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Exam Rept 50-445/OL-86-01,administered During Wks of 860922 & 29.Exam Results:Two of Three Senior Reactor Operators & Six of Seven Reactor Operators Passed Written & Operating Exams.Exam & Answer Key Encl
	ML20214N998
Person / Time
Site:	Comanche Peak
Issue date:	11/20/1986
From:	Cooley R, Whittemore J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20214N835	List: ML20214N832; ML20214N998;
References
	50-445-OL-86-01, 50-445-OL-86-1, NUDOCS 8612040042
	Download: ML20214N998 (92)
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OPERATORLICEh!SEEXAMINATIONREPORT No. 50.445/0L 86-01 3

Docket: 50-445- _ ,

f Licensee: Texas Utilities Generating Company 400 North Olive Street Lock Box 81 Dallas, Texas '

Operator License Examinations administered to the Comanche Peak Station (CPSES)

Chief Examiner: u[G)!f(

hn E. Whittemore ,

Date Approved By: .

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R. A. Cooley 0 Ddte Summary Kritten and operating examinations were administered to three Senior Reactor Operator and seven Reactor Operator; candidates at the Comanche Peak facility during the weeks of September 22 and 29, 1986. Written examinations were administered to two candidates for Instructor Certif.ication. One Senior Reactor Operator and.one Reactor Operator candidates failed the operating portion of the examination and one of the two applicants for Instructor Certification failed the written examination and)all others passed.

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CPSES OPERATOR LICENSE EXM1INAT10N REPORT Report Details

1. Examination Results Written and operating examinations were administered to three (3) Senior Reactor Operator candidates and seven (7) Reactor operator candidates.

Two (2) of the Senior Reactor Operator candidates passed all examinations with one (1) failing the the operating examination. Six (6) of the Reactor Operator candidates passed all examinations and one (1) of the candidates failed the operating examination. One (1) of the two (2) applicants for Instructor Certification failed the written examination.

2. Examiners J. E. Whittemore (Chief Examiner)

D. N. Graves S. L. McCrory T. P. Guilfoil N. Jensen I. G. Kingsley B. Picker

3. Examination Report Performance results for individual candidates are not included in this report as it will be placed in the NRC Public Document Room. This Examination Report is composed of the sections listed below.

a. EXAMINATION REVIEW COMMENT RESOLUTION

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In general, editorial comments or changes made during the examination i or during subsequent grading reviews are not addressed by this resolution section. This section reflects comments and recommended changes to examination answer keys by the licensee. . Examination key

! modifications resulting from these' comments and recommendations 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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4 Comments and Resolutions ..

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Comment: 1.02.b.2 correct answer should be "1065 + or - 100 pcm Resolution: Accepted. Key modified.

Comment: 1.08.b.3 Also accept wording alluding to maintaining acceptable power distribution limits.

Resolution: Accepted. Key modified.

Comment: 1.09.b / 5.05.b 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 the 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: 1.12.a Attached reference material shows that the answer should be "More negative."

Resolution: Accepted. Key modified.

Comment: 1.12.c Correct answer should be "less negative." This question is confusing in that it asks about FTC, however, the question does not differentiate between Doppler Temperature - only and Doppler Power - only coefficients as taught at CPSES. This confusion may have led to the answer key's mistakes and therefore maximum grader discretion is requested.

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4 Resolution: Key modified to reflect the correct answer and comment noted.

Comment: 2.01.b Question asked for " indication or annunciation." While the annunciators listed in the key were fine, credit should be given for indications such as; Decreasing or below normal spray line temperature, Decreasing or below normal surge line : temperature,'

or other valid indications.

Resolution: Accepted. Key modified Comment: 2.02.4 Pressurizer PORVs are nitrogen operated at all times, therefore, a loss of control air has no effect on them.

Resolution: Accepted. Key modified Comment: 2.03.b Steam supply to the turbine driven auxiliary feed pump is from

1 or #4 S/G.

Resolution: Accepted. Key modified.

Comment: 2.04.

Question asked for six locations where CVCS interfaces with the RCS and did not ask for the name of the subsystem.

Consideration should be given to all or a majority of the points to the location, such as: " Cold Leg #4" or " Crossover Leg #1."

Resolution: Accepted. The point allocation has been changed to more heavily weight the location and the key has been modified.

Comment: 2.05 l Question asked for major components in the flow path. The isolation valves should not be considered major. Also " charging l

flow control valve," FCV121 is not listed but HCV182, which l

creates back pressure to ensure RCP seal injection flow, is I included. Inconsistency in choosing major components should warrant maximum grader discretion.

l l Resolution: Noted

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.*TA Comment: 2.10.a Answer should also include MANUAL with setpoint of 1/2 and require 3 of 4 possible answers.

Resolution:

Comment isnot conditions, notactions accepted. Question or switch positions.specifically Will accept'M solicits , ANUAL for no credit.

Comment: 2.10.b For high head injection (CCP), should also accept "immediate" for pressure. For accumulator pressure, anything within the Technical Specification band of 603 -686 psig should be acceptable. For SI and RHR pumps, the key should allow some tolerance such as SI: 1520 = or - 50 psig and RHR: 195 + or - 20 psig.

Resolution: Accepted. key modified.

Comment: 2.12.h 4

CCW isolation valve from RHR heat exchanger throttles open upon receipt of an "S" signal.

Resolution: Accepted. Key modified i Comment: 3.01.a '

Key should allow any response between 1960 and 1829 psig.

Resolution: Accepted. Key modified i

Comment: 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.

4 Resolution: Accepted, key modified.

i Comment: 3.07.b second portion of question asks how the plant is automatically restored to acceptable conditions, but does not ask for a

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6 description of the runback itself to include rate and duration.

Therefore, the complete answer would be "the turbine control system decreases turbine load by running the turbine back."

Additional information supplied should be acceptable but not required.

Resolution: Accepted. Key modified.

Comment: 3.09 Question implies that there are only 7 plant parameters that can be monitored at the Hot Shutdown Panel. This is not the case as documented in IP0008A.

Resolution: Accepted. Key modified to include all parameters.

Comment: 3.10.a Table RPS4, under containment ventilation isolation does not 1

show that a high rad level on the plant vent stack monitors will generate a containment ventilation isolation. However on page 1I110.30, 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.

I Resolution: Accepted. Key modified.

Comment: 3.11 737 3 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.

Resolution: This comment is ignored as there is no reason the key should

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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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,i Comment: 4.01

+ No clarification was given for total quarterly dose. It is uncertain if the 1100 mrem given as the quarterly dose was i

a ,

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inclusive of the 200 mrem weekly dose. Answers in parts a. and 4

b. should accept answers based on 1100 and 1300 mrem quarterly 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: 4.03 1

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: 4.05.c This question posed 2 problems to the candidates and should be

! reviewed accordingly. First, "if a steam generator tube rupture is discovered" is not pertinent to the question. Using this 7

phrase in the question addresses 2 unrelated problems, which caused some confusion and mislead some candidates. Secondly CPSES does not require memorization of procedures or cautions other than immediate action steps. Required memorization of this procedure is particularly inappropriate because both the statement and reason are given in the caution.

Resolution: The first concern is noted. The 4econd concern is responded to i by stating that the candidates were not asked to give the caution. The candidates were asked to provide the reason for

! the caution in accordance with NUREG 1021, ES202, paragraph l

B.4. which states, "The candidate is not expected to have normal

procedures committed to memory, but should be able to explain

reasons, cautions, and limitations of normal operating procedures.

Comment: 4.12.b The question reference states the purpose for the entire

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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.

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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: 5.06.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.

Comment: 6.02.b CPSES does not require operators to memorize detailed procedures other than immediate actions in E0P's and initial action in ABN's. 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: 6.03.d Pressurizer PORVs are nitrogen operated and consequently would not be affected by a loss of instrument air I Resolution: Not accepted. While the comment is true, it does not diminish the applicability of the question. The candidates must determine first if instrument air supplies the component, and second how the component responds to a loss of instrument air.

Several possible answers are provided in the body of the

question, one of these being, " Remain functional."

Comment: 6.06.b r

Span of level transmitter for the CST is 0 - 44'. Control board indication is 0 - 100%. The setpoint for the CST LO-L0 Level i

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9 annunciator is 4' which would equate to approx.10%, which is also the value referenced in the ERG's for switchover to SSW.

Therefore, 10% should be allowed as a correct response.

Resolution: Accepted. Key modified.

Comment: 6.08 ,-

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 th'e 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: 6.09.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: 6.12.b Failure high of N16 detector would cause the loop Tavg to become the auctioneered high Tavg. With rods in automatic they would drive in to attempt to correct a Tavg/ Tref deviation. The correct answer should be decrease.

Resolution: Accept. Key modified.

Comment: 6.14 Question asks why the RCS stabilizes at at 553 deg's. F., yet answer describes how steam dumps operate to maintain temperature. Question alludes to the fact that the steam dump valves have failed open and no interlocks work. Grader discretion is requested based on examinee assumptions.

> 0 8 10 Resolution: Not accepted. The Lo-Lo Tavg interlock affects 3 way solenoid

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valves in such a manner as to remove control air from all steam' dump valves (at 553 F. decreasing) which causes',them to' shut.-

1 The steam pressure controller affects a different 3 way' solenoid ,

4 valve and if the controller fails, it will not affect the Lo-Lo j Tavg protection, m

Comment: 7.01.a .' .

Answer "a" reference (CP TAA/MCD Chapter XIII13.8,9) on pg. 10, lists order utilized in Rev. 0 ERG's Rev. 1 ERG's have superseded Rev. O and prioritize Critical Safety Functions as -

shown in highlighted portion of reference document included.

Answer should be changed to 2,3,5,1,6,4.

Resolution: Accept. Key modified.

I Comment: 7.03.a

! Answer "a" addresses the situations or conditions outlined in reference CP SDII3.25, 26. However, procedural guidance is i also provided for conditions which require " Emergency Boration"

i. of the RCS in EOS 0.1, step 1. Thus, the answer key should be expanded to include all situations or. conditions inclusive of i both sources.

f Resolution: Accepted. Key modified.

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Comment: 7.04 l 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.

1 I Comment: 7.06 i IFIES does not require operators to memorize detailed procedure f steps other than immediate actions in E0Ps or initial actions I in ABNs. This question asks for information contained in subsequent operator actions which would be done af ter consulting the procedure. Although the answer references ABN105A, the question neither references ABN-105A nor mandates utilization of

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l 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

modified to accept the RWST via LCV1120 and E flowpath as an i alternative answer.

