1. Part b. asks for immediate action on failure of manual reactor trip. This may have lead some examinees to list the immediate action steps of FRS-0.1.

Resolution: Noted.

Comment: 4.08.b The question reference states the purpose for the entire procedure, not the purpose of isolating the ruptured steam generator. Therefore the answer should be changed to read:

1. To isolate flow from the steam generators to minimize radiological release.
2. To maintain pressure in the ruptured steam generator greater than the pressure in at least 1 intact steam generator following cooldown of the RCS.

Resolution: Accept. Key modified.

Comment: 5.04.b & c Answers assume that steam dump valves are open. At 10% power the steam dumps may or may not be open. Grader discretion should be used.

Resolution: Accepted. Key modified.

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Comment: 6.04 Question asks for function, key gives design basis. Some candidates were provided clarification, but some were not.

Reasonable answers stating the function should be accepted.

Resolution: Not accepted. The question specifically asks for the safety-related function of the CST which is equivalent to the technical specification basis. Additionally, after several questions to the proctors about this question, a clarification announcement was made detailing the specific information desired to answer this question.

Comment: 6.08.b CPSES does not require operators to memorize detailed procedures other than immediate actions in E0Ps and initial action in ABNs. This question asks for information contained in subsequent operator actions which would be done after consulting the procedure.

Resolution: Not accepted. The question does not require memorization of procedural steps. The question evaluates the candidates knowledge and understanding of the IR/SR NIS interface during a reactor shutdown. The answer key references a CPSES system description, not a procedure.

Comment: 6.09.b Question asks for what automatic actions occur and why. The key answers what and when. In light of this confusion, request that significant latitude be used in grading this question.

Resolution: Accepted. No key modification required.

Comment: 7.01 Question "a" could be interpreted as having no air for operation of the valve since " control air" is not defined. Therefore, manual valve operation should also be accepted as a method of valve operation.

Resolution: Accept. Key modified.

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Comment: 7.03 U5ES does not require operators to memorize detailed procedure steps other than imediate actions in EOP's 'or _ initial actions in ABNs. This question asks for information contained in _ -

subsequent operator actions which would be done after consulting the procedure. Although the answer references ABN105A, the question neither. references ABN-105A nor mandates utilization of the Boric Acid Storage Tank. - Other methods of boration (such as 112D and E via Charging Pumps) are available and should be considered as acceptable means of RCS boration.

Resolution: Partially accepted. The question does not require memorization of detailed procedural steps. The question evaluates the candidate's knowledge and understanding of boration methods without using the boric acid pumps. The answer requires major

-(and very basic) actions to be taken to borate the RCS without using the boric acid pumps. However, the answer key has been medified to accept the RWST via LCV1120 and E flowpath as an alternative answer.

Comment: 7.04.e The question, although prefaced by general usage of IP0's, is answered specifically from IP0-002A. The condition required by IP0-002A (50 deg's subcooling) is not applicable throughout the IP0s. For example, while operating in accordance with IP0s at 100% power, actual subcooling-is approximately 30 deg's. Since the question was general in nature, the answer should accept any correct subcooling from all IP0s.

Resolution: Accept. Key modified Comment: 7.07.a Although the question addresses the basis for precauticn of not.

starting an RCP unless a steam bubble exists in the pressurizer, the answer includes the basis of a steam bubble. This portion of the answer (0.5 points) does not apply to the question and should not be considered for grading purposes.

Resolution: Not accepted. The effect of the steam bubble in reducing the pressure spike is an essential evaluation point as it is the only condition stated by the procedure and is directly related to the intent of the precaution.

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8 Comment: 8.09 15FEi question addresses a poorly worded section of Tech. Spec's.

Although action statement c. under 3.2.1 would allow power to be increased immediately, provided the Axial Flux Difference is within the target band, doing so would be in violation of the LC0 and would place the operator in action statement b. Because of this apparent conflict in the action statement, maximum grader discretion is recommended.

Resolution: Accepted. The key has been modified to allow additional responses.

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b. SITE VISIT

SUMMARY

At the end of the written examination administration, the Licensee was provided a copy of the examination and answer key for the purpose of commenting on the examination content validity. Near the conclusion of the site visit, the examiners met with licensee-representatives the to discuss the site visit and conduct of the examinations. The following personnel were present for'this exit I meeting.

NRC LICENSEE J. E. Whittemore (Chief Examiner) R. A. Jones D. N. Graves W. Melton M. A. Niemeyer M. J. Riggs A. B. Scott C. L. Turner R. R. Wistrand In accordance with NUREG 1021, no preliminary pass / fail results based on operating exam performance were provided to the licensee at this time. It was explained to the licensee that region policy was to have results finalized within 30 days, however this may not be attainable due in part to the large number of examinations administered and the extended time elapsed before. receiving the licensee comments.

c. GENERIC COMMENTS

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The followir.g apparent problems or areas of weakness were observed by' by the examiners and noted to the licensee.

(1) At least two candidates could not find a' procedure for borating the RCS (2) Several operators were r.ot familiar with a problem' that the simulator duplicated whicn actually occurs in the plant. The problem of makeup termination caused by flow deviation was not understood or corrected by the majority of operators.

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(3) At least three individuals were confused on the make up recorder parameters by routinely assuming- that total flow indication was

. primary water flow.

(4) SR0's were very slow in .transitioning from ABNs to ERGS (5) SR0s were slow if not reluctant to initiate ECAs.

(6) SR0s demonstrated a lack of assertiveness in that R0s were frequently requesting guidance on how to proceed and were frequently ignored.

(7)' ERG foldout pages were rarely used.

(8) During the-2-week period of operating examination administration, several problems occurred with the plant specific simulator. The simulator instructors were made aware of-these problems as the examinations progressed and two of the problems were noted to the licensee staff during the exit meeting.

d. EXAMINATION MASTER COPIES The SR0 and R0 master examination and answer keys follow. Changes resulting from licensee comments are reflected in these keys.

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0 0-REQUALIFICATION PROGRAM EVALUATION REPORT Examiner: J. E. Whittemore Date(s)ofEvaluation: 9/23/86 - 10/02/86 Areas Evaluated: X Writen Oral X Simulator Examination Results:

R0 SRO- TOTAL EVALUATION Pass / Fail Pass / Fail Pass / Fail (S,M,orU)

Written Examination 3/4 11/3 14/7 U Operating Examination Oral Simulator 7/0 13/1 20/1 S Evaluation of facility written examination grading Overall Program Evaluation Satisfactory Marginal X Unsatisfactory (Listmajor deficiency areas with brief des-criptive comments)

1. Poor overall performance especially by Reactor Operators, on writtbn examination.

Submitted: Forwarded: Approv'ed':

bl@dda R.6 Branch Chief Examiner Section Chief -

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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _QDd8NQHg_Eg85_1_________

REACTOR TYPE: _EWB-WEC4________________

DATE ADMINISTERED: _Q6/DS/23________________

EXAMINER: _WHIIIEdDBEz_JzlGUILE____ t CANDIDATE: _________________________

l INSIBUGIl0NS_IQ_C8NQ1DalEl i Roed the attached instruction page carefully. This examination replaces tho current cycle facility administered requalification examination, j Rotraining requirements for failure of this examination are the same as {

for failure of a requalification examination prepared and administered by l your training staff. Points for each question are indicated in l l parentheses after the question. The passing grade requires at least 70% l l in each category and a final grade of at least 80%. Examination papers l will be picked up four (4) hours after the examination starts.

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% OF I l CATEGORY  % OF CANDIDATE'S CATEGORY l

! _,_V8LUE_ _IQI8L ___1GQBE___ _V8LUE__ ______________C81EEQBl_____________

' _15tDD__ _2510D ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, I HEAT TRANSFER AND FLUID FLOW I

_15&DD__ _25200 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY i

AND EMERGENCY SYSTEMS

_1510D__ _25z00 ___________ ________ 3. INSTRUMENTS AND CONTROLS

_15ADE__ _21x0Q ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL

! CONTROL

_kDzQQ__ ___________ Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

it__EBINQlELE1_QE_NUQLE8B_EQWEB_EL6NI_QEEB8I1QNi PAGE 2

-IBEBdQQINedIQ1t_HE61_IB8NSEEB_6NQ_ ELM 10_ELQW.

