ML20205C462

Exam Rept 50-285/OL-87-02 During Wk of 870302.Exam Results: All Four Senior Reactor Operator Candidates Passed Exams.Two Candidates Taking Requalification Exams Satisfactory

ML20205C462
Person / Time
Site:	Fort Calhoun
Issue date:	03/20/1987
From:	Cooley R, Whittemore J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20205C377	List: ML20205C372 ML20205C462
References
50-285-OL-87-02, 50-285-OL-87-2, NUDOCS 8703300186
Download: ML20205C462 (52)

Text

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FORT CALHOUN STATION OPERATOR LICENSE EXAMINATION REPORT No. 50-285/0L-87 Licensee: Omaha Public Power District 1623 Harney_ Street Omaha, Nebraska 68102 Docket:

50-285 License No: DPR-40 Operator License examinations at Fort Calhoun Generating Station (FCGS) 3!ao!87 Chief Examiner:

7. E.-Whittemore, Examiner Date Approved by:

bhD R. A. Cooley, Sect 1 on Chief D~ ate '

Summary:

NRC administered four Senior Reactor Operator (SRO) examinations at FCGS during the week of March 2, 1987. All fcur candidates successfully completed the examinations. The two candidates taking operating requalification examinations were evaluated to be satisfactory.

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4 REPORT DETAILS 1.

- PERSONS EXAMINED License Examinations:

TYPEi SR0 UPGRADE

. TOTAL' PASS 4 - 100%

4 - 100%

FAIL 0 - 0%

0 - 0%

TYPE REQUALIFICATION TOTAL PASS 2 - 100%

2 - 100%

FAIL 0 - 0%

'0 - 0%

2.

EXAMINERS J.Whittemore,NRC(ChiefExaminer)

- J. Pellet, NRC

3. -

EXAMINATION REPORT Individual performance results are not included in this report because these reports are placed in the NRC Public Document Room, a.

EXAMINATION REVIEW COMMENT RESOLUTION i

In general, editorial comments or changes made during the-i examination, review, or subsequent grading reviews are not addressed by this resolution section. -This section reflects resolution of substantive coments made in Omaha Public Power District written response to the written examination content. The proposed modifications that have been accepted are included in the master -

examination key, comprising Section 3.d of this report, as are all other changes mentioned above but not discussed herein. Coments not accepted are addressed below.- Note that coments from the licensee response are paraphrased for brevity. The full text of the coments as received from the licensee is included in Section 3.e of this report.

5.08 The question could be misleading since the examinee may have assumed a constant fuel temperature in all cases, in which case (b), (c), and (d) would not change.

It is suggested that the question be revised to include the following: " Assume Tave and reactor power are the same in all four cases.

Resp.

The comment is noted and the examiner sees no basis for agreement.

3

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5.12.b An additional correct answer is one based on the note from Fig. 4-2 in the Emergency Operating Procedures which states:

"Any time the RCS has experienced an uncontrolled cooldown causing RCS temperature to go below 500 oF."

A copy of Fig. 4-2 is attached. E0P-1 instructs the operator to maintain RCS pressure and temperature between the 20 0F subcooled curve and the 200 oF. subcooled curve. Any uncontrolled cooldown causing the RCS temperature to go below 500 oF, which takes the RCS temperature and pressure to the left of the 200 of subcooled curve results in PTS conditions.

Resp.

This comment is noted and the examiner agrees. However there will be no change in the key as this scenario is inherent in the existing key.

6.02.a An additional initiator of load shed is Offsite Power Low Signal (0PLS). This results from coincident occurrence of degraded voltage and SIAS. This is also reflected by the existing answer key for part b.

Resp.

This cannot be accepted as a valid answer as the question asks for the types of load shedding, not individual initiating events.

6.03.b The question could have been interpreted in such a way that an answer which discusses the two method as

" switching to a live bus" and " switching to a dead bus" should also be accepted for full credit. As indicated in the attached pages from Operating Instruction 01-EE-4, the methods used for transferring to a live bus differ significantly from methods used to transfer to a dead bus.

Resp.

This comment is not accepted as the question solicits design features and attributes, not specifics of procedures.

6.03.d There are other design features that add reliability to the inverters. Any of the following answers should be acceptable for full credit:

(1) Each inverter has its own internal frequency reference.

(2) The ability exists to cross-tie instrument AC buses.

(3) The power supply for the inverters is from a battery which provides backup power in case a battery charger fails.

(4) Spare battery charger.

Resp.

Item (2) is acceptable and already exists in the answer key. The other items are not acceptable as this information was solicited in parts a and c.

The answer key will not deduct credit for the suggested answers but will require the bus cross-tie feature for full credit.

'4

.The other answers address reliability of the inverter power supply, not the bus power supply.

6.05.a An alternate correct answer is to prevent a harsh environment for FW-6 created by a rupture of a steam line for FW-10.

Therefore FW-10 is considered to be the designated emergency pump, and FW-6 is given preferential operation.

It is also correct to mention the undesirable conditions introduced due to the fact that use of FW-10 introduces an unmonitored release path'.

Resp.

Comment is not accepted as the question does not solicit action or symptoms based on accident conditions, but operating method as a result of system design.