Comment: 7.07.e

The question, although prefaced by general usage of IP0s, is

answered specifically from IP0-002A. The condition required by IP0-002A (50 deg's subcooling) is not applicable throughout the IP0s. Foe 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.

i Resolution: Accept. Key modified i Comment: 7.09

ThTanswer per reference given (CP IP0 002A) only list 4 of the 6 steps required to be performed for continuation of startup.

The an:wer should be expanded to include all 6 steps required and require 4 of 6 for full credit.

Resolution: Accept. Key modified Comment: 7.12.a Although the question addresses the basis for precaution 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 questi.on 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 i.s directly related to the intent of the precaution. .

l' 12 Comment- 8.03.b Answer should also accept Shift Supervisor in accordance with Technical Specification 6.2.2.e.

Resolution: Accepted. Key modified Comment: 8.03.c Normally the Fire Brigade Leader is the Assistant Shift Supervisor. However, FIR 104, Section 4.3.11 also permits a-fully qualified A0 to fulfill the Brigade Leader position.

Resolution: Not accepted. The question specifically asks, "Who is normally responsible to function as the Fire Brigade Leader?" Therefore, the only acceptable answer is " Assistant Shift Supervisor."

Comment: 8.05 1TEEi 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.

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 to discuss the site visit and conduct of the examinations. The following personnel were present for this exit 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 I e i

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

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

(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) SR0s 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.

U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _QQM6NQUE_PE6K_1_________

REACTOR TYPE: _EWB-WEQ4________________

DATE ADMINISTERED: _Q64Q2423________________

EXAMINER: _WBlIIEdQ8Et_JzfEINQ5____

CANDIDATE: _________________________

INSIBUQIl0S5_IQ_QaNQ1QeIE1 Use separate paper for the answers. Write answers on one side only.

Staple question sheet on-top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at leest 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY

__V8LUE_ _IQI6L ___500BE___ _VeLUE__ ______________QeIEQQBY

_251QQ__ _251s0 ___________ ________ S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_25tDQ__ _25tQQ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_25tQQ__ _25tQQ ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

_25tDQ__ _25100 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS lQQtQQ__ ___________ Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

St__IMEdBl_Qd_NUGLEeB_EQWEB_ELeNI_QEEBoIIQNt_ELu1QS&_eNQ .PAGE 2 IMEBdQQ1Ned1GS QUESTION 5.01 (2.00)

n. 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 (1.00)

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 statements is CORRECT?

a. 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 (1.00)

TRUE or FALSE?

a. As Keff approaches unity, a smaller change in neutron level will result for identical changes in Keff. (0.5)

b. With Keff greater than unity, a constant positive startup rate with increasing neutron level will occur only if net REACTIVITY is NOT changing. (0.5)

QUESTION 5.04 (2.50)

e. List the three most significant contributors to total power coefficient in order of INCREASING n.agnitude. (1.5)

b. How does total power coefficient vary as the core ages? (1.0)

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

51__IMEdBl_Q8_NuGLE88_EQWEB_EL8NI_QEEB8110Nt_ELu1QSt_8NQ PAGE 3 IBEBdQQ1NedIGH QUESTION 5.05 (2.00)

a. List TWO reasons why critical boron concentration decreases over core life. (1.0)

b. How does differential boron worth vary over core life? (1.0)

QUESTION 5.06 (3.00)

For the following situations, indicate whether the final stable power level will be HIGHER, LOWER, or THE SAME as the initial power level. EXPLAIN your answers. Assume the initial power

level is at approximately 10 % following a normal reactor

startup at the end of life. Consider each situation separately,

s. 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) i 1

l

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

5t__IBEQBl_QE_NuGLE88_EQWEB_EL8HI_QEEB8Il0Nt_ELu1Dit_8ND PAGE 4 IHEBdQQ1Ned1GS QUESTION 5.07 (1.50)

Compare the calculated Estimated Critical Rod Position (ECP) for a startup 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months after a trip to the Actual Critical Rod Position (ACP) if the following events / conditions occurred. Consider each independently. Limit your answer to:

a. ACP higher than ECP.

b. ACP lower than ECP.

c. ACP would not be significantly different than ECP.

1. One Reactor Coolant Pump is stopped one minute prior to criticality.

(0.5)

2. The steam dump pressure setpoint is increased to a value j ust below the code safeties setpoints. (0.5)

3. The startup is delayed 2 more hours. (0,5)

QUESTION 5.08 (2.00)

e. 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 effected PRIOR to a reactor trip? (1.0)

QUESTION 5.09 (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 *****)

Hi__IBEQBl_QE_NuGLE88_EQWEB_EL8HI_QEE86Il0Nt_ELu1QHt_86Q PAGE 5 IBEBdQQ188U1GS l.

. QUESTION 5.10 (1.00)

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

e. 405 psig

b. 325 psig

c. 245 psig

d. 165 psig i QUESTION 5.11 (2.00) i

, Will the Departure from Nuclear Boiling Ratio (DNBR) INCREASE, l DECREASE, or REMAIN THE SAME if the following plant parameters INCREASE during power operation? Consider each parameter independently.

c. Reactor Coolant System (RCS) Pressure (0.5)

b. RCS Temperature (0.5)

c. RCS Flow (0.5)

d. Reactor Power (0.5)

QUESTION 5.12 (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 occurs. Natural circulation is established and core delta-T goes to 28 F.

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

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

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-St__IBEQBl_QE_NUGLE68_EQWE8_EL6NI_QEEB8I1QUt_EL91QSt_6NQ PAGE 6 IBE800Q1Ned1GS QUESTION 5.13 (2.00)

Indicate whether the following situations result in SUBCOOLED, SATURATED, or SUPERHEATED fluid conditions.

a. Pressurizer PORV relieving to the PRT (0.5)

b. Steam generator saf ety valve relieving to atmosphere (0.5)

c. Steam from a Moisture Separator Reheater entering a low pressure turbine (0.5)

d. Condensate exiting the condenser hotwell (0.5)

QUESTION 5.14 (1.50)

It is observed that reactor coolant flow through a steam generator is approximately 10 TIMES the feedwater flow into the same steam generator.

However, the feedwater delta-T in that steam generator is only twice the RCS delta-T. What accounts for this apparent heat transfer mismatch?

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

ki__EL86I_SYAIEdS_DESIGNi_CQNIBQLi_8NQ_INSIBudENI8I1QN PAGE 7 QUESTION 6.01 (1.50) i The plant is operating at 100 % steady state power with containment pressure channel IV (PB 934A) failed high. A tochnician troubleshooting the trip bistables inadvertently de-energizes the instrument power for containment pressure channel II. Will a Containment Spray Actuation occur? (0.5)

WHY or WHY NOT? (1.0)

QUESTION 6.02 (2.00)

a. 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)

QUESTION 6.03 (2.50)

State how the following components respond (FAIL OPEN, FAIL CLOSED, REMAIN FUNCTIONAL, DIVERTS TO ..., ETC.) when instrument air pressure is lost with the plant at 100% power.

f a. Letdown pressure control valve (PCV-131) (0.5) l'

b. Volume control tank level control valve (LCV-112A) (0.5)

c. Steam generator atmospheric relief valves (0.5)

d. Pressurizer PORVs (0.5)

e. Auxiliary feedwater regulating control valves (0,5) 3

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

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

6t__EL8MI_1111Ed1_QE11GNt_CQNIBQLt_8NQ_IN11895ENIeI1QN PAGE 8 QUESTION 6.04 (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.

a. Provides a direct input to the Rod Control System,

b. Has a reactor trip function that is blocked at some time between startup and full. power 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".

QUESTION 6.05 (2.50)

The following concern the CVCS.

a. 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 a safety injection initiation? (0.5)

c. Why is letdown flow limited to 120 gpm? (0.5) t

d. With only the positive displacement pump operating at power, which valve (s) is/are utilized to control RCP seal inj ection flow? (Noun l name(s) or number (s) is/are acceptable.) (0.5)

\

l l

QUESTION 6.06 (2.50)

n. List FOUR signals (non-similar/ unique) which will initiate j e motor driven Auxiliary Feedwater Actuation Signal (AFAS). (2.0) i i b. With an AFAS signal initiated, HOW and WHEN is the Auxiliary Feed pump f water supply shifted from the Condensate Storage Tank to the Service

[ Water System? (0.5) l l (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

[

ht__EL881_SYSIEUS_DESIGNt_CQNIBQLt_aND_INSIBudENI611QN- PAGE 9 QUESTION 6.07 (1.60)

Match the following symptoms or causes in column "B" to the specific Rod Control System failure or error in column "A".

"A" "B"

a. Logic Cabinet Urgent Failure 1. Caused by simultaneous zero current order to stationary and movable grippers.

I b. Regulation failure 2. Unselected rod (s) having current

-l flow in movable or lift coils,

c. Phase failure 3. Caused by failure of redundant power supply module.

1

d. Logic error 4. Caused by pulser or slave cycler failure.

5. Caused by full current being i

applied for excessive time.

1 (There is only 1 correct numerical 6. Occurs when voltage to coils has answer for each lettered error or excessive ripple.

! failure at 0.4 each) i t

f QUESTION 6.08 (1.50)

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

(1.0)

{ b. How is the minimum CST water volume required by Technical 4

Specifications ensured to be available at all times considering that

several systems take suction on the CST? (0.5)

QUESTION 6.09 (2.00)

)

a. 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) f

b. 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) i i

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

. . )

ht__EL881_113IEdi_QE11GNt_QQNIBQLt_8NQ_INSIBudENI6110N PAGE 10 QUESTION 6.10 (1.00)

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

a. Excessive vibration

b. Generator differential

c. Generator reverse power j

d. Low lube oil pressure

e. Overspeed

f. - High jacket water temperature QUESTION 6.11 (1.50)

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

a. Centrifugal charging pump #12 b.- Reactor coolant pump #14

c. Condensate pump #12

d. Containment spray pump #14

e. Turbine building cooling water pump #12

f. Service air compressor #11 i

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

6___P(6$1_SlklEdS_QESlqN_CQNIBQL_ANQ_lNSIBUMEN1611QN t t PAGE 11 QUESTION 6.12 (2.00)

The plant is operating at 100% power with all control systems in automatic.