QUESTION 1.01 (2.00)

o. Explain the production and removal of xenon (Xe) in the core (no mathematical equations required). (1.00)
b. After three months of steady state 100% power operations, the reactor trips. After the trip:
1. When will Xe reach its peak concentration in-the core and what will be its maximum reactivity worth? (0.50)
2. When will Samarium (Sm) reach its peak concentration in the core and what will be.its maximum reactivity worth? (0.50) i QUESTION- 1.02 (1.00)

How will cooling the RCS_from 550 F to 450 F affect the Moderator TGmperature Coefficient? BRIEFLY EXPLAIN your answer. (1.0)

QUESTION 1.03 (2.00)

a. List TWO reasons why critical boron concentration decreases over core life. (1.00)
b. How does differential boron worth vary over core life? (1.00)

QUESTION 1.04 (2.00)

If rods are in manual and a steam generator safety valve opens, explain how and why reactor power responds. Assume reactor is at 75% power. (2.00)

(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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iz__EBINQIELER_QE_NVQLE88..EQWEB_EL681_QEE86IIQNt PAGE 3 IBEBdQQ1NedIQ1t_BE6I_IB8NSEEB_eNQ_ELu1D_ELQW QUESTION .l.05 (2.00)

Indicate the. basic problem encountered and the inadequacy of the result obtained (under or overpredicting criticality) for-a fueling 1/M plot

-obtained with the following conditions.

a. A fuel assembly is loaded near the detector. (0.50)
b. The detector is too far from the core. (0.50)
c. The detector is too close to the source. (0.50)
d. The-detector is too far from the source. (0.50)

QUESTION 1.06 (1.25)

For the following parameters, what conditions must exist.to support or indicate natural circulation flow? (1.25)

-1. RCS subcooling

2. Steam generator pressure
3. RCS hot leg temperature

, 4. Core exit thermocouples

  • 5. RCS cold leg temperature QUESTION 1.07 (1.50)

The reactor is at 10% power. What is the initial effect of an increase in oteam demand on steam generator level (increase, decrease, no effect)?

Explain your answer assuming feedwater flow remains constant.

(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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QUESTION 1.08 , (1.00)

Choose. the CORRECT. response. In order to maintain a 200 F subcooling acrgin in the.RCS when reducing RCS pressure to 1600 psig, steam generator pressure must be reduced to approximately:

o. 405 psig
b. -325 psig ,
c. 245 psig
d. 165 psig QUESTION 1.09 (2.25)

Answer the following questions regarding Reactor Safety Limits:

e. State which reactor plant parameters are limited by Reactor Safety Limits. (0.75)
b. What TWO adverse conditions are prevented by adherence to Reactor Safety Limite? (1.00)
c. How are Limiting Safety System Settings (LSSS) related to Reactor Safety Limits? (0.50)

(***** END OF CATEGORY 01 *****)

2t__EL6NI_QESIGN_INCLUQ1NE_16EEIY_6NQ_EMEBEENGl_111IEd3 PAGE 5 QUESTION 2.01 -(1.50)

s. From which busses do the No. 11 and No. 12 Motor Driven Auxiliary Feedwater Pumps draw their power'? (0.50)
b. Which steam generator (s) supplies steam to the No. 13 Turbine Driven 1.uxiliary Feedwater Pump? .

(0.50) c.. List the TWO sources of water that can be supplied to the Auxiliary.

Feedwater System-(AFWS) in the preferred order of use. (0.50)

QUESTION 2.02 (2.00)

Describe the flow path used during Rapid Boration by listing the major components along.the flow path IN THE CORRECT ORDER starting from the o cource and going to the RCS entry point.

' QUESTION 2.03 (2.00)

c. . What effect, if any, does a Phase-A Containment Isolation have on the seal leakage flow path? Can the RCP's continue to be operated in accordance with procedures? (1.00) b.. What-components of a RCP are adversely affected by a Phase-B Contain-ment Isolation? Can the RCP's continue to be operated in accordance with procedures? (1.00)

QUESTION 2.04 (2.00)

a. What buses feed Unit l's Residual Heat Removal (RHR) pumps? (1.00)
b. What modes of operation can the RHR system provide as part of the Emergency Core Cooling System (ECCS)? (1.00)

QUESTION 2.05 (2.00)

During a continued RCS depressurization caused by a LOCA, indicate the order in which the ECCS subsystems will inject into the RCS and the (2.00)

(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

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. QUESTION 2.06- (2.00) ,

For the following components, Jndicate whether-they will receive an OPEN, CLOSE, or N0 signal upon's MANUAL SAFETY INJECTION INITIATION.

c. Control Room outside air isolation valves' (0.20)
b. Main Feedwater bypass valves (0.20) c .- Cold. Leg Accumulator isolation valves (0.20)
d. Charging header isolation valves (0.20) o.- Main steam isolation valves (0.20)
f. RWST.to centrifugal charging pumps suction valves (0.20)
g. RCP seal water return isolation valve (0.20)
h. CCW isolation valve from RHR heat exchanger (0.20)
1. CCW isolation valve from letdown heat ~ exchanger (0.20)
j . - Steam supply valves to turbine-driven AFW pump ,

(0.20)

LQUESTION' 2.07 (1.50)

Under NORMAL OPERATING CONDITIONS at 100% power in what position will the following valves-fail on loss of control air pressure?

1. Letdown orifice isolation valves.
2. VCT level control. valve (LCV-ll2A)
3. Charging flow control valve (HCV-182)
4. Pressurizer power operated relief valves.
5. Steam supply valves to the turbine driven AFW pump.
6. SG pneumaticly operated relief valves QUESTION 2.08 (2.00)
e. What three design accidents is the Containment Spray System (CSS) designed to mitigate? What containment parameters does the CSS maintain?

(1.25)

b. Give the sources of water supplied to the containment spray

. pumps, the sequence in which they are used and the method of sequencing (automatic or manual). (0.75)

(***** END OF CATEGORY 02 *****)

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- 2x__IN11BMMENI1_6NQ_CQNIBQL$ PAGE- 7

- QUESTION 3.01 (2.00)

c. At what pressure is SI blocked during a controlled cooldown? (0.50) b.- How does the operator know that the block permissive is activated? (0.50) c.- How does the operator block SI? (1.00)

QUESTION 3.02 ( .75)

State which region on the gas filled detector curve each of the following detectors operate in:

1) Power range detector (0.25)
2) Inner volume of the intermediate range detector (0.25)
3) Source range detector (0.25)

QUESTION 3.03 (2.00)

The reactor is operating at 40% power steady state with four loops.

00 scribe the-initial automatic actions for the following occurences

_(elarms not required):

c. 2/4 low-low level in 1/4 SG's (0.50)
b. 2/4 low-low level in 2/4 SG's (0.50)
c. 2/3 hi-hi level in 1/4 SG's (0.50)
d. Steam pressure input to the Main Feed Pump speed control system fails high.(Not the steam flow signal) (0.50)

QUESTION 3.04 (2.00)

c. Explain the purpose of the steam pressure input used in the development of:a steam flow signal for the S/G water level control system. (1.00)
b. How would INDICATED steam flow compare to ACTurL cteam flow if, during a power increase from 0-100%, the steam pressure signal stuck at it's 50% power value. (1.00)

(***** CATEGORY 03 CONTINUED ON NEXT PAGT *****)

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2i__IN11BudENI1_8NQ_GQNIBQL1 PAGE 8 QUESTION 3.05 (1.50)

The plant is at 100% power. The' turbine impulse pressure selector switch 10 in its normal position., What will be the response of the steam dump cystem (SDS) if PT-506 fails low? Explain why.

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j QUESTION 3.06 (2.25)

The plant is operating at 100%:oower. The normally selected channel to the pressurizer level. control system fails low.

s. What immediate component responses will be initiated by the channel failure? (1.75)
b. 'If no operator action isitaken, what will eventually trip the plant?

(0.50)

-QUESTION 3.07 (2.00)

Indicate what automatic actions, if any, will occur for the following

'Rodistion Monitoring System alarms. '

c. -High level on the plant vent stack.
b. High level alarm on the steam generator sample line.
c. . High level alarm on the liquid waste processing drain channel B
d.' High level on the Control Room air supply monitor with the ventilation system in normal operation.

l l QUESTION 3.08 (1.50) i l

e.. What~ interlocks prevent the operator from opening the RHR suction from RWST isolation valves (8812A and 8812B)? (1.00)

b. What interlocks prevent the operator from opening the containment sump oto RHR pump isolation valves (8811A and 80118)? C0.50)

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

2t__INSIBudENIS_8NQ_GQUIBQLH PAGE 9-

' QUESTION 3.09 (1.00)

What are four containment parameters available.to the operator as

-pest accident monitoring indication? (1.00) l i

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(***** END OF CATEGOPY 03 *****)

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at__EBQGEQUBE1_ _NQBd8Lt_6HNQBd8Lt_EdEBGENGX_6NQ PAGE 10 88DIQLQQ196L_GQUIBQL QUESTION 4.01 (2.50)

A 24 year old man, with a life time exposure through the last quarter of 23 rem, will be working in a 200 mrem /hr radiation field. In addition to lifetime dose, he has received 1100 mrem in the present quarter and 200 arem in the present week.

o. Who's approval is needed, in this case, to increase this individual's exposure limit? (1.00)
b. How long may this individual work before he reaches the maximum QUARTERLY limit allowed at Comanche Peak under these conditions? (0.50)
c. What provisions must be met to allow an individual,'in non-emergency situations, to increase the quarterly Comanche Peak administrative limit? (1.00)

QUESTION 4.02 (1.50)

-With respect to the RCS state the NORMAL PROCEDURAL cooldown limit and the TECH SPEC limit as well as the basis for the TECH SPEC limit. (1.50)

QUESTION 4.03 (1.00)

Which of'the following statements is correct concerning the status of the Nuclear Instrumentation Recorder (1-NR-45) during control bank rod withdrawal to criticality?

c. Both source range channels and the highest reading intermediate range channel are selected.
b. The highest reading source range and either intermediate range channel are selected.
c. Either source range and the lowest reading intermediate range are selected.
d. Either source range and both intermediate range chanels are selected.

on the recorder.