6.05.c In order to avoid confusion in this question, it might be reworded as follows: "Why is it recommended that Steam generator level be maintained greater than or equal to 80%

when feeding through the auxiliary feedwater nozzles?"

Comment is noted. Licensee should note that proposed Resp.

~ question wording gives half the answer to part b.

6.06.a It is very unclear as to what is meant by "... options available to the operator for reading these indicators..."

Therefore, a discussion of any of the options available for reading the radiation monitor should accepted for full i

credit. Examples include: All positions, any three decades, alert setpoint.. alarm setpoint, local (meter),

and remote (meter). Note that there are six combinations of three consecutive decades available to the operator.

Resp.

The question asked for the options, not any of the options available to the operator. Therefore full credit could not be allowed unless all of the suggested answers were supplied. The comment is not acceptable and the examiner will require the original answer. Those answers requested by the licensee will be allowed for no credit.

6.07.b Part a. addresses only CCW/RW interface valves.

Therefore, it is logical to assume that part b. is addressing the same valves. No indication is given that valves strictly in the RW system should be considered in the response. An answer which discusses the use of accumulators and describes the interface valves equipped with accumulators should be accepted for full credit.

Resp.

The examiner disagrees with the proposed logical assumation by the examinees. However the required answer will )e expanded to require the mention of interface valves for full credit, effectively increasing the

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5 question score for those examinees who mentioned interface valves.

6.08.b An additional correct answer is that the valve is used for shutdown cooling purification as detailed in valve checklist from OI-SC-5, attached.

Resp.

Not accepted. Inclusion of valve on checklist does not support contention that systems interface is used for this evolution.

6.08.c Additional correct answers for Parts (1) and (2) include filling the spent fuel pool, draining the transfer canal, reducing fuel pool level, failure of pool cooling system, and full core off-load.

Resp.

Comment is noted and examiner agrees. However answer key will not be changed as all these answers are inherent in the existing key.

6.09.a The sequencers are started by receipt of an accident signal, therefore, accident' signal should be acceptable in part 3 of this question Resp.

Comment noted. Answer key will not be changed as proposed change is already inherent in existing answer.

6.09.b A number of additional correct answers exist for this question. These include:

(1) Drain and refill the VCT as part of the venting process.

(2) Lower VCT pressure to allow gasses to come out of solution.

(3) Obtain maximum purification and letdown flow (actually(a way of enhancing filtration and domineralization flow.

4) " Burp" the VCT with hydrogen and nitrogen.

Resp.

Not accepted. The proposed additional answers are procedures, not methodology.

6.10.b Additional correct answers include low temperature over-pressurization, shutdown cooling system valve interlock, and pressurizer pressure low (PPLS initiates SI). Any four answers should be accepted for full credit.

It is recommended that the question be worded to say " List four protective action signals that could be initiated by high or low pressurizer.

Resp.

Coment not accepted.

LTOP PORV actuation and PPLS SI initiation are inherent in existing answer.

SDC valve interlock is not considered to be a protective action.

The comment on question rewording is noted.

7.01.a It is unclear as to what is meant by "Be specific as to technique of determination." Therefore, any answer which

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discusses in some detail the process for generating 1/M plots should be accepted for full credit.

Resp.

Not accepted. The proposed answer is acceptable for partial credit, but the number of channels and the number

.of times the count is performed is the required specific information solicited by the question.

7.02.a

-Two additional correct answers exist:

(1) A correct basis is to reduce probability of fuel failure by fuel i

conditioning.

(2) A correct basis is to limit increases in local power density. Any two of the four answers should be accepted for full credit.

Resp.

Partially accepted.

The answer of fuel conditioning limits is accepted and will be incorporated into the key.

i i

Increasing local power density will not be added to the key, but this answer is inherent in the required answer of

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7.02.b Additional correct answers include: Acceptable power peaking. factors, minimum SDM is maintained, CEA ejection consequences are mitigated, core burnup distribution for i

LSSS setpoints, and negligible effects due to Xenon distribution. Any two answers.should be acceptable for full credit.

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Resp.

.Not accepted. Question asks for the "only two considerations." Any two of the answers above are not the only two considerations. The alternative would be to require all five answers above plus any others that may be remotely applicable.

4 7.02.c To clarify this question, it is suggested that it be l

reworded as follows:

"During initial criticality of a j

newly refueled core, Why may the Reactor Engineer place i

limits on maximum withdrawal of CEA's?

Resp.

Comment noted.

l 7.03.b With the RCS pressure at 1500 psig, pressurizer pressure

]

low signal (PPLS) would be blocked since this must occur i

above 1600 psig. Therefore, an alternate correct answer i

would be to initiate PPLS by placing block switch to the emergency reset position.

It should also be noted that j

the answer provided in the key is incorrect. The correct answer can be ascertained by reviewing the block diagram for the Engineering Safeguards Control actuation which is attached.

Resp.

Partially accepted.

The question asked for automatic, not j

manual initiation. The key answer will be expanded to require the manual initiation of automatic initiation for i

7 half the question credit. The second comment required a telephone call to the licensee in order to understand the problem. The answer key has been changed to reflect that CSAS does not occur as a result of CPHS alone..

7.04.b A0P-19, " Loss of Shutdown Cooling", has no specified "last resort" method for removing residual heat subsequent to a complete loss of shutdown cooling for the condition with the head removed with refueling conditions established.