Bank D rods are at 200 steps. Given the following conditions / situations, how will rod height be affected (INCREASE, DECREASE, NO CHANGE)? Assume no operator action and consider each case separately. Assume the reactor does NOT trip.

a. A safety valve on 8 steam generator fails open. (0,5) b.- Loop 1 N-16 power detector fails high. (0.5)

c. C loop narrow-range Tcold instrument fails low. (0.5)

d. Turbine load is reduced to 80%. (0.5)

QUESTION 6.13 (1.40)

The reactor is critical at the point of adding heat. List SEVEN reactor trips which are DISABLED in this condition.

QUESTION 6.14 (1.00)

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

r i

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

J

~Zi__EBQ6EQUBES_=_NQBd6Lt_6BNQBdekt_EMEBGENGl_8ND PAGE 12 88DIQLQQ1G8L_GQNIBQL

QUESTION 7.01 (2.00)

e. Number the Critical Safety Functions in order of priority. (1 -

highest, 6- lowest) (1.2)

Core Cooling _______

Core Heat. Sink _______

Containment Integrity _____.,_

S ub c r it ic ality_______

j RCS Inventory _______

RCS Integrity _______.

l

b. Which of the following CSF conditions take precedence in requiring operator response? (0.8)

1. RCS Inventory - Orange Path j

2. Core Heat Sink - Yellow Path

! 3. Containment Integrity - Red Path

4. Subcriticality - Orange Path l S. RCS Integrity - Red Path QUESTION 7.02 (2.00)

a. 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) a QUESTION 7.03 (3.00) i n. List THREE situations or conditions (unrelated) which require the

! operator to commence EMERGENCY BORATION of the RCS via HV-8104.

j (1.5) j b. Briefly describe the " normal" method of emergency boration. (operator

, actions only) (1.5) t i

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

Zt__EBQdEQUBEH_:_NQBdekt_8BNQBd8Lt_EMEBGENQ1_6NQ PAGE 13 86DIQLQGIQ6L_GQNIBQL t

QUESTION 7.04 (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)

c. 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? (0.5)

QUESTION 7.05 (1.00)

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

1 RCS SG CORE EXIT SUBC00 LING PRESSURES Thot Tcold THERMOCOUPLES

e. 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 i

e. 10 F Constant Constant Constant Decreasing e

QUESTION 7.06 (2.00) i In the event that both boric acid transfer pumps are inoperable, how can

RCS boration be accomplished (lineup) and controlled (flow increased and ,

decreased) until the boric acid pumps are repaired?

{

l 4

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

'Zi__EBQdEQUB$1_:_NQBdekt_8BNQBd8Lt_EMEBGENQ1_8NQ , PAGE 14 88DIQLQQIQ8L_QQNIBQL QUESTION 7.07 (2.50)

Supply the following.which must be observed during plant operation in accordance with INTEGRATED OPERATING PROCEDURES.

c. Maximum RCS cooldown rate , _ (0.50)

b. Maximum pressurizer cooldown rate (0.50) l

c. Maximum boron concentration differential between RCS and pressurizer (0.50)

d. Maximum differential temperature between pressurizer and spray fluid (0.50)

e. Minimum RCS subcooling (0.50)

QUESTION 7.08 (2.50)

e. Complete the following statements regarding a reactor s'tartup.

1. The reactor must be critical no greater than ________(0.50) hours following the Estimated Critical Concentration calculation.

2. RCS Tave must be verified to be greater than ________(0.50) F every

_______(0.50) minutes until criticelity is reached.

! 3. Source range counts must read at least ________(0.50) cps on the j highest channel prior to rod withdrawal.

l

b. List TWO alarms which are expected to clear AS A RESULT OF Group A control rods reaching 6 steps. (0.5) l l QUESTION 7.09 (2.00)

List FOUR actions required to continue's reactor startup if the reactor is still subcritical when Group D control rods reach 228 steps.

(

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

Zi__EBQdEDMBIS_:_NQBdokt_eRUQBdekt_EMEBQENQ1_6NQ PAGE 15 86DIQLQQ1Q6L_QQNIBQL QUESTION 7.10 (1.00)

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

i

s. 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 7.11 ( T.'0 0 )

a. How and where is the RCS monitored for fuel element failures? (1.0)

b. What are the Technical Specification limits for RCS specific activity?

(1.0)

c. What plant system change (s) is/are required in response to a verified fuel element failure with RCS specific activity below Technical Specification limits? (1.0)

QUESTION 7.12 (2.00)

n. SOP-108A states that an RCP should not be started 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)

(***** END OF CATEGORY 07 *****)

I i

Sz_ 8DM NISIB811YE_EBQQEQVBESt_GQNDIIlQNSt_8ND_LidlI611QUS PAGE 16 QUESTION 8.01 (2.00h ,

The concentratio'n of the boric acid solution in the Boric Acid Storage Syst'em must be verified,once a week in cccordance with Technical Specificat. ion 4.1.2.5. The chemist sampled the boron concentration on the following ochedule. (All samples taken at 1,200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months ).

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

c. Explain why surveillance time interval requirements WERE or WERE NOT exceeded on Mar 16. (1.0) i

b. Explain why surveillance time interval requirements, '

WERE or WERE NOT exceeded on Mar 24. , (1.0)

)

i QUESTION 8.02 (1.50)

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

QUESTION 8.03 (2.00)

a. How many members per shift are required on the fire brigade per Technical Specifications? [0.5)

b. Who may NOT be included as members of the , fire brigade? [1.0)

c. Who is normally responsible to. function a's the fire brigade leader?

(0.5)

QUESTION 8.04 (3.00) ,

In acccordance with STA-205," Temporary Changes to Procedures", temp'orary changes may be made to plant procedures governed by Technical ,,

Specifications if four criteria are met. Provide these FOUR criteria.

N

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

At__eQUINISI5611YE_E80CEQUBESt_GQUQ1IIQNSt_6NQ_LIdII6IIQNS PAGE 17 QUESTION 8.05 (2.50) i s. 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)

QUESTION 3.06 (1.50)

a. During a valve alignment, it is reported that the discharge valve for No. 1 Centrifugal Charging Pump is closed and can not be physically opened. Is No.1 Centrifugal Charging Pump OPERABLE? (1,0)

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? (1.0)

c. What should the Control Room Operator actions (required by Technical Specifications) be during the next 1 HOUR 7 (1.0)

QUESTION 8.07 (1.50)

Is it permissable to disable an annunciator which frequently alarms and clears, other than for repair? WHY or WHY NOT?

QUESTION 8.08 (3.00)

List SIX conditions or occurrences that require notification of the NRC within ONE hour.

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

St__8Dd1NISIB8IIVE_ESQGEQUBESt_CQNQ1110Nat_8NQ_ lid 1I811QNS PAGE 18 4

4 QUESTION 8.09 (2.50) ;

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

i

a. What is the MINIMUM number of persons that must be present at any location where fuel handling is taking place? (0.5) l

b. List TWO conditions which require emergency boration during core alterations. (1.0) i i

c. L ist TWO conditions which require suspension of fuel loading and containment evacuation during fuel loading. (1.0)

QUESTION 8.10 (1.50) i

s. 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)

J QUESTION 8.11 ( .50)

Complete the following statement with one of the provided terms.

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

a. Controlled

b. Pressure Boundary

.c. Identified

d. Unidentified QUESTION 8.12 (1.50)

a. 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)? (0.5) l

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

r Rz__605fNISIB611YE_EBQQEQUBEft_CQNQ1Il0NSt_8UQ_L151I611QU$ PAGE 19 QUESTION 8.13 (2.00)

a. In accordance with EPP-201, WHO IS RESPONSIBLE to assess abnormal or emergency conditions to determine whether or not an emergency classification is warranted? C0,5)

1. Emergency Coordinator

2. Operations Superintendant

3. Plant Manager

4. Shift Supervisor

5. Resident NRC Inspector

b. Who assumes immediate responsibility as the Emergency Coordinator upon activation of the Emergency Plan? C0.5)

1. Shift Supervisor

2. Resident NRC Inspector

3. Person discovering condition

4. Plant Manager

5. Operations Superintendant

c. Lint the EMERGENCY ACTION LEVELS in order of severity (1 = most severe). (1.0)

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

! C_*******$*****_END OF EXAMINATION

- - - - )

51__IBEQBl_QE_NVQLEeB_EQWEB_ELeNI_QEEBoIIQNt_ELVIDSt_6NQ PAGE 20 IBEBdQQ1NedICS ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING i

ANSWER 5.01 (2.00)

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

CR1/CR2 = (1-Keff23/(1-Keff1) 115/345 = 1/3 n (1-Keff2)/(1 .9756) (0.5)

Keff2 = 0.992, Rho 2 = -0.806% delta-k/k Reactivity added = .0806% - C-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 al ANSWER 5.02 (1.00) a (1.0) ,

REFERENCE Westinghouse Reactor Theory and Core Physics, Chapter I - 4.28 ANSWER 5.03 (1.00)

s. False (0.5)

b. True (0.5)

REFERENCE CP Fundamentals of Nuclear Reactor Physics, Chapter 8 - 54,60 ANSWER 5.04 (2.50)

a. 1. Void coefficient (0.4 each, 0.3 for correct order)

2. Moderator temperature coefficient

3. Doppler power (or fuel temperature) coefficient

b. Total power coefficient becomes more negative from BOL to EOL. (1.0)

St__IMEdBY_QE_UUQLE6B_EQWEB_EL6BI_QEEB8Il0Nt_ELUIQ1&_8ND PAGE 21 IBEBUQQ1Ned191 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING REFERENCE CP Reactor Core Control, Chapter 3 - 42 ANSWER 5.05 (2.00)

e. 1. Fission poison buildup

2. Fuel depletion (0.5 each)

b. Differential boron worth becomes less negative from BOL to EOL. (1.0)

(ASSUMES CONSTANT (B])

If (B) is assumed to decrease with core age, then OBW becomes more negative initially and then less negative at EOL.