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

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st__EBQQEQUBER_:_NQBdekt_eBNQBdeLt_EMEBGENCY_6NQ PAGE 11 88010LQQ196L_QQNIBQL QUESTION 4.04 (3.00)

Answer the following questions utilizing information that would be found in FRS-0.1, Response to Nuclear Power Generation /ATWS.

c. What three conditions must you observe to verify a reactor trip? (1.00) b.- What immediate action must be taken if the reactor will not trip on a manual reactor trip initiation? (0.50)
c. What actions are required to initiate emergency boration of the RCS?

(1.50)

QUESTION 4.05 (1.50)

e. What is the condenser low vacuum alarm setpoint? (0.25)
b. Identify three conditions which will result ina loss of condenser vacuum. (0.75)
c. What automatic action will result if the' operator is unable to restore vacuum and when will it occur? (0.50)

QUESTION 4.06 (2.50)

In accordance with E0P-2.0, FAULTED STEAM GENERATOR ISOLATION, what are five-of the seven actions required to isolate a steam generator. (2.50)

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'(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

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'st__EBQQEQWBES_=_NDBd8Lt_8BNQBd8Lt_EMEBGENQX_8NQ PAGE 12 88Q10LQQ198L_QQNIBQL QUESTION 4.07 (1.00)

In'the process of determining if SI can be terminated, it is' determined that the secondary heat. sink is available. In which of the following cituations could SI be terminated?

PZR LVL SUBC00 LING PRESSURE

c. 4% 60 degrees. stable
b. 25% 10 degrees increasing
c. 10% 40 degrees decreasing
d. 20% 15 degrees stable QUESTION 4.08 (2.00)
a. Once in the E0P's, what are four conditions that result in trans-itioning to EOP-3.0, Steam Generator Tube Rupture? (1.60)
b. What is the purpose behind isolating the ruptured steam generator?

(0.40) i

(*****.END OF CATEGORY 04 *****)

(************* END OF EXAMINATION ***************)

iz__EBINQ1ELEl-QE NUGLE88 EQWEB EL8NI-QEEB8110Nt PAGE 13 IBEB50018851Q1t BE81 IB8NSEEB_8ND ELMID ELQW

. ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 1.01 (2.00)

o. Xe is produced directly as a fission product (0.25) and also indirect-ly from the decay of Te-135 to I-135 which decays to Xe-135 (0.25). Xe is removed by decay (0.25) and by burnout-(0.25). (1.00)
b. 1. 8 to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (0.25) 6300 + or - 200 pcm(0.25) (0.50)
2. 15 to 17 days (0.25) 1065 + or - 100 pcmCO.25) (0.50)

(360 to 408 hours0.00472 days <br />0.113 hours <br />6.746032e-4 weeks <br />1.55244e-4 months <br />)-

REFERENCE Roactor Core Control Ch. 4 pgs. 4-11 to 4-13 3.1 001 000 K 5.13 3.7 5.38 3.5 ANSWER 1.02 (1.00) f LESS NEGATIVE (0.50) Water density changes are less as tcmperature is reduced.(0.50) (1.00)

REFERENCE Roactor Core Control, pgs. 3-17,3-18,3-20,and 3-25 3.1 001 000 K 5.15 3.4 ANSWER 1.03 (2.00)

e. 1) Fission poison buildup. (0.50)
2) Fuel depletion. (0.50)
b. Differential boron worth becomes less negative from BOL to EOL for a constant boron concentration. (Also accept becomes more negative due to decreasing boron concentration over core life.) (1.00)

REFERENCE CP Reactor Core Control, Ch. 3, pg. 26 and Ch. 5, pg. 14 3.1 001 010 K 5.21 3.4

-'li__EBINQIELES_QE_NVQLE88_EQWEB_EL8NI_QEEB8IIQN t PAGE 14 IEEBdQQINedIQSt_BE81_IB8NSEEB_8ND_ELVIQ ELQW ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 1.04 (2.00)

-When steam demand increases Tcold will decrease which will cause T-ave to decrease (0.50). When T-ave decreases Moderator Temperature Coefficient cdds positive reactivity which causes reactor power to increase (0.50).

When reactor power increases fuel temperature increases which adds negative roactivity from the doppler coefficient (0.50). Reactor power will increase until the reactivity changes from MTC and doppler are equal (0.50).

REFERENCE Thermal-Hydraulic Prinicples II, Ch. 12, pgs. 12-39 t0 12-43 3.1 001 000 K 5.29 3.7 ANSWER 1.05 (2.00)

e. Loading a fuel assembly into the core close to a neutron detector increases the fraction of neutrons in the core reaching the detector.

The detector count rate increases more than the core neutron population increases (0.25). Criticality is under predicted (0.25). (0.50)

b. The detector will not see neutrons until there are a great number.

(0.25) Criticality is over predicted. (0.25) (0.50)

c. The initial' count rate is too high and the detector is insensitive to core changes. (0.25) Criticality is over predicted. (0.25) (0.50)
d. Initial flux level is low so that ICRR is low. (0.25) Criticality is under predicted. (0.25) (0.50)

REFERENCE Fund. of Nuclear Physics, pg. 8-27 to 8-35

, 3.1 001 010 K 5.16 2.9 f

.I' .

Iz__EBINQIELES_QE_NMQLEeB_EQWEB_EL8NI_QEEBellQNt PAGE 15 IHEBUQQ1Ned1Qft_BEeI_IB8NSEEB_eND_ELu1R_ELQW I

. ANSWERS '-- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL

' ANSWER 1.06 (1.25)

1. Greater than 15 degrees F (0.25)
2. Stable or decreasing (0.25)
3. Stable or decreasing (0.25)
4. StC le or. decreasing (0.25)
5. At saturation temperature for SG pressure (0.25)

REFERENCE CP EOS-0.1 ATTACHMENT 2 3.4 000 015 EK 1.01 4.4 ANSWER 1.07 (1.50)

Increase (0.50). As the mass flow rate from the SG increases, the pressure in the SG is quickly reduced, and more vapor bubbles are formed (0.50).

The rapid formation of bubbles carries additional water into the moisture coperators (and! increases resistance to flow'in.the downcomer) causing the .

downcomer level to increase above normal (0.50).

REFERENCE Thermal-Hydraulic Principles pg. 12-52 3.4 035-010 k5.03 2.8 ANSWER 1.08 (1.00)

c --_- (1.00)

REFERENCE Steam Tables

. . - - - _ . - _ _ _ _ . . __,,c_ , . , _ , , , , , . - _ , ,..-._._ _ , _, ___.,_,.__,,s._._,.__,,__,_,,, ,

. '~ .

e it;_EBINQIELER_QE_NMQLE88_EQWEB_EL8HI_QEEB6110Nt PAGE 16 IBEBdQQ1NoMIQ1t_BE8I_IB8NSEEB_6NQ_ELu1R_ELQW cANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 1.09 (2.25)

c. 1) Thermal Power (0.25) 2).RCS pressure (0.25)
3) Tave (0.25)
b. 1) DNB (0.50)
2) Loss of RCS integrity (0.50)
c. LSSS's are automatic' protective device setpoints(0.25) which are chosen so that protective action will prevent exceeding a Safety Limit.(0.25)

(0.50)

REFERENCE Thermal-Hydraulic Prin. II, pg. 13-53 Toch *pec 2.1 and 2.2 2.1 System-wide Generic 5 2.9

~* '

2x__EL8NI_ DESIGN _INGLVQ1NG_S6EEIl_aND_EdEBGENGl_SYSIEd3 PAGE 17 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 2.01 -(1.50) 0.- No. 1EA1 6.9KV Safeguards Bus (0.25)

No. 12 - 1EA2 6.9KV Safeguards Bus (0.25)

b. 1 and 4 (0.50)
c. 1. Condensate Storege Tank. (0.20)
2. Service water (0.20) (0.10 for preferred order)

. REFERENCE CP SD VIII, pgs. 8.6 and 8.8 P&ID - El-0004 3.5 061 000 K 2.02 3.7 K 1.03 3.5 K 4.01 3.9 Sys Gen 5 3.3 ANSWER 2.02 (2.00)

Boric acid tanks (0.20)

Boric acid transfer pumps (0.20)

Emergency boration control valve (HV-8104) (0.20)

Charging pumps (0.20)

Charging flow control valve (HCV-182) (0.20)

Charging line isolation valves (8105 and 8106) (0.20)

Rogenerative Heat Exchanger (0.20)

Charging isolation valve (8146) (0.20)

Enters the RCS via #4 cold leg (0.20)

(for correct order 0.20)

(2.00)

REFERENCE CP SO Vol II, pgs. 2.19,2.32,3.26, and 3.42 3.1 004 010 K 6.09 4.4

-g- p- - - - ,

p -.g, ,,mn , --.-g. 7

T',- . .