Since this question was misleading, it requested that question 7.04.b either be deleted from the examination or any reasonable flow path be accepted for full credit. The appropriateness of a flowpath can be verified by referring to the P&ID's for the Fort Calhoun Station.

Resp.

Not accepted.

The procedure describes a flow path to be used for cooling when the head is removed and the refueling pit is flooded with the transfer canal open.

7.05.a Full credit should be given for stating that the recirculation valves have been closed on a RAS or that the recirculation to the SIRWT has been isolated on a RAS.

A0P-23 does not discuss deadheading of the pumps.

Resp.

Coment is noted. Answer is inherent in the existing key.

7.06.a An additional vent path exists, that of venting the pressurizer to the VCT. This path is described in 01-CH-3. While this flow path is not part of the reactor coolant gas vent system, it would be an available flow path and would be considered for use following an accident.

If only the answers in the key are desired, it is suggested that the question be revised as follows:

"During recovery from an accident, it becomes necessary to vent the RCS of non-condensable gasses using the reactor coolant gas vent system..."

Resp.

Not accepted. There is no documentation that this method would be used during accident or accident recovery conditions.

7.07.a An alternate correct answer is that a star (*) preceding a step in the Safety Function Status Check portion of the Emergency Operating Procedures requires the operator to record the value of the parameter.

Resp.

Not accepted. The star (*) requiring a value to be i

recorded does not precede a step but appears as usually within the step.

7.07.b A description of actions taken to attempt to establish reactivity control should be accepted for full credit a

_ _ ~ _

i 8

instead of the statement of attempting to establish -

reactivity control.

In order to aid clarity, it is suggested that the question be reworded as follows: Whct course of action does E0P-20 direct the operator to take if acceptance criteria for. reactivity control cannot be met?

Resp.

The comment concerning rewording the question for clarity is noted. Action necessary to establish reactivity.

reactivity control will be accepted in lieu of the blanket i

statement, However it must be noted that the operator is responsible for other safety functions to obtain full credit for this question and this response will be required.

+

7.09.b It is requested that the answer in the key be replaced with the following: "Cooldown below no-load Tave with no means of makeup to the RCS may result in loss of RCS pressure and inventory control." Void formation is not mentioned in this statement from E0P-02.

Resp.

This comment is noted and ignored as the statement is obviously inherent in the existing answer key.

7.10.b Credit should also be given for the following responses:

(1)

Isolating the steam generator, as detailed in step 3.14 of E0P-04.

(2) Controlling RCS pressure 0-100 psid above isolated steam generator pressure, as detailed in step 3.18 of E0P-04.

Resp.

Not accepted as these reasons were not delineated for this purpose within the procedure.

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7.10.d Voiding in the head should be an acceptable answer in addition to voiding in the RCS.

Resp.

This answer-is already.in the existing key.

.8.02.a This question is extremely difficult to interpret.

It appears that a specific item from the surveillance section of Technical Specifications or a surveillance test is.

being solicited.

It is not apparent that an item appearing in the definitions section of the Technical Specifications is the desired response. Due to the misleading nature of this question, it is requested that it be deleted from the examination. Should this infomation be desired in the future, a multiple choice fomat might be appropriate.

Resp.

Not accepted. All candidates who asked received l

clarification.

Further all Technical Specification instrument surveillances are classified as a particular i

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9 type, eg; calibration, channel check, response time check, etc.

8.02.c Alternative correct responses to the reason for the required action to be taken are contained in the last paragraph of the basis for Technical Specification 2.15.

Some of the other reasons include: Preventing a limiting condition for operation on CEA insertion, CEA deviation, CEA overlap, or CEA sequencing from being approached.

Resp.

This comment is ignored as the question does not require the limiting event to be named and the above statements are the reasons for installing a rod block.

8.07.a In order to clarify the question, it is suggested that the second sentence be reworded as follows: What conditions eliminate the need for the requirement to disable the pumps?

Resp.

Noted.

8.08.a This question could be clarified by adding that each of the first two barriers start at the core and work out.

Resp.

Noted 6

8.10.c The question as written is in error. Technical Specification 5.11.1 addresses High Radiation Areas.

Technical Specification 5.11.2 addresses Very High Radiation Areas.

From the answer key it appears that the desired response was the methods required for High Radiation Area. As such, the candidates could have been mislead by the question and responded with the items listed in Technical Specification 5.11.2 Due to the confusion created, this question should be deleted from the examination.

Resp.

This comment is not understood in that a response concerning T. S. 5.11.2 was desired. However no change will be made to the key as t M requirements'of 5.11.2 are referenced from 5.11.1 and they are identical.

b.

MARCH 5, 1987 EXIT MEETING

SUMMARY

At the conclusion of the site visit, the NRC examiners met with representatives of the plant staff to discuss the site visit. The following personnel were present.

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-0 PPD OPPD J. Hannon R. Andrews J. Fluehr P. Harrell J. Gasper W. Gates J. Pellet L. Kusek D. Leather T. Westerman:

K. Morris D. Munderloh J. Whittemore J. Tesarek L. Wigdahl Mr. Whittemore started the meeting by expressing thanks for the cooperation received by the examiners during the site visit. The operations staff was singled out-for the cooperation of-the on-shift control room operators during examination administration. The following general topics were discussed.