REFERENCE CP Reactor Core Control, Chapter 3- 26 and 5 - 14 ANSWER 5.06 (3.00)

a. 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 (0.50) 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 5.07 (1.50)

1. c (same)

2. a (ACP higher)

3. b (ACP lower) (0.5 each)

Si__IBEQ'BY_Qb_NMQLE6B_EQWEB_EL6NI_QEEB6Il0Nt_ELU1QSt_6NQ PAGE 22 IHEBdQQYNad101 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING REFERENCE CP Reactor Core Control, Chapter 3 - 4,5; Chapter 4 - 20,21 ANSWER 5.08 (2.00)

c. The instantaneous amount of reactivity by which the reactor would be subcritical from its present condition assutning 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) j (Full credit if paraphrased)

b. It increases. (1.0)

REFERENCE CP Reactor core Control, Chapter 7 - 13 through 23 CP Technical Specifications, Section 1.28 ANSWER 5.09 (1.50)

S/G heat transfer = Q = UACTavg - Tstm)

Q, U, and Tstm remain constant; 1

A1(Tavg1 - Tstm) = A2(Tavg2 - Tstm) (0.5)

Given: A2 = 0.9 x Al From Steam Tables: Tsat for 995 psia = 544 F (0.5)

A1C587 - 544) = 0.9A1(Tavg2 - 544)

Tavg2 = 591.8 F (591 to 592.5 F acceptable) (0.5) l REFERENCE CP Thermal-Hydraulic Principles, Chapter 12 - 8 I Steam tables l

ANSWER 5.10 (1.00) l

, C l

l I

S&__IBEdBl_Q$_NVQLEeB_EQWEB_ELeUI_QEE86110Nt_ELVIDSt_6NQ PAGE 23 IBEBdQQ1Ned103 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING REFERENCE Steam tables ANSWER 5.11 (2.00)

a. Increase

b. Decrease

c. Increase

d. Decrease (0.5 each)

REFERENCE CP Technical Specifications, Figure 2.1-1, B2-1 CP Thermal-Hydraulic Principles, Chapter 13 - 23,24 ANSWER 5.12 (2.00)

To determine flow in NC:

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

cp

60 => cp = 100 100

60 cp = .0167 (1.00)

THEREFORE: 2% = m * .0167

28 => m= 2%

= 4.28% (1.00)

.0167

28 REFERENCE CP Thermal-Hydraulic Principles, Chapter 2- 39 ANSWER 5.13 (2.00)

a. Saturated B. Superheated

c. Superheated

d. Subcooled (0.5 each)

REFERENCE Steam Tables l

51__IBEQBY_Q$_NVQLE88_EQWEB_EL8NI_QEEB8IIQut_ELUIQSt_88Q PAGE 24 IHEBMQQ1N8MIQS ANSWERS -- COMANCHE PEAK 1 - 86/09/23-WHITTEMORE, J./ KING ANSWER 5.14 (1.50)

The delta-T of the feedwater is not a true indication of heat transfer through the U-tubes.(0.5 for concept) Most of the heat added to a steam greerator provides the heat of vaporization necessary to change the foedwater into steam at a constant temperature,(0.5) and therfore is not eccounted for in the temperature rise of the feedwater.(0,5)

REFERENCE Steam Tables 4

)

., .. . , . - -- , . - - . , - - - - . , . . -n--,____, ,n. - , . . - .

'ht__PL8$1_31hlgd3_QEllGNt_GQNIBQLt_8NQ_lN11EydEN1811QN PAGE 25 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 6.01 (1.50)

NO. (0.5)

Channel II.must energize to ..ctuate for en 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)

n. 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 111-1.24,35 ANSWER 6.03 (2.50)

a. Fail open

b. Goes to VCT

c. Remain functional, closed (provided with backup air accumulators)

d. Remain functional, closed (provided with backup nitrogen acci ulators)

e. Remain functional, open (provided with backup air accumulators)

(0.5 each)

REFERENCE CP SD 11-2.9,13, IX-1.5, Table PG-1

6t__ELedI_S151 EMS _QESIGNt_QQNIBQLt_6ND_INSIBUMENI6IIQN PAGE 26 4

ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 6.04 (2.00)

n. POWER

b. SOURCE, INTERMEDIATE and POWER

c. INTERMEDIATE and POWER

d. INTERMEDIATE and POWER (0.5 each)

REFERENCE CP SD III-1.6 through 13

' ANSWER 6.05 (2.50)

a. - 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 ANSWER 6.06 (2.50)

's. 1. Manual

2. S/G low-low level on 1 steam generator

3. Safety Inj ection sequence signal

4. Trip of both Main Feed Pumps

5. Blackout sequence signal (Any 4 at 0.5 each)

b. Manually by the operator (0.25)at a CST level of 4 feet (or " CST LVL LO-LO" alarm or 10 percent) (0.25)

REFERENCE

. CP SD VIII-8.6,10 CP ALB-88, 1.6 l

l l

l

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

61__EL6N'I_1151EdS_DE11GNt_QQNIBQLt_68Q_INSIBudENI6IIQN PAGE 27 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 6.07 (1.60)

c. 4

b. 5

c. 6 (0.4 each)

d. 1 REFERENCE CP SD III-3.23 through 25 ANSWER 6.08 (1.50)

a. Store a sufficient volume of water to maintain the RCS in Hot Standby for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months while discharging steam to the atmosphere concurrent with loss of offsite power,(0.5) or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 Technical Specifications, section 3.7.1.3 ANSWER 6.09 (2.00)

a. 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 temperature (0.5).

REFERENCE CP SD II-12.15, Owg. number 2323-M1-0231 1

r

- , . , - - - - - - - - t -,- ,r ,- my ,-.; - - - - . - - - . , . - . , -

Ez__ELeNI_SI51EUS_DE1196t_QQUIBQLt_6NQ_ldSIBQUENI611QN PAGE 28 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 6.10 (1.00) b (0.5 for each correct answer) e REFERENCE CP SD VII-4.34 ANSWER 6.11 (1.50) b.

c.

c. (0.5 each)

REFERENCE CP Dwg. number El-0003 and El-0004 ANSWER 6.12 (2.00)

c. INCREASE

b. DECREASE

c. NO CHANGE

d. DECREASE C0.5 each)

REFERENCE CP SD III-3.10 ANSWER 6.13 (1.40)

1. Source range high flux

2. RCP low voltage

3. RCP underfrequency

4. Pressurizer low pressure

5. Pressurizer high level

6. Loop low flow

7. Turbine trip (0.2 each) l REFERENCE l CP SD III-9, Table RPS-1 i

ht__EL8NI_S15IEUS_ DES 10Nt_99 BIB 9Lt_aNQ_INSIBQUENI6IIQN PAGE 29 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 6.14 (1.00) 553 F (0.5); 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 velves will close (0.25). The six effected steam dump valves will open when decay heat increases Tave above 553 F and reclose at 553 F, thus maintaining Teve at approximately 553 F (0.25).

REFERENCE CP SD III-7.8, Logic 10 l

a Zi__EBQQEQuB51_:_NQBd8Lt_8HNQBd8Lt_EMEBQENQ1_8eD PAGE 30 i

88Q1QLQQ108L_GQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING I

4 ANSWER 7.01 -(2.00) o.- 2, 3, 5, 1, 6, 4 (0.2 each)

b. 5 (0.8)

REFERENCE

! CP TAA/MCD Chapter XIII - 13.8, 9 I

1 ANSWER 7.02 (2.00)

a. At least one Centrifugal Charging Pump or Safety Inj ection 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

+

i l ' ANSWER 7.03 (3.00) j o. 1. Failure of more than one RCCA to fully insert following a reactor trip.

! 2. CRH below RIL.

I 3. Uncontrolled RCS cooldown following reactor trip or shutdown not I

requiring ECCS.

4. Unexplained or uncontrolled reactivity increase.

(any three at 0.5 each) l 1.

b. Open emergency boration control valve (HV-8104)

2. Start boric transfer pump

3. Verify boration by observing boration flow rate or response of plant to addition of negative reactivity.

(0.5 each)

Append to a. above:

5. SDM less than required.

6. Boron concentration than required during refueling.

7. ATWT.

t- , _-_.-_., . .... .,_.-,. - . _ . . _ . . . _ _ - . _ - . - . . _ . _ _ . . - . - - - _ - . -

Zt__EBQQ'EQUB$1_:_NQBd8Lt_6BNQBd6Lt_EMEBGENGY_eND PAGE 31 bod 10LQQ196L_GQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ,

REFERENCE CP SD II - 3.25, 26 ANSWER 7.04 (2.00)

a. They have backup air accumulators and can be positioned from the control room. (0.5) (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 7.05 (1.00)

c. (1.0)

REFERENCE CP EOS 0.1, Attachment 2 ANSWER 7.06 (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. (0.53 (Also accept using RWST via LCV-1120 and E.)

REFERENCE CP ABN - 105A, pp 7 and 8

e .

l Zz__EBQG'EQUB$S_:_NDBd6Lt_6BNQBdekt_EMEBGENGX_6ND PAGE 32 bed 10LQG108L_GQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 7.07 (2.50)

e. 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)

REFERENCE CP IPO 002A and 005A ANSWER 7.08 (2.50)

a. 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 7.09 (2.00) j

1. Insert Group D to O steps

2. Contact Engineering Support and recalculate the ECC for Group D at 100 steps.

3. Dilute RCS as necessary.

4. Obtain permission from SS to perform startup.

5. Initiate 1/M plot.

6. Pull bank 0 as needed. (any four at 0.5 each)

REFERENCE CP IPO-002A-

{

ANSWER 7.10 (1.00) e.

i

Zt__EBQQiQuBf1_:_UQBd6Lt_8HNQBd8Lt_EMEBGENQ1_8NQ PAGE 33 2

88QIQLQQ1Q8L_QQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING f

REFERENCE y CP EOP 0.0 a

ANSWER 7.11 (3.00) i

a. A gross failed fuel monitor (RE - 406) (0.5) . measures fission product activity (downstream of PCV - 131) in the CVCS. (0.5)

b. Less than or equal to 1 microcurie per gram dose equivalent I-131 (0.5) and less than or equal to 100/E microcuries per gram of gross radioactivity. (0.5) -

c. Increase CVCS letdown flow to 120 gpm (0.75) and ensure charging flow j increases. (0.25)

. REFERENCE CP SD Chapter II-2.9 j, ABN-102A CP~ Technical Specifications i

I ANSWER 7.12 (2.00)

a. 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 pressurizer will prevent a large j . pressure spike. (0.5)

b. Jogging or running RCPs for filling or venting (0,5)(to take the RCS solid) s Also accept when RCS < 160 F and all loop WR temperatures are within 20 F.

i

, REFERENCE CP SOP-108A, IPO-001A l

i J

At__8Dd1'NISI58IIVE_EBQQEQUBESt_QQUQ1Il0NSt_6NQ_Lld1I6I1QU$ PAGE 34 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 8.01 (2.00)

n. 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) '

because 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 (2.00)

c. 5 (0.5)

b. The fire brigade shall NOT include the SS and the two other members of the minimum shift crew required for safe S/D of Unit 1 (0.5) and any personnel required for essential functions during the fire.(0.5)

c. The Assistant Shift Supervisor is designated as the leader. (0.5)

REFERENCE CP FIR-104, Fire Brigade CP Technical Specifications, Section 6

Az__8Dd1NISI58IIVE_EBQQEDUBESt_QQND1IIQUSt_8ND_L10118I1QNS PAGE 35 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING ANSWER 8.04 (3.00)

1. The intent of the original procedure is not altered.(0.75)

2. The change is approved by two members of the plant staff, et least one of whom is a member of station management staff. If the change affects station equipment, a third approval by the SS or ASS is required.(0.75)

3. No QA requirements may be deleted.(0.75)

4. The change is documented, reviewed by SORC and approved by the Manager, Plant Operations within 14 days of implementation (only after fuel load).(0.75)

REFERENCE CP STA-205, Temporary Changes to Procedures ANSWER 8.05 (2.50)

n. 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation exists or after 1614 on 2-19-86.