2t__EL8NI_DE11GN_INGL9Q1NG_18EEI1_8NQ_EMEBGENRY_1111Ed3 PAGE 18 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL 1 ANSWER 2.03 (2.00)

O. 1) Seal leakoff cannot exit containment. It exits the system through a relief valve upstream of the isolation valves (0.37) and goes to the PRT (0.38). (0.75)

2) Yes (0.25)
b. 1) Upper motor bearing (0.20)

Motor windings (0.20)

Lower motor bearing (0.20)

Thermal barrier (0.20) (0.80)

2) No (ABN_101A requires.the RCP be secured within one minute of loss of thermal barrier cooling.) (0.20)

REFERENCE CP SD Vol II, pgs 2.22 and 12.10 ABN-101A, pg. 15 3.10'008 000 K 3.01 3.4 3.4 003 000 K 4.11 3.0 ANSWER 2.04 (2.00)

e. 11 - 1EA1 6.9KV Safeguards Bus (0.50) 12 - 1EA2 6.9KV Safeguards Bus (0.50)
b. Low head inj ection during inj ection phase (0.25)

Recirculate water from the containment sump back to the RCS during recirculation phase (0.25)

Provide suction to the: High head CC pumps (C.25)

Intermediate head SI pumps (0.25)

REFERENCE P&ID El-0004 CP SD Vol II, pgs. 6-19 to 6-21 3.4 005 000 K 2.01 3.0 System generic 3.6 K 4.02 3.2 1

, . - . - - , - . . , c , - _ . _ _ - - . - --

-Zu._EL6NI_ DESIGN _INGLUDING_S6EEI1_8NQ_EMEBSENQ1_1111Ed1 PAGE 19 ANSWERS -_ COMANCHE PEAK'l _86/09/23-WHITTEMORE, J./GUIL

.SWER 2.05 (2.00)

High head inj ect ion (0.20) 1829 psig(immediate) (0.25)

Intermediate head inj ect ion (0. 20 ) 1520 + or - 50 psig (0.25)

Accumulators (0.20) 650(603 to 686) . psig (0.25)

_Lew head inj ect ion (0.20) 195 + or - 20 psig (0.25)

(0.20 for.the correct order) (2.00)

REFERENCE CP SD Vol II, pgs.- 6.19, 8.24, and 8.10 to 8.18 3.2 006 000 K 4.05 4.3 K 6.02 3.4 K 6.03 3.6 ANSWER 2.06 (2.00)

(0.20 pts each)

c. CLOSE
b. CLOSE
c. OPEN
d. CLOSE
o. N0
f. OPEN 9 .CLOSE
h. THROTTLE OPEN 1.. NO
j. NO REFERENCE CP SD VOL II, pg. 7.19 Table ESF-1 3.2 006 000 K 4.09 3.8

2t__EL6HI_QEllEN_INGLUDING 18EEI1_aND_EdEBGENQ1_111IEd3 _PAGE 20

ANSWERS 1-- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL' ANSWER 2.07 (1.50)

(0.25 each)

1. Fail closed 2.EFail flow-to the VCT
3. Fail open1
4. Fail'as'is (operated by nitrogen)
5. Fail open
6. Fail closed (provided with backup air accumulator) (1.50)

REFERENCE CP SD VOL II, pgs. 2.7, 2.13, and'2.19 P&ID M1-0202, 0251 and 0262 3.8 078 000 K 3.02 3.4 ANSWER 2.08 (2.00)

.c. 1) Design basis LOCA (0.25)

Steam line rupture (0.25)

Feed line rupture (0.25)

2) Pressure (0.25)

Temperature (0.25)

b. 1) RWST_ (0.25)
2) Containment Sump (0.25) 3)' Manual (0.25)

REFERENCE CP SD Vol-II, pgs. 10_2 and 10_14 P&ID El-0004 3.6 026 000 K 4.04 3.7 3.6 028 000 System generic 4 3.5 K 5.03 2.9

at__INSIBudENIS_6NQ_GQNIBQLS PAGE 21 ANSWERS -_-COMANCHE PEAK 1 _86/09/23-WHITTEMORE, J./GUIL ANSWER 3.01 (2.00)

c. Below 1960 psig and above 1829 psigCany value in between). (0.50)
b. The PRZR PRESS SI BLK PERM P_11 blue annunciators come on- (0.50)
c. Place both STEAMLINE SI BLOCK switches to the BLOCK position (0.50) and both PRESSURIZER SI BLOCK switches to the BLOCK position (0.50).

REFERENCE CP IPO-005A, pg. 13 3.2 006 000 K 4.13 4.1 ANSWER 3.02 ( .75)

1) Ionization '(0.25)

.2)

Recombination (0.25).
3) Proportional (0.25)

REFERENCE CP SD VOL III, pgs. 1.4 and 1.5 3.9 015 000 K 6.111 2.9

n

-2i__IN11BVMENII_6NQ_GQNIBQLS PAGE 22 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 3.03 (2.00)

G. Reactor-trip (0.25) and start both MD AFW pumps (0.25).

b. Reactor trip (0.25) and start all AFW pumps (0.25).
c. Turbine trip (0.17), FW' system isolation (0.17), MFP trip (0.16).
d. Main Feed Pump speeds up (control system see's large decrease in delta P and acts to raise it to conform with programmed delta P.

(0.50)

REFERENCE CP SD VOL III, pgs. 8.7, 9.65 and 9.66 3.9 016 000 K 1.12 3.5' A 2.01 3.0 ANSWER 3.04- (2.00)

c. Steam pressure is'used to compensate.the steam flow signal for density variations in the steam as power-increases. (1.00)
b. Indicated steam flow will be higher than actual. (1.00)

REFERENCE CP SD VOL III, Table RPS-4 3.2 SYS GEN 4 3.8

'21__INIIBUNENI1_6NQ_QQNIBQLf; PAGE 23 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 3.05 (1.50)

The system will arm (PT506 feeds the arming signal) (0.75) but there will bo no demand signal because PT505 provides the_ Tref input for the demand-signal (0.75). (1.50)

REFERENCE-CP ABN-709A

.3.9 016 000 K 3.03- 3.0 A 2.01 3.0 ANSWER 3.06 (2.25) e.- CVCS letdown' isolation valve closes (0.43)

All CVCS letdown orifice isolation valves close (0.43)

All pressurizer heater groups are turned off (0.43)

Increased charging flow (0.46) (1.75)

b. Reactor will trip on high pressurizer level at 92%. (0.50)

REFERENCE CP SO VOL III, pg. 6.7 3.2 011 000 K 3.01 3.2 K 4.05 3.7 A 2.01 3.0

~2t__IN11BudENI1_eNQ_GQNIBQLS PAGE 24 ANSWERS -- COMANCHE PEAK-1 -86/09/23-WHITTEMORE, J./GUIL

. ANSWER 3.07 (2.00)

c. 1) Initiates Control Room emercancy recirculation. -(0.25)
2) Closes HCV-014 in the GWPS. (0.25)

(also accept containment ventilation isolation for no credit.)

b. 1) Isolates SG blowdown. (0.25)
2) Isolates blowdown sample lines. (0.25)
c. ' Closure of discharge valve to CST (RCV-5252) (0.50)
d. Initiates Control Room ventilation emergency recirculation (0.50)

REFERENCE CP SD VOL III, pg. 12.7 3.11 029 000 K 1.01 3.4 068 000 K 6.10 2.5 000 036 EA1.02 3.1 000 059 EA2.05 3.6 ANSWER 3.08 ( l '. 5 0 )

e. Either sump isolation valve (8811A or 8811B) is open. (1.00)
b. The RWST to RHR pump isolation valves (8812A and 88128) are closed and one of the two RCS suction series valves (8701A & 8702A or 87018 &

8702B) are closed. (0.50)

REFERENCE CP SD VOL II, pg. 8.36

-3.2 006 000 K 4.08 3.2 K 4.06 3.9 3.6 026 020 K 4.03 4.1 4

I

c

.?" .