(1) AsrequiredbyNUREG:1021,theExaminerSdndards, preliminary results are no longer provided in the exit meeting.

(2) NRC would attempt to finalize results as soon as possible, but grading could not commence until the licensees comments on the written exam were received in the RIV office.

(3) The licensee expressed concern that they had little or no control over the candidates selected to receive NRC requalification examinations.

It was explained that examinations were often cancelled or delayed due to personal or scheduling problems. It was also noted that the NRC has regulatory responsibility to ensure that individuals holding active operator licenses are in fact competent to operate the facility safely.

c.

GENERIC WEAKNESSES The generic weaknesses noted below were noted during the administration of the operating oral walk-through examinations.

They are provided for the benefit of the licensee and no response or corrective actions are required.

(1) Candidates could not recognize abnormal indications on the Qualified Safety Parameter Display System (QSPDS) and when abnormalities were pointed out, they could not propose corrective action.

(2) At least three candidates stated that boration would cause a noticeable shift in core flux axial profile shape because of higher boron concentration entering the bottom portion of the Core.

(3) Candidates were at a loss to explain why the core axial flux varied with changes in core power and lifetime.

11 (4) At least two candidates stated that a Radiation Work Permit could not be made_ active on back shift as the supervisory personnel needed to approve were not on site and would have to be called out.

(5) Knowledge of plant response to small reactivity changes at low power levels was weak in some individuals.

(6) Candidates were confused about the procedure to extend radiation exposure limits, some even to the point of inability to execute the opened procedure.

d.

EXAMINATION MASTER COPY Master copies of the SR0 license examination questions and answers follow this page.

e.

FACILITY EXAMINATION COMMENTS The licensee examination review comments as received by the NRC follow the examination master copies.

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S.

NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

_Elz_Q8LUQuN REACTOR TYPE:

_EWB-QE__________________

DATE ADMINISTERED: _QZLQ24Q2________________

EXAMINER:

_WBlIIEdQBgt_Jt__________

CANDIDATE:

INSIBVQIl0NS_IQ_QaNQ10aIEL Uno separate paper for the answers.

Write answers on one side only.

Stcple question sheet on top of the answer sheets.

Points for each qucstion are indicated in parentheses after the question.

The passing grcde requires at least 70% in each category and a final grade of at locst 80%.

Examination papers will be picked up six (6) hours after tho examination starts.

% OF CATEGORY

% OF CANDIDATE'S CATEGORY

__ VALVE _ _IQI8L

___SQQBE___

_V8LUE__ ______________CaIEQQBl_____________

_25zQQ__ _25tGQ

________ S.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS

_2510D__ _25100

________ 6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_25tDQ__ _25tDQ

________ 7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 2EAQQ__

2EAQQ

________ 8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100t00__

Totals Final Grade All work done on this examination is my own.

I have neither given nor received aid.

Candidate's Signature

5t__IBEQBl_QE_NUQLEaB_E0 WEB _ELeNI_QEE86110Nt_ELVIDSt_6NQ PAGE 2

IBEBdQQ1N6dIQ1 QUESTION 5.01 (1.00)

With the reactor critical at 10 EE-04 %,

rod withdrawal is used to increase power to 10 EE-03 %.

Select the statement that correctly describes the position of rods after the power is stabilized at 10 EE-3%.

c.

The rod position will be higher than at 10 EE-04% because more fuel must be exposed to the available neutrons to maintain the higher power level.

b.

The rod position will be higher than at 10 EE-04% to overcome the power defect.

c.

The rod position will be the same.

The outward rod motion needed to achieve a given startup rate equals the inward motion needed to reduce the startup rate to zero.

d.

The rod position will be lower than at 10 EE-04% due to the increased delayed neutron population associated with the higher power level.

(1.0)

QUESTION 5.'02 (2.00)

What are 4 of 5 automatic reactor trips which act to prevent the core from violating DNBR limits?

(2.0)

-QUESTION 5.03 (1.00)

During a reactor startup, an initial reactivity addition causes power to increase from 20 to 40 CPS.

A second reactivity addition causes power to change to 80 CPS.

Which of the following statements is correct?

e.

The first reactivity addition was smaller, i

b.

The second reactivity addition was smaller.

c.

The first and second reactivity additions were equal.

d.

There is insufficient data to determine a relationship between the reactivity additions.

(1.0)

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CATEGORY 05 CONTINUED ON NEXT PAGF *****)

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5t__IBEQBX_QE_NUQLEaB_EQWEB_EL8NI_QEEBBI1QN _ELVIDSt_aNQ PAGF 3

t IBEBdQQ1Nad103 QUESTION 5.04 (1.00)

The unit' is prepared for startup after refueling.

K eff 0.95 and count

=

rete is 10 CPS with all control rods inserted.

The operator pulls the chutdown banks fully out, adding + 3.5% dK/K of reactivity.

Without further operator action, nuclear power will: (Choose the correct response) c.

Continue increasing until the fuel or moderator temperature increases.

b.

Increase while rods are moving but will decrease by an observable amount as soon as rod motion stops.

c.

Increase while the. rods are moving, but decrease back to 10 CPS after rod motion stops.

d.