REFERENCE CP TS 3.2.1 CP TS 3.10.2.7 ANSWER 8.06 (1.50)

c. NO (0.5)

b. Y2S (TS 3.1.2.4 and 3.5.2) (0.5)

c. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , (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

I Az__6Dd1NISIB611VE_EBQQERMBESt_GQNQ1110NSt_6NQ_LidII6110Ng PAGE 36 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING l

l ANSWER 8.07 (1.50)

Yes, (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 ANSWER 8.08 (3.00)

1. Declaration of any emergency classification.

2. The initiation of any plant shutdown required by Technical Specifications.

3. Any authorized deviation (10CFR50)from Technical Specifications.

4. Any event or condition during operation that results in the condition of the plant or safety barriers being seriously degraded, or results in the plant being:

a. In an unanalyzed condition that significantly compromises plant safety, or

b. In a condition that is outside the design basis of the plant, or

c. In a condition not covered by the plant's operating and emergency procedures.

5. Any natural phenomenom or external condition that poses an actual threat to safety of plant or significantly hampers site personnel while performing duties required for safe operation of the plant.

6. Any event that results or should have resulted in ECCS actuation on a valid signal.

7. Any event that results in a major loss of emergency assessment capability.

8. Any event that poses an actual threat to the safety of the plant or significantly hampers site personnel while performing duties necessary for safe operation of the plant including fires, toxic gas releases, or radioactive releases.

9. Any violation of a safety limit.

(any 6 at 0.5 each)

(No. 4 counts as 4 separate conditions)

REFERENCE CP Technical Specifications CP STA-501, Reporting of Operating Information to the NRC 10CFR50.72

1

. . i 1

I St__eQd1'UISIB'el1VE_EBQQEQUBESt_QQUQ1110NHt_aNQ_ Lid 116I10NS PAGE 37 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING l

ANSWER 8.09 (2.50)

c. 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 8.10 (1.50)

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

(1.0)

b. Yes. (0.5) i REFERENCE CP STA-605, Clearance and Safety Tagging ANSWER 8.11 ( .50)

c. (0.5)

REFERENCE CP Technical Specifications, Definitions I

At__8 Dbl'NISIb6IIVE_EBQQEQUBESt_QQNQlIl0NSt_6NQ_LidII6IlQNS PAGE 38 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./ KING 1

4 i

ANSWER 8.12 (1.50)

e. Shift Supervisor and Assistant Shift Supervisor (0.5 each)

b. Only during emergencies (when not available from Radiation Protection personnel) (0.5)

REFERENCE CP OWI-102, Operations Department Key Control i

ANSWER 8.13 (2.00)

c. 4 (0.5)

b. 1 (0.5)

c. 1. General Emergency

2. Site Area Emergency

3. Alert

4. Unusual Event (0.15 for each classification, 0.1 each for proper order)

REFERENCE CP EPP-201 l

i _ _ _ . . - _.

U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _QQU6NGBE_EE6K_1_________

REACTOR TYPE: _PWB-WEQ4________________

DATE ADMINISTERED: _QhLQQl2Q________________

EXAMINER: _WBlIIEdQBEz_JzlGUILE____

CANDIDATE: _________________________

INSIBUCIl0NS_IQ_QaNQ1DaIE1 Use . separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer . sheets. Points for each question are indicated in parentheses 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 six (6) hours after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY

__Y6LUE_ _IQIaL ___1GQBE___ _Y8LME__ ______________Q6IEQQRY_____________

_25zQQ__ _2EzQQ ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

_2EzQQ__ _251QQ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS

_25zQQ__ _25zQQ ___________ ________ 3. INSTRUMENTS AND CONTROLS

_2EzQQ-- _25tDQ ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL IQQzQQ__ ___________ Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

it__EBINQ1 ELE 1_QE_NUGLEeB_EQWEB_ELoNI_QEEBoI1QNt PAGE 2 IHEBdQQ1NedIQ1t_BE81_IB8NSEEB_8ND_ELu1D_ELQW 3

QUESTION 1.01 (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 1.02 (2.00)

c. Explain the production and removal of xenon (Xe) in the core (no mathematical equations required). (1.00) i

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? C0.50)-

2. When will Samarium (Sm) reach its peak concentration in the core and what will be its maximum reactivity worth? (0.50)

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

11__EBINCIELES_QE_NUGLE88_EQWEB_EL8HI_QEEB8I1QN t PAGE 3 IBEBdQQ1Ned1GSt_BE81_IB8NSEEB_8ND_ELU10_ELQW QUESTION 1.03 (2.50)

e. A variable speed centrifugal pump is operating at 1/4 rated speed in a " CLOSED" system with the following parameters:

1) Power = 200KW

2) Pump ' delta' P = 80 psid

3) Flow = 500 gpm What are the new values for these parameters when the pump speed is increased to full rated speed? (1.50)

b. Choose the answer that most correctly completes the sentence.

"In.a. closed system, two single stage centrifugal pumps operating in parallel will have ___(choose from below)___, as )

compared to the same system with one single stage centrifugal pump operating with one pump isolated."

1) a higher head and higher flow rate.

2) the same head and a higher flow rate.

3) the same head and the same flow rate.

4) a higher head and the same flow rate. (0.50)

c. How is the available NP'H for the pump affected by an increase in system flowrate? (0.50)

QUESTION 1.04 (2.50)

How will the following affect the Moderator Temperature Coefficient?

BRIEFLY EXPLAIN your answer.

a. The charging pump suction inadvertently switches to the Refueling Water Storage Tank. (1.00)

b. The core ages from BOL to EOL. (0.50)

c. The RCS is cooied down from 550 F to 450 F. (1.00)

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

i

li__EBINCIELES_QE_ NUCLE 88_EQWEB_ELANI_QEEB6I1QNt PAGE 4 IHEBMQQ1N6MIGSt_HE6I_IBeNSEEB_8ND_ELu1D_ELQW QUESTION 1.05 (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,

s. 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 (1.50) -

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

Explain your answer assuming feedwater flow remains constant.

QUESTION 1.07 (2.25)

Answer the following questions regarding Reactor Safety Limits:

a. State which reactor plant parameters ar e limited by Reactor Safety Limits. (0.75)

b. What TWO adverse conditions are prevented by adherence to Reactor

Safety Limits? (1.00) ,

c. How are Limiting Safety System Settings (LSSS) related to Reactor Safety Limits? (0.50)

QUESTION 1.08 (2.50)

s. Define Shutdown Margin (SDM). (1.00)

b. Give three reasons for Rod Insertion Limits. (1.50) i

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

Iz__EBINQIELES_QE_NUQLE68_EQWEB_ELANI_QEEBal1QN t PAGE 5 IBEBdQQ1N80101&_HE81_IB6NSEEB_8ND_ELVIQ_ELQW QUESTION 1.09 (2.00)

e. 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) i QUESTION 1.10 (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)

l. QUESTION 1.11 (1.00)

Choose the CORRECT response. In order to maintain a 200 F subcooling margin 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 1.12 (2.00)

How does the Fuel Temperature (doppler) Coefficient (FTC) vary (more neg.,

less neg., unchanged) for the following conditions / situations. Consider each separately.

a. BOL to EOL. (0.50)

b. RCS boron concentration increases by 50 ppm. (0.50) i

c. 0% to 1004 power. (0.50)

d. Group D rods changes position from 200 steps to 150 steps. (0.50)

]

f

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

5

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

1, e e J I

i i '

1&__EBINQIELES_QE_NUGLE68_EQWEB_EL8NI_QEEB611QNt - PAGE 6 IBEBdQQ1N851QSt_BE8I_IB6NSEEB_8ND_ELVIQ_ELQW '

't

, y ii r

/

-QUESTION 1.13 (1.50) <'

r,; f, The plant is ' op e r s't ing at 100% power with RCS Tave at 587 F and a steam pressure of 980 psig. What must Teve be changed to in order to maintain these conditions with 10% of the steam generator tubes plugged. SHOW ALL WORK, including any applicable formulas.