, ,e 2t__INSIBudENI1_8NQ_QQNIBQL1 PAGE 25-ANSWERS - COMANCHE PEAR 1 -86/09/23-WHITTEMORE, J./GUIL

. ANSWER- .3.09 (1.00)

1. Pressure

.2. Radiation

'3. Sump Level

4. H2 (0.25 each)

REFERENCE CP SD VOL III, pgs. 18.11 -18.14 3.6 103 000'A 1.01' 3.7

D :. . ..- . -

. .z.- _

Liz__EBQGEQUBE1_:_NQBueLt_8BNQB56Li_EMEBEENGY_8NQ PAGE 26 88Q1QLQGIG8L_GQNIBQL

. ANSWERS--- COMANCHE PEAK 1 -86/0g/23-WHITTEMORE, J./GUIL ANSWER 4.01- _(2.50) c .' The Radiation Protection Engineer-must approve exceeding the

-administrative quarterly limitCO.50). The Radiation Protection Supervisor must: approve exceeding the administrative weekly limit.

(If assumed'1300 mrem total dose received for quer,ter then only1the Red-

-iation Protection Supervisor's approval is needed) (0.50) .

(1.00)

.b. '3000 mrem /Q'= 1100 mrem + 200 mrem /hr X T ; T'= 9.5-hours (0.50) 1300 mrem +'200 mrem /hr-X T ; T = 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

c. Dose to whole body should not exceed 3 rem per quarter (0.33)

The SCN - 18)-limit must not be exceeded (0.33)

The individual's exposure history is documented on NRC Form 4 .(0.34)

REFERENCE 10 CFR 20.101 b AP_24 pg 13 System wide and plant wide generic knowledge 15 3.5 ANSWER 4.02. (1.50)

The normal cooldown rate for the RCS is 50 degrees /hr(0.50). The RCS cool-down rate shall not exceed 100 degrees F/hr(0.50). This limit is imposed

.to minimize thermal stresses (0.50). (1.50)

REFERENCE' CP IPO-005A, pg. 3 TECH SPEC pg. 3/4 4-30 and 4-47 3.2 002-020 Sys Gen 5 2.9 t

1 i

. . _ . , . - . . , - - - - - . _ . . . . , . , . _ . . . .,.,m. . . _ _ , , ,,,.,...._.._r,_._.#-.,-._.-,__m.,_..-.-_,,,_ - , - . , .

r- '

-e st__880GEQMBEQ_ _NQ856Lt_8BNQBd8ke._EMEEGENQ1_6NQ PAGE 27 88DIQLQQ1Q8L_QQNIBQL ANSWERS -- COMANCHE' PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 4.03 (1.00)

--b.-- (1.00)

REFERENCE CP IPO-002A, . pg. 6 3.9 015 000 3.01 3.8 ANSWER 4.04 (3.00)

c. 1) Reactor trip and by-pass breakers open (0.34)
2) Neutron flux decreasing (0.33)
3) All DRPI rod bottom lights on (0.33)
b. Manually-insert control rods (accept rapid boration) (0.50)
c. 1) Start both boric acid transfer pumps. (0.50)
2) Open emergency boration valve 1/1-8104 (0.50)
3) Verify emergency boration flow (accept start CCP's) (0.50)

REFERENCE CP FRS-0.1, pg. 3 3.1 000 007 EA 1.06 4.4 EA 2.04 4.4

E..- ,. .. ]

=+ *

'st__EBQQEQMBER_: NQBdekt_aBNQBd8Lt_EdEBQENQ1_6NQ PAGE 28 88DIQLQGI98L_QQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER -4.05 (1.50)

s. 22.6 + or - 1 in. Hg vacuum (0.25)
b. 1) Loss of circulating water
2) High condenser water level
3) Loss of gland seal steam system
4) Loss of CEV pumps
5) Air leakage Cany three for 0.75)
c. Turbine trip (0.25) at 21 in. Hg vacuum (0.25) (0.50)

REFERENCE CP Alarm Procedure 1-ALB-90 ANSWER 4.06 (2.50)

1) Isolate main steam line.
2) Isolate main feed line.
3) Isolate AFW flow.
4) Place TDAFW pump steam supply. valve in PULL-TO-LOCK.
5) Isolate blowdown and sample lines.
6) Verify GG steam line atmospheric dump valves closed.
7) Verify main steam line drip pot isolation valves closed.

(Any 5 at 0.5 ea.] (2

.50)

REFERENCE E0P-2.0,'pgs. 3 and 4 3.5 000 040 EK 3.04 4.5

3z_;EBQQEQUBE1_ _NQBdekt_8ENQBdeLi EMEBQENQ1_6ND PAGE .29 88DIQLQGIQ8L_CQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23_WHITTEMORE, J./GUIL i

ANSWER 4.07 (1.00)

--d.-- (1.00) l REFERENCE '

E0P_0.0, pg. 11 3.2 006 020 K 4.06 3.9 ANSWER 4.08 (2.00)

o. Condensate Exhausting Vacuum (CEV) pump radiation readings abnormal SG blowdown radiation readings abnormal Main steamline radiation readings abnormal SG sample radiation readings abnormal SG water level increasing (any four for a total of 1.60)
b. 1) To isolate flow from the ruptured steam generator to minimize radio-

, logical release. (0.20)

2) To maintain pressure in the ruptured steam generator greater than the pressure in at least one intact steam generator following cool-l down of the'RCS. (0.20)

REFERENCE E0P-0.0, pg. 10 i E0P-3.0, pg. 2 l

ERG-HP_ Background E_3 ECA-3 l

l 3.3 000 037 EK 3.07 4.2 L ___________________________________________________________________________

t

+ \

s U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY: _QQd8NQBE_ PEAK _Il________

REACTOR TYPE: _EW8-WEQ4________________

DATE ADMINISTERED: _H6402422________________

EXAMINER: _WBlIIEdQ8Et_JtfKINQ1____

CANDIDATE: _________________________

INSIBVQIl0NS_IQ_QaNQIQ8IEi Road the attached instruction page carefully. This examination replaces tha current cycle facility administered requalification examination.

Rotraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff. Points for each question are indicated in pcrentheses after the question. The passing grade requires at least 70%

in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.

% OF

. CATEGORY  % OF CANDIDATE'S CATEGORY

__V8LVE_ _IQI8L ___SQQBE___ _Y8LVE__ ______________Q6IEQQBY

_1510D__ _25tQQ ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_15t00__ _25tDQ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_1510D__ _251QQ ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_15t00__ _251Q0 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS

_hQtQQ__ ___________ Totals Final Grade

.All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature k

a 3;.; ;,

g e .. e Et__IBEQBX_QE_NWQLE88_EQWEB_EL8NI_QEE86IIQNt_ELUIQSt_6NQ PAGE 2 IBEBdQQXNe5101 a

1

' QUESTION 5.01- (2.00)

c. The reactor is subcritical by 2.5% delta-k/k. The count rate is 115 CPS. After a positive reactivity' insertion, the count rate increases to 345.- How much reactivity was added to the core? (1.5)
b. 'Why'does-it take longer, after.each reactivity' addition, for the neutron population to reach equilibrium as Keff approaches 1.07 (0.5)

QUESTION 5.02 (l'.00)

During a reactor startup, the first reactivity addition' caused count rate to' increase from 20 cps to 40 cps. The second reactivity addition caused count rate to increase from 40 cps to 80' cps. Which of the following otatements is CORRECT?-

o. The first reactivty addition was larger.
b. The second reactivity addition was larger.

c.- The first and second reactivity additions were equal.

d. There is not enough data given to determine relationship of reactivity values.

-QUESTION ~ 5.03 (2.50)

o. List the three most significant contributors to total power coefficient in order of INCREASING magnitude. (1.5)
b. How does' total power coefficient vary as the core ages? (1.0)

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

St__IBEQBl_QE_NUCLEaB_EQWEB_EL8HI_QEEB8Il0Nt_ELu1 dst _8NQ PAGE 3 IBEBdQQ1Ned1GH

. QUESTION 5.04 (3.00)

For the following situations, indicate whether the. final stable pcwer level will be HIGHER, LOWER, or THE.SAME as the initial power level. EXPLAIN your answers. Assume the initial power 1cvel is at approximately 10 % following a normal reactor-otartup at the end of life. Consider each situation separately.

a. Steam dump pressure setting is lowered by 20 psig while in Steam Pressure mode. (1.0)
b. A small (1 %) main steam leak develops inside containment that is insufficient to initiate SI or Containment Spray. (1.0)
c. RCS boron concentration is increased by 5 ppm. (1.0)

QUESTION 5.05 (2.00)

c. What is the operating definition of Shutdown Margin? (1.0)
b. The plant is operating at 85% power with all systems in automatic.

The operator inadvertently aligns charging pump suction to the RWST.

How is shutdown margin affected PRIOR to a reactor trip? (1.0)

QUESTION 5.06 (1.50)

The plant is operating at 100 % power with RCS Tave at 587 F cnd a steam pressure of 980 psig. What must TAVE be changed to in order to maintain these conditions with 10 % of the tubes in each steam generator plugged? SHOW ALL WORK, including any applicable formulas.