Increase until the positive reactivity addition is offset by the Isothermal, Temperature Coefficient.

o.

Increase to a new equilibrium count rate depending on K eff and source neutron level.

(1.0)

QUESTION 5.05 (3.00) a.

Explain how and why an individual CEA rod worth is affected by moderator temperature.

(1.0) b.

If the worth of an individual CEA rod is measured at a flux level of 10(EE-5) % power, how is the worth of the CEA affected if flux level is raised to 10(EE-4)% power AND there is NO Change in relative flux distribution?

Explain.

(1.0) c.

Beside temperature and neutron flux. what are two factors of design affecting the worth of an individual CFA?

(1.0)

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

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[ QUESTION 5.06

'(2.50)

After.a reactor startup, power is maintained at 50 %.

I c.

.What'will happen.to-Samarium concentration if power is increased to 100% after 2 months at 50%?

( 0.~ 7 5 )

b..

. Explain the initial effect on Xenon concentration as reactor power'is increased to 100%.

(1.0) l-L

c.

If shutdown from 100% power results in the insertion of-approximately

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0.005'dK/K of negative reactivity due;to the buildup of. Samarium, how I:

much reactivity would be inserted from shutdown or scram at 50% power.

Assume equilibrium Samarium conditions for both-cases.

(0.75) l

. QUESTION 5.07 (1.50) l

-How does an increase in RCS temperature affect.the relationship between indicated and actual core power as measured by the excore nuclear instruments?

Explain your answer.

l QUESTION ~ 5.08 (3.00) l Explain how and why the following factors affect the magnitude of the Doppler temperature coefficient over life.

Consider independently.

a.

Pu-240 buildup l

b.

Fission gas buildup i-c.

Fuel pellet swell d.

Fuel densification (3.0)

QUESTION 5.09 (1.00)

What are four common indications of centrifugal pump cavitation?

(1.0)

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CATEGORY 05 CONTINUED ON NEXT PAGF *****)

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'IBEBdQQ1NedIQS U

' QUESTION 5.'10

'(1.50)

LIndicate on the answer sheet whether the following statements are'true or

< f else:'

a.

Pump runout is a term used to describe a centrifugal-pump operating 4

withLit's-discharge flow totally restricted.

(0.5) j:

b.

A-law applying to the' operation'of centrifugal pumps.in closed loop systems states that the power required:is directly proportional to'the square of the pump speed.

(0.5) g E

'c.

For two IDENTICAL centrifugal pumps' running at DIFFERENTispeeds.in l IDENTICAL systems, the' faster. running pump requires a greater'NPSH to prevent cavitation.

(0.5) 1 i

l '

. QUESTION 5.11 (1.50)

Considering ONLY condensate depression;in the condenser, how will it's-

.vorience effect the REQUIRED and the AVAILABLE Net Positive Suction Head

+

i (NPSH) of.a condensate pump? VERY BRIEFLY EXPLAIN.

(1.5)_

?

i QUESTION 5.12 (2.00) 6 e.

What TWO parameters can the operator control to prevent pressurized thermal shock?

(0.5) i b.

What_is the sequence of events that could lead to pressurized thermal shock conditions?

(1.0) i c.

Why does the concern about brittle fracture of the reactor pressure vessel increase as the plant ages? Include in your answer the specific material property that is affected.

(0.5) i i

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CATEGORY 05 CONTINUED ON NEXT PAGF *****)

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. QUESTION

'5.13 (3.00)-

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..x Would' fuel center:line temoersture INCREAhE, DECREASd. or hEMAIN THE.SAME in each of. the f trlIowiig situationa?

BRIEFLY EXPLAIN.

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a.

Power decreases with" constant Ta've.

.(0.75) x.c

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(0.75) b..

Tave. increases with constant-power.

g 0.75)

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c..

Core ~ age increases with constant power.

=d.

Pressurizer pressure increases with.const&nt power.

(0.75).

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' QUESTION 5.14 (1.00) 9 WhichTof the following will,cause plarat effidiency to increase?

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a.

Total S/G blowdown is changed.from 30 gpq to 40 gpm..

b.

Steam quality changes from 99.7% to 99.9%.

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c c.-

Level increase doihigher than-normal in a feedwater hedter.

I d.

Absolute conde'nser pressure changes from 1.0 psi to 1.5 psi.

(1.0) l' I

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

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. UESTION 6.01

-(2.50)

Q 3.

0escribe Nqw the.following operate:to. manipulate the lineup or. change 3

.3-functions 4f the; Reactor' Protection System.

State. specific setpoints.

r l-1.

Wide Range. Log Channel.

(0,5).

12.

Power Range Safety. Channel.

-(0.75)

N:

b.'

Describe how the Nuclear InstrumentJSystem (NIS) functions to alert the Loperator of a dropped CEA.

(0.75) n c.. How is reliability improved by buffering-NI channel outputs?

.(0.5)

. QUESTION 6.02-(3.'00) c.

Describe the 3 types of " Load Shedding" designed into the Engineered

-Safeguards Control. system and state-the situations under which they.

-4

~

'l would occur.

T

-(2.25)

'b.

What are the CONDITIONS required to trip the Offsite Power LowLlockout' -

relays and what? ACTION results from-. relay trip?