Y i

f 1

i la J ,

t 2

l

/

4 i l

2-(***** END OF CATEGORY 01 *****)

2t__EL8NI_ DESIGN _INCLUDINQ_18EEI1_8NQ_EMEBGENQ1_11SIEd3 PAGE 7 QUESTION 2.01 (2.00)

e. List TWO reasons for maintaining a minimum pressurizer spray

! line flow during normal "at power" operations. (1.00)

b. What indication or annunciation is available to alert the operator that minimum spray flow is not being maintained? (0.50)

c. What creates the motive force for pressurizer spray flow? (3.50)

QUESTION 2.02 (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-112A)

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 2.03 (1.50)

e. 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 Auxiliary 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 2.04 (2.00)

State six locations where the Chemical and Volume Control System (CVCS) interfaces with the Reactor Coolant System (RCS). (2.00)

QUESTION 2.05 (2.00)

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

(

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

21__EL8MI_QE11GN_INCLUQ1NG_S8EEIY_8NQ_EMEBGENQ1_1111Ed3 PAGE 8 QUESTION 2.06 (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 2.07 (2.00)

. List four signals (non-similar/ unique) which will initiate a motor driven Auxiliary Feedwater Actuation Signal (AFAS).(setpoints not required) (2.00)

QUESTION 2.08 (3.On)

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)

c. What buses supply power to the containment spray pumps? (1.00) l i

QUESTION 2.09 (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)

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

2t__ELeMI_QE11EN_ INCLUDING _SaEEIl_aND_EMEBEEUQ1_S11IEd3 PAGE 9

, QUESTION 2.10 (2.90)

a. What are three conditions that will generate a Safety Inj ect ion signal? Include setpoints and coincidences. (0.90)

b. During a continued RCS depressurization caused by a LOCA, indicate the order in which the ECCS subsystems will inj ect into the RCS and the pressure at which each will inject. (2.00)

QUESTION 2.11 (2.10)

Starting from a diesel generator, explain how power would be supplied to a Vital Instrument Bus. Include major components and voltages. (2.10)

QUESTION 2.12 (2.00) l i For the following components, indicate whether they will receive an OPEN, CLOSE, or N0 signal upon a MANUAL SAFETY INJECTION INITIATION.

a. 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)

e. Main steam isolation valves (0.20)

f. RWST to centrif ugal 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)

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

l j

l Sz__IN11BVMENIl_8NQ_QQNIBQL1 PAGE 10 !

QUESTION 3.01 (2.00)

e. At what pressure is SI blocked during a controlled cooldown? C0.50) '

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 (1.75)

a. Identify each of the regions on the gas filled detector curve (figure EXC-1). (1.00)

b. 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.

Describe the initial automatic actions for the following occurences (alarms not required):

a. 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) i

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

i

at__IN11BudENI1_ANQ_QQNIBQLS PAGE 11 QUESTION 3.04 (2.00)

a. 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 ACTUAL steam flow if, during a power increase from 0-100%, the steam pressure signal stuck at it's 50% power value. (1.00)

QUESTION 3.05 (1.50)

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

QUESTION 3.06 (2.25)

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

a. .What immediate component responses will be initiated by the channel failure? (1.75)

b. If no operator action is taken, what will eventually trip the plant?

(0.50)

QUESTION 3.07 (2.50)

a. What is the purpose of the Overtemperature N-16 and the Overpower N-16 trips? (1.00)

b. When is the Overtemperature N-16 Turbine Runback signal initiated? How is the plant automaticaly restored to acceptable conditions? (1.50)

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

-2t__INSIBudENI1_eND_GQNIBQL1 PAGE 12 QUESTION 3.08 (2.50)

e. If both Intermediate Range Detectors are overcompensated, what could happen during.a reactor shutdown regarding the source range detectors and the Reactor Protection System? Explain. (1.00)

b. How is nuclear power indication adj usted to match secondary power following a calorimetric? (0.50)

c. How is the output signal from the detector current comparator generat-ed? What is the output signal used for? Indicate setpoint(s) if appro-priate. (1.00) 4 QUESTION 3.09 (2.50)

What are five (5) plant parameters which can be monitored from the Hot Shutdown Panel?

QUESTION 3.10 (2.00)

Indicate what automatic actions, if any, will occur for the following Radiation Monitoring System alarms.

a. 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 8

d. High level on the Control Room air supply monitor with the ventilation system in normal operation.

l l

l i

l

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

l

2t__lNil8MMENIS_8NQ_GQN18QL3 PAGE- 13 QUESTION 3.11 (3.00)

c. 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 to RHR pump isolation valves (8811A and 88118)? (0.50)

c. What interlocks prevent the operator from opening the RHR to safety in-jection pump and charging pump discharge isolation valves (8804A and

-88048)? (1.50)

QUESTION 3.12 (1.00)

What are four containment parameters available to the operator as post accident monitoring indication? (1.00)

J 4

(

(***** END OF CATEGORY 03 *****)

st__EBQQEQMBEl_:_NQBd8Lt_8ENQBd8Lt_EdE8QENGl_8NQ PAGE 14 88Q10LQQ108L_QQUIBQL 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 j lifetime dose, he has received 1100 mrem in the present quarter and 200 mrem in the present week.

4 c. 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)

There are four ' Design Basis Accidents' for which the ECCS systems are designed to limit or mitigate the consequences. Give three of the four. (1.50) i QUESTION 4.03 (3.00) l Answer the following questions utilizing information that would be found I

in FRS-0.1, Response to Nuclear Power Generation /ATWS.

a. 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 4.04 (3.50)

In accordance with E0P-2.0, FAULTED STEAM GENERATOR ISOLATION, what are the seven actions required to isolate a steam generator. (3.50)

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

dz__EBQQEQMBE1_:_NQBd6Lt_6ENQBd6Lt_EdEBGENQ1_6NQ PAGE 15 86010LQQ1Q6L_GQNIBQL QUESTION 4.05 (1.50)

According to ECA-0.0, LOSS OF ALL AC POWER:

O. When can the recovery actions be implemented? (0.50)

b. What action is required if SI is active or actuates while in ECA-0.0?

(0.50)

c. If a steam generator tube rupture is discovered, why must the steam generator pressure NOT be reduced to less than 170 psig? (0.50)

QUESTION 4.06 (1.50)

n. What is the condenser low vacuum alarm s.etpoint? (0.25)

b. Identify three conditions which will result in 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 4.07 (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 4.08 (2.00)

a. In accordance with E0S-1.2, Post LOCA Cooldown and Depressurization, why might pressurizer level rapidly increase? (0.50)

b. In accordance with E0P-1.0, Loss of Reactor and Secondary Coolant, what is the cold leg recirculation switchover criteria? (0.50)

c. In accordance with EOS-1.3, Transfer to Cold Leg Recirculation: 1) What is the criteria for stopping any pump taking a suction from the RWST (0.50), and 2) what is the criteria for switching to an alternate AFW water supply (0.50)? (1.00)

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

'dA__E8QQEQMBE1_ _NQBd8Lt_8BNQBd8Lt_EdEBGENQX_8NQ PAGE 16 B8Q10LQQ1Q8L_GQNIBQL 1

QUESTION 4.09 (1.00) i 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 situations could SI be terminated?

PZR LVL SUBC00 LING PRESSURE

.a. - 4% 60 degrees stable

b. 25% 10 degrees increasing

c. 10% 40 degrees decreasing

d. 20% 15 degrees stable 4

QUESTION 4.10 (1.00)

Which of the following statements is correct concerning the status of the Nuclear Instrumentation Recorder (1-NR-45) ouring control bank rod

, withdrawal to. criticality?

c. Both source range channels and the highest reading intermediate j 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.

4 t

t J

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

= - - ~- - - , , ----- ,,-- ,,-vn-n ---~;..,,-,n -,--.----.---,,,,------~-,,,.-------,e -u----en----p-n,-,---- - - .,-,-~v -----e

8t__88QQEQUBEl_:_dQBd8Lt_898QBd8Lt_EDE89ENQX_8NQ PAGE 17 88QIDLQQ198L_QQNIBQL QUESTION 4.11 (1.50)

The plant is being maintained in hot standby.

c. What is the Shutdown Margin limit? (0.50)

b. When conducting a Reactor Shutdown Margin Verification, what is the difference between calculating a UNCORRECTED maximum boron concentrat-ion and calculating a CORRECTED maximum boron concentration? (exclude the requirement of when they are to be calculated and do not use any formulas) (0.50)

c. If the Shutdown Margin is not in specification how is it brought into specification? (0.50)

QUESTION 4.12 (2.00)

a. Once in the E0P's, what are four conditions that result in trans-itioning to E0P-3.0, Steam Generator Tube Rupture? (1.60)

b. What is the purpose behind isolating the ruptured steam generator?

(0.40)

QUESTION 4.13 (2.50)

What are the four precautions stated in SOP-102A, Residual Heat Removal System, that must be followed during startup of the system during plant chutdown from hot standby to cold shutdown? (2.00)

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

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

. It__EBINCIELES_QE_NVQLE88_EQWEB_EL6NI_QEEBaIIQNt PAGE 18 IHEBdQQ1NedIQSt_HE61_IB6NSEEB_6NQ_ELMIQ_ELQW ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL i

1 ANSWER 1.01 (1.25)

1. Greater than 15 degrees F (0.25)

2. Stable or decreasing (0.25)

3. Stable or decreasing (0.25)

4. Stable 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 1.02 (2.00)

a. 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 0.00278 hours 1.653439e-5 weeks 3.805e-6 months (0.25) 6300 + or - 200 pcm(0.25) (0.50)

2. 15 to 17 days (0.25) 1065 + or - 100 pcm(0.25) (0.50)

)

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

REFERENCE Reactor Core Control Ch. 4 pgs. 4-11 to 4-13 3.1 001 000 K 5.13 3.7 l 5.38 3.5 i __________________________________________________________________________

7

It__EBIMGIELES_QE_NUGLE88 EQWEB_EL8NI_QEEBellQNi PAGE 19 IBEBdQQ1N851GSt_BE61_IB6NSEEB_eNQ_ELu1D_ELQW ANSWERS -- COMANCHE PEAK 1- -86/09/23-WHITTEMORE, J./GUIL ANSWER 1.03 (2.50) 3 e. Power (2) = Power (1) * (N2/N1)**3 = 200 * (4)**3 = 12.8 Mw (0.50)

. Delta PC2) = Delta PC1) * (N2/N1)**2 = 80 * (4)**2 = 1280 psid (0.50)

Flow (2) = Flow (1) * (N2/N1) = 500

4 = 2000 gpm (0.50)

b. #1 (0.50)

c. Decreases (0.50)

REFERENCE

, Thermal-Hydraulic Prin., pgs. 10-32 to 10-48 Appendix 9 Components _ Pumps 2.7 l

i ANSWER 1.04 (2.50)

I

c. LESS NEGATIVE (0.50) More boron to leave core area per degree temperature change. (0.50) (Or equivalent answer) (1.00)

b. MORE NEGATIVE (0.25) Less boron, opposite result as above.(0.25) (0.50)

c. LESS NEGATIVE (0.50) Water density changes are less as j temperature is reduced.CO.50) (1.00)