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

4

  • S t _ _IllE Q Bl_ Q E _N U Q L E 68_ E QW E B _ E L 8 NI_Q E E B 6110 N t _E L V ID S t _6N D PAGE 4 IBEBdQQ1NedIQS-QUESTION 5.07 (1.00)

Choose the CORRECT response. In order to maintain a 200 F subcooling ccrgin in the RCS when reducing RCS pressure to 1600 psig, steam generator pressure must be reduced to approximately:

a. 405 psig
b. 325 psig
c. 245 psig
d. 165 psig QUESTION 5.08 (2.00)

The reactor-is producing 100% rated thermal power at a core delta-T of 60 degrees and a RCS mass flow rate of 100% when a station blackout

'eccurs. Natural. circulation.is established and core delta-T goes to 28 F.

If decay heat is 2%, what is the core mass flow rate (in %)?

e

(***** END OF CATEGORY 05 *****)

r -.- .,.- , , .;

PAGE 5 ht__EL8HI_111IEb1_QE11GNt_QQNIBQLt_8NQ_{N11BubENI611QN

. QUESTION 6.01 (1.50)

The plant is operating at 100 % steady state power with centainment pressure channel IV (PB 934A) failed hfgh. A tcchnician troubleshooting the trip bistables inadvertently dO-energizes the instrument power for containment press ~ure channel II. Will a Containment Spray Actuation-occur?. (0,5)

WHY or WHY NOT? , (1.0)

QUESTION 6.02 (2.00)

Indicate which of the Excore Nuclear instrumentation Ranges (SOURCE, INTERMEDIATE, or POWER), will correctly match with the following statements. More than one may apply to each.

c. Provides a direct input to*the Rod Control System.
b. Has a reactor trip function that is blocked at some time between startup and fullTpower operation,
c. Utilizes a Boron-10 coating in.it's detectors._

(

d. Operates in the " Ion Chamber" region of the ." Gas Filled Detector Characteristic Curve". e QUESTION 6.03 (2.50)

The following concern the CVCS.

c. What are the TWO functions (purposes) of the Letdown Pressure Control Valve (PCV-131)? (1.0)
b. If left in automatic control, what position'should PCV-131 be found in two minutes after assafety inj ect ion initiation? (0.5)
c. Why is letdown flow limited to 120 gpm? (0,5)
d. With only the positive displacement pump operating at power, which valve (s) is/are utilized to control RCP seal inj ect ion flow? (Noun name(s) or number (s) is/are acceptable.) (0.5)

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

-ht__EL6NI_111IEdi_DE11DNt_QQNIBQLt_6NQ_INSIBMMENI6IIQN PAGE 6,'

-QUESTION 6.04 (1.50)

O. Describe the safety-related function of the Condensate Storage Tank.

(1.0) b.. How is the minimum CST water volume required by Technical Specifications ensured to be-available at all times considering that several systems take suction on the' CST?. (0.5)

QUESTION 6.05 (1.50)

Solect from the following list of electrical loads, THREE loads which would bo deenergized following the loss of the 1A4 bus.

s. _. Centrifugal charging pump #12
b. Reactor coolant pump #14
c. Condensate pump #12
d. Containment spray pump #14
o. Turbine building cooling water pump #12
f. Service air compressor #11 QUESTION 6.06 (1.00)

The plant is critical at the point of adding heat during a reactor ctartup. A malfunctioning ^ steam header pressure controller causes six Oteam dump valves to fully'open. Assuming the reactor does not trip, et what average temperature (Tave) will the Reactor Coolant System etabilize? Why?

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

t

, e *

  • 6t__EL8NI_SYSIEMS_QESIGNt_QQNIBQLt_8NQ_INSIBudENI6IIQU PAGE 7 QUESTION 6.07 (1.00)

Following a loss of offsite power with a safety inj ection signal, which of the following abnormal conditions, if occurring separately, will result in o diesel generator trip? (More than one answer may be correct.)

c. Excessive vibration
b. Generator differential
c. Generator reverse power
d. Low lube oil pressure
o. Overspeed
f. High jacket water temperature i

QUESTION 6.08 (2.00)

c. Describe an IR instrument response if the circuitry is undercompensated during a reactor shutdown, including any effects on SR instrumentation. Include any applicable setpoints. (1.0)
b. What operator action (s) is required to continue a reactor shutdown if one IR channel has failed high ? (1.0) l QUESTION 6.09 (2.00) i
c. List two indications or symptoms which will be observed if a tube leak occurs in a RCP thermal barrier heat exchanger. Assume NO alarm setpoints are reached. (1.0) l
h. If the tube leak continues to increase in severity, what AUTOMATIC action will occur and why will it occur to minimize its effect on the rest of the CCW system? (1.0)

(***** END OF CATEGORY 06 *****)

~~

Zi__EBQGEQUBEl_=_NQBd8Lt_8BNQBd8Lt_EdEBGENGl_8HD PAGE 8 88010LQQ1G8L_GQNIBQL QUESTION 7.01 (2.00)

During a loss of all AC power, ECA 0.0 has the operator depressurize the RCS using steam generator atmospheric relief valves.

c. How should these valves be operated without control air? (0.5)
b. Why must the RCS be depressurized? (0.5)
o. Why shouldn't the RCS be depressurized below 170 psig? (0.5)
d. Regarding the depressurization, how should the operator respond if pressurizer level is lost or vessel head voiding occurs? C0.5)

QUESTION 7.02 (1.00)

Solect the group of indications which provide verification of natural circulation.

RCS SG CORE EXIT SUBC00 LING PRESSURES Thot Teold THERMOCOUPLES

s. 30 F Constant Decreasing Decreasing Increasing
b. 25 F Decreasing Constant Constant Increasing
c. 20 F Constant Decreasing Constant Decreasing
d. 15 F Decreasing Increasing Decreasing Increasing
o. 10 F Constant Constant Constant Decreasing QUESTION 7.03 (2.00)

In the event that both boric acid transfer pumps are inoperable, how can RCS boration be accomplished (lineup) and controlled (flow incr(ased and docreased) until the boric acid pumps are repaired?

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

Zi__EBQGEQUBES_:_NQBd8Lt_8HNQBd6Lt_EDEBGENGl_6ND PAGE 9 86D10LQQ1G6L_GQNIBQL QUESTION -7.04 (2.50)

Supply the following which must be observed during plant operatioti in occordance with INTEGRATED OPERATING PROCEDURES.

c. Maximum RCS cooldown rate (0.50)
b. Maximum pressurizer cooldown rate (0.50)
c. Maximum boron concentration differential between RCS and pressurizer (0.50)
d. Maximum differential temperature between pressurizer and spray fluid (0.50)
o. Minimum RCS subcooling (0.50)

QUESTION 7.05 (2.50)

c. Complete the following statements regarding a reactor startup.
1. The reactor must be critical no greater than ________(0.50) hours following the Estimated Critical Concentration calculation.
2. RCS Teve must be verified to be greater than ________(0.50) F every

_______(0.50) minutes until criticality is reached.

3. Source range counts munt read at least ________(0.50) cps on the highest channel prior to rod withdrawal.
b. List TWO alarms which are expected to clear AS A RESULT OF Group A control rods reaching 6 steps. (0.5)

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

Zt__EBQGEQUBE1_ _NQBd6Lt_6BNQBd8Lt_EMEBGENQ1_8ND PAGE 10 88Q10LQQ1G8L_GQNIBQL QUESTION 7.06 r,1.00)

Which of the following operator actions is NOT among the immediate actions of E0P 0.0, Reactor Trip or Safety Injection?

a. Check if_ steam generators are not faulted.
b. Check if main steamlines should be isolated.
c. Check RCS average temperature.
d. Verify containment spray not required.
o. Verify ECCS flow.

QUESTION 7.07 (2.00)

c. SOP-108A states that an RCP should not be starte'd unless a steam bubble exists in the pressurizer. What is the basis for this precaution?

(1.5)

b. When can RCPs be started when the pressurizer is full? (0.5)

QUESTION 7.08 (2.002

c. E0P 0.0 lists two conditions that together require the operator to trip all RCPs. What are these conditions? (1.0)
b. What are the bases for tripping the RCPs under these conditions? (1.0)

(***** END OF CATECORY 07 *****)

At__6DDIN1118811VE_EBQQEDVBEft_QQNDII1QNit_6NQ_ lid 116110NS PAGE 11 QUESTION 8.01 (2.00)

The concentration of the boric acid solution in the Boric Acid Storage System must be verified once a week in sccordance with Technical Specification 4.1.2.5. The chemist sampled the boron concentration on the following ochedule. (All samples taken at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />).

Mar 1 --- Mar 8 --- Mar 16 --- Mar 24 --- Mar 31-

c. Explain why surveillance tise interval requirements WERE or WERE NOT exceeded on Mar 16. (1.0)
b. Explain why surveillance time interval requirements WERE or WERE N01 exceeded on Mar 24. (1.0)

QUESTION 8.02 (1.50)

What is the TECHNICAL SPECIFICATION basis for the requirement to roduce Tavg to less than 500 degrees when specific activity limits on the RCS are exceeded?