(0.75) 1 QUESTION 6.03 (3.00) t

' State.the power pupplies available to a Safety Related 120VAC

.(

e.

Instrument PoweriInverter units and indicate the normal source.

(0.75) s b.-

What are two met fds by which the units can be switched to the other

'(alternate) power Mupply?

(0.5) 4 W.

'How many and what type (Safety /Non-Safety) of 120VAC Instrument Power inverters would be directly impacted by the loss of a single 125VDC

.a bus?

(0.75)

.a d.

Besides the attributes discussed in a.

and b.

above, what other design

. feature adds [ reliability to the power supply for the Safety related 120VAC Instrument busses?

(1.0) gu

.s i

i L

[

D L

(~

(***** CATEGORY 06 CONTINUED ON NFXT PAGF *****)

c a

s sk 6t__EL6NI_SISIEdS_ den 1 Gut _CQUIBQLt_880_INSIBudENIeI1QU PAGE 8

QUESTION 6.04

+(3.00) i 0.

What ' two conditions are necesnary to IN.TI ATE the automatic ramped closing feature of the Feedwat6r Regulat'ing Valves (FRVs)?

(0.5)

i b.

What are two ways that the operator can stop the " Ramp Function", once it has been initiated?

(1.0) c.

Describe and explain the response of the FRV to a high downcomar level in the Steam Generator.

Describe the transient for an extended period with no operator action.

(1.5)

QUESTION 6.05 (2.50) a.

Why is it that the Turbine Driven Auxiliary Feed. Pump is usually stopped and placed in standby after a transient stabilizes while the motor driven pump is used to maintain level?

(1.0) n i

b.-

What are the two auxiliary feed water flow paths to the steam generator

\\

rh

.and why are two flow paths necessary?

(0.75) c.

Why is tt recommended that the operator maintain a minimum S/G level when adding auxiliary feed via the normally designed flowpath?

(0.75)

\\

QUESTION 6.06 (2.50) j c.

State the normal range of the Area Radiation Monitor indicators and describe the options available to the operator for reading these indicators (not recordurs).

(1.0)

b.

What is the basis or reason for the low setpoint alarm on Area Radiation Monitors?

(0.5) c.

What are the four r e le ase., po ints where;potentially-radioactive gaseous effluents are monitored by the crocess RMS?

(1.0)

E 4

(*****

CAY 2 GORY 06 CONTINUED ON NEXT PAGE *****)

m ;-

ti__EL8NI_111IEdi_QE11GNt_CQNIBQLt_eUQ_IN11BVdENIoI100 PAGE 9

QUESTION 6.07 (2.50) e.

Explain how Raw Water (RW) system interface valves.use 2 solenoid valves in their control and. operation. Be specific ~as to generic power.

j supplies, interlocks, switches, and locations for control.

(1.5) b.

What RW system design feature assures that the system can be properly-i aligned during accident conditions.in the event of a loss of normal instrument air and what are the specific components affected?

(1.0)

-QUESTION 6.08 (2.50) e.-

What are.three uses of the connection between the discharge side of the charging pumps and the SI redundant high pressure header?

(1.0) b.

What is the: purpose of-the connection between the Saf ety.Inj ection Tank recirculation line and the CVCS?

(0.5) c.

Why are there connections between:

1.

The" Spent Fuel Pool system and the SIRWT?

(0.5) 2.

LThe SI system and the Spent Fuel Pool System that are installed when needed?'

(0.5)

QUESTION 6.09 (1.50) a.

What.are 2 parameters (not breaker features) that will trip the charging pumps AND what condition (s) will override these trips to restart the' pumps?-

(0.75)

. hat are the:three methodologies employed in the CVCS system to b.

W reduce the Reactor Coolant System activity prior to cooldown?

(0.75)

(*****

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

Et__ELoul_11SIEUS_QE1100t_QQUIBQLt_800_1GSIBUDEUIBI196 PAGE 10 QUESTION 6.10 (2.00) a.

Explain WHY the following statement is TRUE or FALSE:

The only time and reason the operator uses Pressurizer Vapor Space RTO-indication is when the pressurizer is full to determine the upper portion water temperature.

( 1. 0 ).

b.

Detail the protective action signals that may be initiated by Pressurizer Pressure.

(1.0) 1 l

t

(*****

END OF CATEGORY 06 *****)

i

Zz__EBQQEQuBES_:_NQBdeL&_8BNQBd8Lt_EdEBGENQ1_6NQ PAGE 11 88010LQQ108L_QQNIBQL-QUESTION

'7.01 (2.50) e.

Describe-how it-is ascertained that it is. safe to release a. fuel-assembly that has j ust.been CORRECTLY installed in the Reactor. vessel-during refueling,Eend commence loading the next assembly.

Be specific as to techniqueoof-determination.

(1.0).

b.

What two specific communications systems.or portions of systems may te-used for refueling operations?

(0.5)

c.. List 5 of the 7 alarms listed in the fuel handling incident procedure-(AOP-8) that could possibly indicate a fuel handling incident has.

occurred.

(1.0)

QUESTION 7.02 (2.50) a.

. Precautions.in OP-3-(Start up to minimum load)-AND OP-4 (Power operation, load : changes) state that Rate of power change and CEA movement shall be restricted to that specifiedJin the Technical Data Book.