REFERENCE

Reactor Core Control, pgs. 3-17,3-18,3-20,and 3-25 3.1 001 000 K 5.15 3.4 P

l i

It__EBIBQ1ELES_QE_NVQLEeB_EQWEB_EL6NI_QEEB8I196t PAGE 20 IBEBdQQ1Nad10St_UEeI_IBauSEEB_86D_ELVID_ELQW ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 1.05 (2.00)

o. 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 ANSWER 1.06 (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 soparators (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

11__EB1501ELES_QE_N90LEaB_EQWEB_EL6NI_QEEB6IIQN t PAGE 21

'IBEBdQQ1NedIQ1t_HE8I_IB8NSEEB_8ND_ELul0_ELQW ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL 4

i ANSWER 1.07 (2.25)

s. 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) 4 (0.50) i REFERENCE j Thermal-Hydraulic Prin. II, pg. 13-53 Tech Spec 2.1 and 2.2 2.1 System-wide Generic 5 2.9 f

ANSWER 1.08 (2.50)

a. SOM shall be the instantaneous amount of reactivity (0.25) by which the

reactor is subcritical or would be subcritical from its presen. cond_

ition(0.25) assuming all full length rod cluster assemblies are fully inserted (0.25) except for the single rod cluster assembly of the high-reactivity worth which is assumed to be fully withdrawn (0.25). (1.00)

! b. 1. To minimize the consequences of an ejected rod accident. (0.50)

2. Quarantee a sufficient shutdown margin from a given power level.

(0.50)

3. Provide an axial flux distribution which prevents high Iccal peak

> power levels.(Also accept wording alluding to acceptable power distribution limits are maintained.) (0.50) i REFERENCE Technical Specifications (TS) Unit 1, pg. 1-5; TS BASES 99 3/4 1-3 l Reactod Core Control Ch. 6, pg. 6-30 l

j 3.1 001 000 K 5.08 3.9 4

i

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

lx__EB1501ELES_QE_ NUCLE 68_EQWEB_EL8NI_QEEBoI1QNt PAGE 22 IHEBdQQIN851 cst _BE6I_IB88SEEB_6NQ_ELu1D_ELQW ANSWERS -- COMANCHE PEAK 1 _86/09/23-WHITTEMORE, J./GUIL ANSWER 1.09 (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 ANSWER 1.10 (2.00)

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

When reactor power increases fuel temperature increases which adds negative reactivity 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 1.11 (1.00)

c ---- (1.00)

REFERENCE Steam Tables

- it__EBINCIELE1_QE_NVQLE88_EQWEB_EL8NI_QEEBal1QNt PAGE 23 LINEBMQQIN8MIQit_BEaI_IB6NSEEB_eNQ_ELVIQ_ELQW ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 1.12 (2.00)

e. More negative. (0.50)

b. No change. (0.50)

c. Less negative. (0.50)

d. No change. (0.50)

. REFERENCE Reactor Core Control Ch. 2, pgs. 2-23, 2-39 and 2-40 3.1 001 000 K 5.48 3.3 K 5.49 3.4 ANSWER 1.13 (1.50)

S/G heat transfer = Q = UACTavg - Tstm)

Q,U, and Tstm remain constant; A1(Tavg1 - Tstm) = A2(Tavg2 - Tstm) (0.50)

GIVEN: A2 = 0.9 X Al From Steam Tables: Tsat for 995 psia = 544 F (0.50)

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

Tavg2 = 591.8 F (591 to 592.5 F acceptable) (0.50)

REFERENCE CP Thermal-Hydraulic Principles, Ch. 12, pg. 8 i

4

,, -,v_.-~,,n--.,_ . _ - . _ , , . .. -. ._..._, ._,.-,-,-- - , - .,.-,..--,-.,--,,.,,,,,-,,,,n.n_-,---___,,-., -

21-_ELeMI_DE110N_INDLUDINQ_18EEIl eND_EMEBQENQ1_11 HIED 1 PAGE 24 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL 1

ANSWER 2.01 (2.00)

o. 1) To reduce thermal stress to the spray line and spray nozzle. (0.50)

2) To maintain Pzr. chemistry uniform with the RCS. (0.50)

b. 1) Spray line low temperature alarm (decreasing or below normal temp-erature). (0.25)

2) Surge line low temperature alarm (decreasing or below normal temp-erature). (0.25) t

c. Differential pressure across the reactor vessel (will also RCP dP) (0.50)

REFERENCE CP SD VOL II, pgs. 1.26 to 1.27 ALM-0053A, pgs. 13 and_19 3.3 010 000 K 4.01 2.7 K 1.03 3.6 K 6.06 2.3 ANSWER 2.02 (1.50)

(0.25 each)

. 1. Fail closed

2. Fail flow to the VCT

3. Fail open

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 i

I

2t__PL8MI_QE$1GN_INQLQQ1NQ_18EEII_6NQ_EdEBQENQ1_111IEd3 PAGE 25 ANSWERS _- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL 1

ANSWER 2.03 (1.50)

e. No. 11 - 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 Storage 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 2.04 (2.00)

Letdown from crossover leg #3 (0.10 for system, 0.25 for loop) i Excess letdesn from crossover leg # 1

Charging to cold leg #4

, Alternate charging to cold leg # 1

{ Auxiliary spray (0.30) l RCP seal inj ect ion (0.30) i REFERENCE CP SD VOL II, pg. 2.32 3.1 004 000 K 4.05 3.3 j __________________________________________________________________________

i l

5 L

l

2A__EL8MI_DE11GN_INGLMQ1NG_18EEI1_eNQ_EMEBGENGl_11SIgd$ PAGE 26 l

l ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 2.05 (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)

Regenerative 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 SD Vol II, pgs. 2.19,2.32,3.26, and 3.42 3.1 004 010 K 6.09 4.4 ANSWER 2.06 (2.00)

a. 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 berrier (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, pg3 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

21__EL6HI_QE11EN_INGLUDING_SeEEI1_6NQ_EMEBGENGY_111IEd3 PAGE 27 s

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

i ANSWER 2.07 (2.00)

1. Manual

2. S/G low - low level on 1 steam generator

3. Safety injection sequence signal

4. Trip of both Main Feed Pumps

5. Blackout sequence signal Cany 4 at 0.50 each) i REFERENCE

, CP SO VOL III, pgs. 3.23 to 3.25 i~

3.5 061 000 K 4.02 4.5 i __________________________________________________________________________

4 i

ANSWER 2.08 (3.00)

, c. 1) Design basis LOCA (0.25)

Steam line rupture (0.25)

Feed line rupture (0.25)

2) Pressure (0.25)

, Temperature (0.25)

J

. b. 1) RWST (0.25)

2) Containment Sump (0.25)

3) Manual (0.25) i

c. 1) 11 - 1EA1 6.9KV Safeguards Bus (0.25)

'2) 12 - 1EA2 6.9KV Safeguards Bus (0.25)

3) 13 - 1EA1 6.9KV Safeguards Bus (0.25)

I 4) 14 - 1EA2 6.9KV Safeguards Bus (0.25) l REFERENCE

,- CP SD Vol II, pgs. 10-2 and 10-14 P&ID El-0004 i

! - 3. 6 026 000 K 4.04 3.7 3.6 028 000 System generic 4 3.5

K 5.03 2.9 l

l i

i

s 2t__EL6MI_DE11GN INQLUDINQ_16EEI1_6NQ EMEBQENQ1 111IEMS PAGE 28 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 2.09 (2.00)

c. 11 - 1EA1 6.9KV Safeguards Bus (0.50) 12 - 1EA2 6.9KV Safeguards Bus (0.50)

b. Low head injection during injection 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 (0.25)

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 ANSWER 2.10 (2.90)

e. Low pressurizer pressure (0.10) 1829 psig(0.10) 2/4(0.10)

High contaiment pressure (0.10) 3.35 psig(0.10) 2/3(0.10)

Low compensated steamline pressure (0.10) 605 psig(0.10) 2/3(0.10)

(Manual (accepted for no credit) 1/2 3/4)

(0.90) i b. High head inj ect ion (0.20) 1829 psig(immediate) (0.25)

Intermediate head inj ect ion ( 0. 2 0 ) 1520 + or - 50 psig (0.25)

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

Low head inj ect ion (0.20) 195 + or - 20 osig (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 4 K A.03 3.6 l

l i

l l

s 2z__ PLANI _ DESIGN _INGLUQ1NQ_SeEEIl_eNQ_EMEBGENQ1_SYSIEd$ PAGE 29 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 2.11 (2.10)

The diesel generators tie into the 6.9 KV Safeguards Buses (0.30). The voltage is stepped down by 6900/480 V transformers (0.30) to the 480 V Scfeguards Buses (0.30). 480 V is then supplied to an inverter where it to transformed to 125 V C0.30) and rectified to DC (0.30). The rectified 125 VOC is auctionetred with the 125 V DC Bus voltage (0.30) and the high-or voltage is supplied to the inverter which feeds the 118 V AC instrument bus-(0.30). (2.10)

REFERENCE CP SD VOL III, pgs. 1.13 to 1.20

.P&ID El-0018. 0005, AND 0004 3.7 064 000 K 1.01 4.1 ANSWER 2.12 (2.00)

(0.20 pts each)

e. CLOSE

b. CLOSE

c. OPEN

d. CLOSE

e. N0

f. OPEN

g. 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__IN11BWNENI1_8NQ_QQNIBQL1 PAGE 30 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER- 3.01 (2.00)

.e. Below 1960 psig and above 1829 psig(any value in between). (0.50)

o. 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- (1.75)

a. Figure EXC-1 (1.00)

b. 1) Ionization (0.25)

2) Recombination (0.25) !

3) Proportional (0.25)

REFERENCE CP SD VOL III, pgs. 1.4 and 1.5 f 3.9 015 000 K 6.01 2.9 4

4 1

J j

}

i 1

r 1

l

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

2z__INSIBydENIS_6NQ_QQN18QL$ PAGE 31 ANSWERS -- COMANCHE PEAK 1 _86/09/23-WHITTEMORE, J./GUIL ANSWER 3.03 (2.00)

c. 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

's',

-x ,

m.

- 2A__INSIBMMENll 88Q GQN18QL3 S PAGE 32

; t s- .. .