QUESTION 8.03 (1.50)

c. During a valve alignment, it is reported that the discharge valve for No. 1 CentrifJgal Charging Pump is closed and can not be physically opened. Is No.1 Centrifugal Charging Pump OPERABLE? (0.5)
b. The Shift Supervisor orders an operability check to be performed on No.

2 Centrifugal Charging Pump but it fails to start. Does NONCOMPLIANCE with any Technical Specification exist? (0.5)

c. What should the Control Room Operator actions (required by Technical Specifications) be during the next 1 HOUR? (0.5)

QUESTION 8.04 (1.50)

Io it permissable to disable an annunciator which frequently alarms and

_ clears, other than for repair? WHY or WHY NOT7

(***** CATEGORY 08 CONTINUED ON NEXT FAGE *****)

At__8Dd1N11188I1VE_EBQGEQUBEft_GQNQlIl0 Nit _8NQ_LIMII8IIQNS PAGE 12 QUESTION 8.05 (1.50)

c. When may an Auxiliary Operator be permitted to attach Danger Tags to the main control board? (1.0)
b. When clearing non-radioactive equipment assigned to the Operations Department for maintenance, does the independent verification have to be performed by a licensed operator? (0.5)

QUESTION 8.06 ( .50)

Complete the following statement with one of the provided terms.

Primary to secondary steam generator U-tube leakage is classified as

__________ leakage.

a. Controlled
b. Pressure Boundary
c. Identified
d. Unidentified QUESTION 8.07 (1.50)
e. What shift personnel are permitted to issue a " Controlled Key"? (1.0)
b. When may High Radiation Area keys be issued (by the personnel in a.

above)? C0.5)

QUESTION 8.08 (2.50)

According to ISU-001A, Initial Fuel Load Sequence:

e. What is the MINIMUM number of persons that must be present at any location where fuel handling is taking place? (0,5) b.- List TWO conditions which require emergency boration during core alterations. (1.0)
c. List TWO conditions which require suspension of fuel loading and containment evacuation during fuel loading. (1.0)

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

. . . a. ,

Az__8Dd1N1118811VE_EBQQEQUBEft_QQNQ1Il0 Nit _6NQ_LidII6110NS PAGE 13

~

QUESTION 8.09 (2.50)

c. . Assume that it is 0300 on 2-19-86_and the reactor is presently at 45%

power.- Considering the Delta-I target band history listed below, calculate the associated Delta-I penalties. (1.5)

Date TimeCout) Time (in) Power (%) Penalty (min)

1. 2-18-86 0300 0318 85 _______
2. 2-18-86 1557 1633 65 _______
3. 2-19-86 0138 0300 45 _______
b. When may power be increased above 50%? (1.0) l W

(***** END OF CATEGORY 08 *****)

(*************-END OF EXAMINATION ***************)

St__IBEQBl_QE_NUGLE68_EQWEB_EL6NI_QEEBoI1QNt_ELUIDSt_6SD PAGE 14 1BE85001860101 ANSWERS -- COMANCHE PEAK.1 -86/09/23-WHITTEMORE, J./ KING ANSWER 5.01 (2.00)

c. Rhol = -2.5% delta-k/k Keffi = 1/C1-rho) = 1/(1-C .0253) = 0.9756 (0.5)

CR1/CR2 = (1-Keff23/(1-Keff1) 115/345 = 1/3 = (1-Keff2)/(1 .9756) (0,5)

Keff2 = 0.992, Rho 2 = -0.806% delta-k/k Reactivity added = .0806% - (-2.5%) =1.69% delta-k/k (0.5)

b. More neutron generations will be required for the neutron level to reach equilibrium. (0.5)

REFERENCE CP Fundamentals of Nuclear Reactor Physics, Chapter 8 - 51 ANSWER 5.02 (1.00) a (1.0)

REFERENCE Wostinghouse Reactor Theory ar.d Core Physics, Chapter I - 4.28 ANSWER 5.03 (2.50)

a. 1. Void coefficient (0.4 each, 0.3 for correct orde r) 2. Moderator temperature coefficient
3. Doppler power (or fuel temperature) coefficient
b. Total power coefficient becomes more negative from BOL to EOL. (1.0 3

REFERENCE CP Reactor Core Control, Chapter 3 - 42

. it__IHEQBl_QE_NWQLE88_EQWEB_EL8NI_QEEB8IIQNt_ELUIDSt_6NQ PAGE 15 IHEBdQQ1NadIQS ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING

,,c

  • h ANSWER 5.04 (3.00)
o. HIGHER-(0.50) steam dump pressure setting decrease causes RCS temperature to decrease. MTC and FTC both add positive reactivity to increase power. (0.50)
b. THE SAME (0.50) the steam dump. system will compensate for steam leak by shutting valves to maintain demanded steam generator. pressure. (0.50) (HIGHER if steam dumps initially closed)
c. . LOWER C0.50) the negative reactivity will cause power and TAVE'to decrease. Steam dumps will reduce steam _ flow to maintain a constant steam pressure. MTC and FTC will add positive reactivity to offset boration.-(0.50) (REMAIN THE SAME if steam dumps initially closed)

REFERENCE CP Thermal-Hydraulic Principles, Chapter 12 - 27 ANSWER 5.05 (2.00)

~

o. The instantaneous amount of reactivity by which the reactor would be subcritical from its present condition assuming all full-length RCCAs are fully inserted except for the single RCCA of highest reactivity worth which is assumed to be fully withdrawn. (1.0)

(Full credit if paraphrased)

b. It increases. (1.0) i 2

REFERENCE CP Reactor core Control, Chapter 7 - 13 through 23 CP Technical Specifications, Section 1.28 5

Z

+ - g --- - - -

~e- , - - - ,ynn- g-- - - - , - - - - , , . - - - -

Et__IBEQBl_QE_NUGLE68_EQWEB_EL6NI_QEEB6Il0Nt_ELVIDSt_6NQ PAGE 16 IBEBdQQIN65101 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 5.06 (1.50)

S/G heat transfer = Q = UA(Tavg - Tstm)

Q, U, and Tstm remain constant; A1(Tavg1 - Tatm) = A2(Tavg2 - Tstm) (0.5) 5)

Given: A2 = 0.9 x Al

'From Steam Tables: Tsat for 995 psia = 544 F C0.5)

A1(587 - 544) = 0.9A1(Tavg2 - 544)

Tevg2 = 591.8 F (591 to 592.5 F acceptable) (0.5)

REFERENCE CP Thermal-Hydraulic Principles, Chapter 12 - 8 Steam tables ANSWER 5.07 (1.00) c eccccccccccccccc REFERENCE Steam tables ANSWER 5.08 (2.00)

To determine ~ flow in NC:

Q = n cp delta-T => 100 = 100

  • cp
  • 60 => cp = 100 100
  • 60 cp = .0167 (1.0
0) . .

THEREFORE: 2% = m * .0167

  • 28 => m= 2%

= 4.28% (1.0

0) .0167
  • 28

it__IBEQBl_QE_NMGLE88_EQWEB_EL8NI_QEEB8110Nt_ELVIQ1t_8NQ PAGE 17 IBEBBQQ1N85101 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING REFERENCE-CP Thermal-Hydraulic Principles, Chapter 2 !-

l i

l f

l l

/

ht__EL6NI_1111Edi_DESIGNt_QQNIBQLt_8NQ_INSIBudENI6IIQN PAGE 18

ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING

-ANSWER. 6.01 (1.50)

NO. (0.5)

Channel II must energize to actuate for an unsafe condition (to avoid. inadvertent spray actuation in the event of a loss of instrument power). (1.0)

REFERENCE CP Logic 10 ANSWER 6.02 (2.00)

c. POWER
b. SOURCE, INTERMEDIATE and POWER
c. INTERMEDIATE and POWER
d. INTERMEDIATE and POWER C0.5 each)

REFERENCE

.CP SD III-1.6 through 13 i

ANSWER 6.03 (2.50)

e. - Maintain backpressure on orifices:to prevent flashing.

- Maintain RCS pressure when solid. (0.5 each)

b. SHUT (0,5)
c. Prevent (resin channeling due to) excess flow through demin resin. (0,5)
d. HCV-182 (Charging flow control valve, Back pressure regulator or RCP Labyrinth Delta-P Control Valve). (0.5)

REFERENCE

-CP SD II-2.9,19 i

i

_.__. e,. ,, _ _ . - - , _ _ _ _ , - _ _ _ _ _ ,. , , _ _ _ . , , , - , . . _ _ _ _ - , - - - - . _

c . ..

ht__ELeNI_111IEdi_DE110Nt_QQNIBQLt_8NQ_IN11BudENI811QN PAGE 19 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER- 6.04- (1.50)

e. Store a suf ficient valume of water to maintain the RCS in Hot Standby for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> while discharging steam to the atmosphere concurrent with loss of offsite power,(0.5) or 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in Hot Standby followed by a cooldown to 350 F at 50 F/ hour. (0.5)
b. Other systems which use the CST have elevated suction nozzles (which tap in above the minimum required level). (0.5)

REFERENCE CP SD VIII-8.5 Tochnical-Specifications, cection 3.7.1.3 ANSWER 6.05 (1.50) b.

c.

e. (0.5 each)

REFERENCE CP.Dwg. number El-0003 and El-0004 4

. ANSWER 6.06 (1.00) 553 C C0.5); The increased steam dump will cause Tave to decrease. At the P-12 setpoint, 553 F, positioning air will be vented and all steam dump valves will close (0.25). The six affected steam dump valves will open when decay heat increases Tave above 553 F and reclose at 553 F, thus acintaining Tave at approximately 553 F (0.25).