What are the 2 bases of those limits found in the Technical Data Book?

(1.0)

. hat'are the only 2 considerations for limiting CEA insertion during W

~

b.

power operation?

(0.5) c.

During the initial star. tup of a newly refueled core, why may the Reactor Engineer place limits on CEA withdrawal?

(1.0)

QUESTION 7.03 (2.50) a.

Why is the operator required to use correction graphs during cooldown and/or accident conditions to ascertain and control the pressurizer level?

Provide a specific answer for each situation.

(1.5) b.

Explain how automatic safeguards protection is initiated if a LOCA occurs during a normal plant cooldown with RCS pressure at 1500 psig and RCS temperature at 460 deg's.

(1.0) l

?..

(*****

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

- Z t. _ BQQEQV8El_:_UQBU6Lt_8EUQ808Lt_EDEBQEUQ1_660 PAGE 12

)8Q1QLQQ1QaL_QQNIBQL QUESTION 7.04 (2,50)

Dcscribe the LAST RESORT methods, including flowpaths, for removing residual heat subsequent to a complete loss of Shutdown Cooling for the following. conditions.

c.

Reactor head installed.

(1.25) b.

Head removed with refueling conditions established.

(1.25)

QUESTION 7.05 (2.50)

According to A0P-23:

G.

Why is the operator cautioned not to terminate containment spray by closing the spray valves when RAS has been initiated?

(1.0) b.

With forced circulation, what are 2 of the 3 conditions that will verify to the operator that adequate core heat removal exists prior to securing safeguards equipment?

C0.6) c.

What are 3 of 4 conditions the operator must confirm prior to stopping or throttling a safety inj ection train?

(0.9)

QUESTION 7.06 (2.00)

During recovery from an accident it becomes necessary to vent the RCS of non-condensible gasses:

c.

What are the 4 different vent paths available for venting?

(1.0) b.

Briefly describe how the operator will determine the duration of the venting period for any given vent path.

(1.0)

(*****

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

-Zz__EBQQEQUBES_:_UQBdakt_aBNQBd6Lt_EdEBGEUQ1_aNQ PAGE 13.

88DIQLQQIQaL_QQUIBQL QUESTION 7.07 (3.00) c.

What does a star (*) preceding a step in the Emergency Operating Procedures CEOPs) indicate to the operator?

(0.75) b.

What is the operators responsibility if acceptance criteria for Reactivity Control is not met upon completion of the Functional Recovery Procedure?

(1.0) c.

According to the Resource Assessment Tree for: " MAINTENANCE OF VITAL AUXILIARIES", what are the 4 initial success paths in order of priority?

(1.25)

QUESTION 7.08 (1.00)

Select the correct answer:

According to FC Emergency Procedures, once the operator has satisfied all Sefety Function criteria, he or she should; o.

Perform a-Safety Function Status check.

b.

Return to the appropriate E0P 01 thru 06.

c.

Proceed to Long Term Action.

d.

Restore backup vital auxiliaries.

(1.0)

QUESTION-7.09 (2.50)

According to E0P-02, Electrical Emergency, for a loss of all AC power:

a.

Describe the prescribed method for the operator to remove core decay heat.

(1,0) b.

Why is the operator cautioned not to cooldown below No-Load T-Avg?

(0.5) c.

What are 3 of the 4 steps the procedure dictates in an attempt to prolong the useful availability of the DC power source?

(1.0)

(*****

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

~

sZr__EBQQEQUEE1_:_NQBdakt_8BNQBdekt_EdEBGEUQY_6ND.

PAGE 14

'88010LQQIQaL_GQUIBQL

' QUESTION 7.10

- ( 2.' 5 0 ) -

~

During.the-performance of EOP-4, Steam Generator-Tube Rupture:

a.

What are 2 considerations specified far determining where RCS pressure should'be controlled?

(1.0) b.

What can be done to prevent overfilling of the affected S/G?-

(0.5) c.

How can the'affected S/G be cooled down after'it has been isolated'with natural circulation in progress?

(0.5)

P Dd. 'What problem should be suspected if difficulty is encountered in-reducing pressure to place'SDC in' operation during natural circulation cooldown?

( 0. 5)'

s

. QUESTION 7,.11 (1.50)

Briefly indicate the specific action necessary to avoid having to' shutdown the unit should it be discovered-that the SIRWT has been inadvertently

- diluted below its minimum boron concentration.

(1.5)

(*****

END OF CATEGORY 07 *****)

-Ai__6DMINISIB611YE_EBQQEQUBEnt_Q0UQ11106Ht_680 L10116I1QN3 PAGE 15 QUESTION. 8.01 (3.00) a.

. Explain why.the following. leakages are.IN or OUT of compliance with the Limiting Condition for Operation.-

Consider each individually and-separately.

.1.

Unidentified----------------------------------- 0.65 GPM 2.

Steam Generator Tube Leakage:

A----------------------------------------------0~.45 GPM B----------------------------------------------0.50 GPM 3.

Various RCS manual vent and drain valves seat and packing gland leakage.---------------------4.6 GPM 4.

Back leakage through SI: check valves detected by leak check on previous shift.----------------3.9 GPM (1.5) b.

Evaluate compliance if the above leakage is simultaneous.

(0.75) c.