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

z 1

~'

- ANSWER 3.05 (1.50) i The system will arm (PT506 feeds the arming signal) (0.75) but there will

- be no demand signal because PT505 provides th'e T~ref input for the demand cignal (0.75). (1.50)

,+ L' ,

1 REFERENCE CP ABN-709A

- 3.9 016 000 K 3.03 3.0 A 2.01 3.0

[

ANSWER 3.06 (2.25)

a. 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 C0.46) (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 4

t J

1

- - , - - --.n.-- ,. -, .c, -n_-.,..--,-,---..,,w--.- ,-,e ,e,n+ , ~ ,,.ew--g, . ,,en.ne,---r-, ,--r . -r- - -

21__INSIBudENI1_8NQ GQUIBQL1 PAGE 33 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL 4

4 ANSWER 3.07 (2.50)

c. OT N To ensure operation within the DNB criteria. (0.50)

OP N To prevent excessive fuel rod power (Kw/Ft) (will accept limit heat generation rate). (0.50)

b. The signal is initiated at 3% below the reactor trip set point. (0.50).

The turbine control svatem decreases turbine load (0.50) by running the

turbine back (0.50). (1.50)

REFERENCE CP SD VOL III, pgs. 9.11, 10.4 and 10.5 2

VIII, pg. 9.49 I

3.9 012 000 K 4.02 3.9 K 1.03 3.7

, K 1.06 3.1 i AMSWER 3.08 (2.50)

e. Overcompensation results in a lower indicated flux which could caus<:

j the source range detectors to be reinstated too early (0.50) thus causing a reactor trip from source range high flux trip (0.50).

(1.00)

b. Adj ust the gain of the summing and level amplifier.(Also accept gain adj ustment on the front of the power range drawer.) (0.50)

c. Signal outputs from the detectors are compared with the average of the corresponding signal from the appropriate detector sections (0.50). A 2% deviation (0.25) from the average will actuate an alarm.(0.25)

(Also accept compares each of four inputs (upper / lower][0.50) with 102% of the average input. If any one input equals or exceeds the 102%

value(0.2S] and reactor power is above 50% power an alarm will actuate.

l (0.25)) (1.00)

REFERENCE CP SD VOL III, pgs. 1.5, 1.10, and 1.11 I

3.9 015 000 K 3.01 3.9 System generic K 4 3.6 4

i i

2A__ld3I8VUEU13_AUQ_QQ31BQL3 PAGE 34 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 3.09 (2.50)

PZR pressure PZR level AFW flow to each SG RCS wide range Th Source Range NI SG 1evel SG pressure Letdown flow Charging pump to CVCS charging and RCP seals flow (any 5 at 0.50 each)

REFERENCE CP IPO-008A attachment 1 3.9 016 000 K 4.01 2.8 l

l l

i ANSWER 3.10 (2.00)

a. 1) Initiates Control Room emercency recirculation. (0.25)

2) Closes HCV-014 in the GWPS. (0.25)

(also accept containment ventilation isolation for no credit.1

b. 1) Isolates SG blowdown. ' 'I . 2 5 )

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) l 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 m._

31__INSIBUDEUI1_8ND_GQUIBQL3 PAGE 35 ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL l

l l

ANSWER 3.11 (3.00)

c. Either sump isolation valve (8811A or 88118) is open. (1.00)

b. The RWST to RHR pump isolation valves (8812A and 8812B) are closed and one of the two RCS suction series valves (8701A & 8702A or 8701B &

8702B) are closed. (0.50)

c. 1. Either 8702A AND 8701A OR 8702B AND 8701B (RCS hot leg suction isolation valves) are open. (0.50)

2. Both 8511A and 85118 are OPEN (CCP emergency miniflow isolation valves) (0.50)

3. Either 8814A or 88148 are OPEN (Safety Inj ection miniflow isolat-ion valves) (0.50)

REFERENCE CP SD VOL II, pg. 8.36 l 3.2 006 000 K 4.08 3.2 K 4.06 3.9 3.6 026 020 K 4.03 4.1 ANSWER 3.12 (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

s e st__EBQEEQMBE1_:_NQBd8Lt_8BNQBd8Lt_EMEBGENQ1_8NQ PAGE 36 88019LQQ1G8L_GQUIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 4.01 (2.50) l

a. The Radiation Protection Engineer must approve exceeding the I administrative quarterly limit (0.50). The Radiation Protection l Supervisor must approve exceeding the administrative weekly limit.

(If assumed 1300 mrem total dose received for quarter then only the Rad-istion Protection Supervisor's approval is needed) (0.50) (1.00)

b. 3000 mrem /Q = 1100 mrem + 200 mrern/hr X T ; T = 9.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (0.50) 1300 mrem + 200 mrein/hr X T ; T = 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

c. Oose 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)

(Any 3 of 4, 0.50 each)

1. Rod Ejection

2. Loss of Secondary Coolant Accident (steamline rupture)

3. Steam Generator Tube Rupture

4. Loss of Coolant Accident (LOCA) which results in a coolant leak greater than the capability of the normal charging system. (1.50) l REFERENCE CP SD VOL II, pg. 8.5 3.2 006 030 A 2.01 4.5

~ .

4 t__EBQQEQUBEl_=_NQBd6Lt_6BNQBd6Lt_EMEBGENQ1_6NQ PAGE 37 B6DIQLQQIQ6L_GQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 4.03 (3.00)

s. 1) Reactor trip and by-pass breakers open (0.34)

2) Neutron flux decreasing (0.33) -

3) All DRP1 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 ANSWER 4.04 (3.50)

1) Isolate main steam line.

2) Isolate main feed line.

3) Isolate AFW flow.

4) Place TOAFW pump steam supply valve in PULL-TO-LOCK.

5) Isolate blowdown and sample lines.

6) Verify SG steam line atmospheric dump valves closed.

7) Verify main steam line drip pot isolation valves closed. (0.50 each)

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

, n --w. -

m-y. - , - - , , ,m, - . . . - - - - - - , , ,r- - - - . - - , - - -. - - = . . , - -

. _ - . . - .. .. . . = . .-. .. ..

~ .

st__EBQEEQUBEl_ _NQBd8Lt_8HNQBd8(t_EMEBQENQ1_8NQ PAGE 38 88Q10LQQ1Q8L_QQNIBQL ANSWERS -- COMANCHE PEAK 1 _86/09/23-WHITTEMORE, J./GUIL ANSWER 4.05 (1.50)

e. When at least one AC Safeguards Bus is restored. (0.50) c b. .SI should be r e s t '- (to permit manual loading of equipment on an AC i Safeguards 3us. (0.50)

c. To prevent inj ection of accumulator nitrogen into the RCS. (0.50)

REFERENCE CP ECA-0.0, pgs. 5 and 9 3.7 000 055 EK 3.02 4.3 ANSWER 4.06 (1.50)

s. 22.6 + or - 1 in. Hg vacuum (0.25)

b. 1) Loss of circulating water

i. 2) High condenser water level j 3) Loss of gland seal steam system j 4) Loss of CEV pumps

5) Air leakage Cany three for 0.75) i l c. Turbine trip (0.25) at 21 in. Hg vacuum (0.25) (0.50)

REFERENCE CP Alarm Procedure 1-ALB-90 I

i ___________________________________________________________________________

I i

l f

i

. ~ , , , . ._,,____,__,_____m - _ _ _ . . _ _ , _ , _ _ , _ , . ,_ ___ , , , . _ _ _ , . . _ . . , _ . _ . . . _ . ,

~ .

st__EBQEEQUBES_ _NQBd8kt_6ENQBd6Lt_EdEBQENQY_6ND PAGE 39 88DIQLQQ106L_QQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL ANSWER 4.07 (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 ANSWER 4.08 (2.00)

a. Voiding in RCS (0.50)

b. When RWST level decreases to less than 40%. (0.50)

c. 1) When the RWST EMPTY alarm actuates at 15% level. (0.50)

2) When CST level decreases to less than 10%. (0.50)

REFERENCE EOS-1.2, pg. 7 E0P-1.0, pg. 13 EOS_1.3, pgs. 4 and 8 3.3 000 009 EK 3.10 3.4 3.3 000 011 EK 3.12 4.4 3.3 000 011 EA 1.10 4.1 i

l

! ANSWER 4.09 (1.00)

l

--d.-- (1.00)

REFERENCE E0P_0.0, pg. 11 k-

s e 8t__BBQEEQMBEl_: NQBd8Lt_8ENQBd8Lt_EdEBQENQ1_6NQ PAGE 40 88Q10LQQIQ8L_QQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL 3.2 006 020 K 4.06 3.9 ANSWER 4.10 (1.00) i

_-b.-- (1.00)

REFERENCE CP IPO-002A, pg. 6 3.9 015 000 3.01 3.8 ANSWER 4.11 (1.50)

a. Limited to greater than 1.6% delta k/k. (0.50)

b. The uncorrected maximum boron concentration calculation does not take into account Xenon or Samarium in the reactor. (0.50) l b. Immediately initiate rapid boration. (0.50) l i

l REFERENCE TECH SPEC 3.1.1.1 and 4.1.1.1.1 CP OPT-301, PG. 2

! 3.1 004 020 Sys Gen 5 2.9 l _____---______--_-_______----________---__________-__---______-____________

l l

l i

8t__E8QEfQUEll_: NQBd8kt_8DNQ858Lt_EME8QENQ1_8NQ PAGE 41 B8Q10LQQ1Q8L_QQNIBQL ANSWERS -- COMANCHE PEAK 1 -86/09/23-WHITTEMORE, J./GUIL 4

ANSWER 4.12 (2.00)

c. 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-down of the RCS. (0.20)

REFERENCE E0P_0.0, pg. 10 E0P-3.0, pg. 2 ERG-HP Background E-3 ECA-3 3.3 000 037 EK 3.07 4.2 4

i ANSWER 4.13 (2.50)

1) RCS temperature must be less than 350 degrees F (0.25)

RCS pressure must be less than 450 psig (0.25)

2) Boron concentration in the RHRS must be equal to or greater than the RCS before placing the RHRS in operation. (0.50)

3) Flow through the RHRS must be initiated slowly to prevent thermal shock. (0.50)

4) RHRS heatup(0.25) and cooldown(0.25) rates should not exceed 50 degrees F in any one hour (0.50) (1.00)

REFERENCE CP SOP-102A, pg. 5 3.4 SYS GEN 7 3.5

-- . . _. . _ _ . -

ML20214N998

Text

SUMMARY

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