REFERENCE CP SD III-7.8, Logic 10 ANSWER 6.07 (1.00) b (0.5 for each correct answer)

O REFERENCE CP SD VII-4.34

ht__EL6NI_SYSIEUS_DESIGNt_QQNIBQLt_6NQ_INSIBuuENI6110N PAGE 20 ANSWERS -- COMANCHE PEAK 1- -86/09/23-WHITTEMORE, J./ KING I

. ANSWER 6.08 (2.00)

s. Undercompensation results in a higher than actual reading, (0.50) and if > 10E-10 amps will prevent the SR detectors from automatically energizing (0.50).
b. The operator must manually energize the SR detectors with the Source Range Manual Reset pushbuttons (0.50) when the operable IR channel drops below the P-6 setpoint(0.50).

REFERENCE CP SD III-1.24,35 ANSWER 6.09 (2.00)

e. 1. Rising surge tank level
2. Increasing CCW system radioactivity
3. Increasing thermal barrier heat exchanger outlet temperature (any two at 0.5 each)
b. The CCW outlet valve on the affected thermal barrier heat exchanger will shut (0.5) on high thermal barrier heat exchanger CCW outlet temperatureCO.5).

REFERENCE CP SD II-12.15, Dwg. number 2323-M1-0231 l

l

  • ^'

Zz__EBQQEQUBE1_:_NQBdekt_8BNQBdakt_EMEBQENQ1_8NQ- PAGE 21 86DIQLQQIQ8L_QQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING

. ANSWER 7.01 (2.00)

c. They have backup air accumulators and can be positioned from the control room. (0.53 (MANUAL is also acceptable)
b. To minimize RCS inventory loss (via the RCP seals). (0.5)
c. To prevent inj ection of accumulator nitrogen into the RCS. (0.5)
d. Continue depressurizing. (0.5)

REFERENCE CP ECA 0.0 ANSWER 7.02 (1.00)

c. (1.0)

REFERENCE CP EOS 0.1, Attachment 2 ANSWER 7.03 (2.00)

1. A boric acid gravity flow valve lineup must be performed. (0.5)

-2. Emergency borate valve, HV-8104, must be opened. (0.5)

-3. Boric acid flow can be increased by lowering VCT pressure (0.5) and decreased by raising VCT pressure or shutting HV-8104. (0.5)

(Also accept using RWST via LCV-112D and E.)

REFERENCE CP ABN - 105A, pp 7 and 8 ANSWER 7.04 (2.50)

c. 50 F/hr
b. 100 F/hr
c. 50 ppm
d. 320 F (260 F acceptable)
e. 50 F (>15 F acceptable) (0.50 each)
Zz__EBQQEQuBEs_:_NQBd6Lt_6BNQBd6Lt_EdEBGENQ1_6NQ PAGE 22

-86Q10LQQ1Q6L_QQNIBQL-ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING REFERENCE CP IPO 002A and 005A ANSWER 7.05 (2.50)

e. 1. four (0.50)
2. 551; 15- (0.50 each)
3. two (0.50)
b. 1."ANY R0D AT BOTTOM" 2.">0R= 2 RODS AT BOTTOM" (0.25 each)

REFERENCE CP IPO 002A ANSWER 7.06 (1.00) e.

_ REFERENCE CP E0P 0.0 ANSWER 7.07 (2.00)

c. Starting an RCP with unequal temperatures (0,5) in the RCS may result in a pressure excursion in the RCS when the colder water heats up and expands. (0.5) A steam bubble in the pr.essurizer will prevent a large pressure spike. (0.5)
b. Jogging or running RCPs for filling or venting (0.5)(to take the RCS solid)

Also accept when RCS < 160 F and all loop WR temperatures are within 20 F.

REFERENCE CP SOP-108A, IP0-001A

4

.- ,y ,

.'. .7 o o ,

Zi__EBQQEQUBE1_:_NQBd8Li_8BNQBd8Li_EMEBGENQi_8NQ PAGE 23 88DIQLQQIQ8L_QQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 7.08 (2.00)

'c. At least one Centrifugal Charging Pump or Safety Injection Pump running, (0.5) and RCS subcooling < 15 F. (0.5)

b. To prevent' continued mass loss from a break (LOCA) (0.5) which may result in prolonged core uncovery if RCPs were subsequently lost. (0.5)

REFERENCE CP E0P 0.0 Foldout CP TAA/MCD Chapter XIII-11.10

. o .- a- ,

At__8Dd1N11188IIVE_EBQGERMBEli_GQNQ1Il0 Nit _8ND_LIdII8IIQN3 PAGE 24 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING-ANSWER ;8.01 (2.00)

e. Interval requirement not exceeded (0.5). Eight days does not exceed 1.25 times the specified interval (0.5).

(1.0)

b. Interval requirement exceeded (0.5). The last 3 consecutive intervals exceed 3.25 times the specified interval (0.5).

(1.0)

,. REFERENCE CP TS 4.0.2 ANSWER 8.02 (1.50)

Prevents a release of activity in event of a SGTR (1.0) bocause the saturation pressure for 500 degrees is less than atmospheric steam relief valve setpoint (0.5). (1.5)

REFERENCE CP TS B 3/4 4-6 ANSWER 8.03 (1.50)

e. NO (0.5)
b. .YES~(TS 3.1.2.4 and 3.5.2) (0.5)
c. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, (0.25) take action to place the plant in Hot Standby (0.25)

REFERENCE CP' Technical Specifications 3.0.3, 3.1.2.4, 3.5.2 ANSWER 8.04 (1.50)

Yos, (0.5) The Intermittent alarm can reduce the operators' awareness of other plant alarms or indications. (1.0)

-REFERENCE CP ODA-401, Disabling of Control Panel Annunciators / Instruments

_ ~, _ _ . ._ . _ .__._ _, _ _ - . _ .- ._,

x -

. vgy q. ,_

~

~. .oy*s ,

e.

At__8 QUIN 11188IIVE_PBQQEQWBEft_QQNQlI1QNit_8NQ_ Lid 1I811QN1 PAGE 25

-ANSWERS -- COMANCHE 1 PEAK 1: -86/09/23-WHITTEMORE, J./ KING 1 ANSWER 8. 0 5L (1.50)

.o. .Only.when he is under the direct supervision of a Reactor Operator.

( 1. 0) b'. Yes. -(0.5)

REFERENCE CP STA-605, Clearance and Safety Tagging l

l

ANSWER _ 8.06 ( .50)
c. (0.5)

REFERENCE CP Technical. Specifications, Definitions' l

ANSWER 8.07 (1.50) l l; e. Shift Supervisor and' Assistant ShiftLSupervisor' (0.5 each)

~b. Only dur'ing emergencies-(when not available from Radiation Protection personnel) -(0.5)

REFERENCE CP OWI-102, Operations Department Key Control l

mty ..

0',6 e At__6DMIN11186I1YE_EBQGEQUBEft_GQNQ1IIDN1t_6NQ_ lid 116IIQN3 PAGE~ 26

^ ANSWERS - LCOMANCHE PEAK'1 -86/09/23-WHITTEMORE, J./ KING ANSWER: 8.08 (2.50)

a. 2 -(0.5)
b. 1. RCS. boron concentration < 2000 ppm. (0.5)
2. Reactor critical or approaching criticality. (0.5)
c. 1. Unexpected-increase in count rate by a factor of 5 on any temporary nuclear. instrument after the initial 8 fuel. assemblies are loaded.
2. Unexpected increase in count rate by a factor of 2 on all responding nuclear ~ instruments after the initial 8 fuel assemblies are loaded.

-3. Containment. evacuation alarm actuates due to an unexpected increase in count rate by a factor of 5 on NIS source range channels.

(any 2 at 0.5 each)

REFERENCE CP ISU-001A, Initial Startup Test Manual ANSWER 8.09: (2.50)

e. 1. 18 minutes

.2. 36 minutes (54 minutes total)

3. 41' minutes (0.5 each)
b. Immediately (as long as Delta-I remains in the target band) (1.0)

Also accept: When less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of cumulative penalty deviation exists or after 1614 on 2-19-86.

REFERENCE CP TS 3.2.1 CP TS 3.10.2.7

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