Fort Calhoun Technical Specifications require that at least.one system or method be operable.for detecting leakage to the containment AND-one for detecting leakage to the secondary system.

1.

State 3 systems or. methods used'to detect leakage to the containment.

2.

State 2 systems or methods used to detect leakage to the secondary-system.

(0.75) i

. QUESTION - 8.02 (2.50) n.

What Technical Specification instrument surveillance (s) will include the adj ustment of any alarm, interlock, or trip setpoints?

(0.5) i-b.

.Why do the Technical Specifications state that at rated power the minimum operable high power level channels is three (3) and what must be done immediately if the number.of operable becomes two (2)?

Explain the reason for the required action.

(1.0) c.

What must be done to continue operation if the " Rod block" system should become inoperable for an extended period of time due to a malfunction of the secondary CEA position indication system?

Explain the reason for-the required action.

(1.0) c.

(*****

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

r n

w-

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

.-----.c

,,~.-..

-.~,--,---,..+.---n.

Aia_aQUIN11IB6I1YE_EBQQEQMBEft_QQNDIIIQNic_6NQ_L15116IIQNh

'PAGE 16

,QUESTIGN-8.03

.(2.00)

' Motch: one =def inition to each term.

TERMS.

OEFINITIONS s.:

1.25 Rem /Qtr.

1.

The 10 CFR 20 maximum allowable annual dose not to exceed-5(N-18) b.

400-500 Rem or 3. Rem /qtr with-exposure documented.

c.

Chronic exposure 2.

Median lethal exposure-for man.

.d.

12_ Rem.

3.

The 10 CFR 20' Dose limit for the whole body in a restricted area.

4.

Small exposure rate.over a long period of time.

5.

Radiation effects transferred-from

. parent to child.

6.

One time limit for lifesaving.

.i (2.0)

QUESTION 8.04 (2.50)

Assume that an Operations Incident Report-Form.has been initiated on shift:-

e...Who (by title) exercises administrative _ control of the form before it

.goes into.the review chain?

(0.5) b.'

Who (by title) are the 2 individuals responsible for documenting event description AND immediate action pertaining to the event?

(1.0) c.

Describe the entire review chain for the completed report.

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

< At__8Dd1811IB611YE_EBQGEQuBEnt_GQNQ1110N1t_8ND_L10118IIQU$

PAGE 17 t

QUESTION 8.05 (2.003-l0.

Who is responsible for. preparation and promulgation of surveillance

testing' schedules?

(0.5)

'b.

What.is the mechanism for scheduling surveillance testing for

~ operability of safety related equipment that is undergoing repair?

( 0. 5).

What are 2 documents where test requirements deemed necessary by the c..

Technical Review Supervisor may be specified?

(1.0) 4!

QUESTION 8.06 (3.00)

-According to Standing Order G-28, Fire Protection Plan:

a.

Who is responsible for' making initial off site notification?

.(0,5) b.-

During backshift, who will be in charge at the scene?

(0.5) c.

What is the Shift Supervisor's responsibility during a fire on backshift?

.(0.5)._

d.

-I f a nuclear emergency develops during a fire,.where should non-operating and non-security personnelEassemble?

(0,5)-

'o.

Describe how the plant communications system is-utilized during fire fighting activities.

Also explain backup methods-in the event of normal system failure.

(1.0) 1-L-

(*****

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

K k

At__6Dd1NISIB611VE_EBQCEQuBESt_90NDIl10NSt_6NQ_ Lid 11811QNS.

PAGE-18

. QUESTION 8.07 (2.50).

l la.-

Protection against low temperature overpressurization requires disabling of the HPSI pumps.

What conditions mitigate the requirement-to disable the pumps with the plant ~in cold shutdown?

(1.0) 1

b.

. State'how many HPSI pumps must be disabled for the followingJRCS l

< co nd it io ns :

1.

115 deg's 2.

~335 deg's 3.

275 deg's 4.

250 deg's

5..

315 deg's (1.5)

, QUESTION-8.08 (2.50)-

a.

JO default method ~of Emergency Classification is the Barrier Method.

List 3 of the 4 symptoms that will indicate a challenge to or failure L

of EACH of the first TWO barriers.

List parameters or symptoms only and ignore setpoints.

-(1.5)

.b.

State thelbarrier failure / classification guidelines.

(1.0)

QUESTION-8.09 (2.50)

In the event of a bomb threat:

a.-

Who are the two individuals (by title), one of which MUST be notified by a person receiving a written or oral threat?

(0.6)

.b.

What 3 off site agencies are to be notified immediately?

(0.9) c.

. What are the' specific precautions to be observed by an individual finding-a written bomb threat?

The answer should include the reason the precautions are observed.

(1.0)

(*****

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

e x

JAz__6Dd1N1118611YE_EBQQEQUBESt_CQNQ1Il081t_6NQ_ Lid 118I19NS

'PAGE 19 UQUESTION '8.10 (2.50)

>av What are:the NORMAL 0.P.P.~D., Permissible Accumulated Dose Limits for whole body penetrating radiation?

(1.0) b.'

If maintenance is to be performed in the containment at' power over a-period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ~and will likely result in an exposure of more than 100 mrem, who are the'3 individuals (by title),.'any-2 of which are required.to approve the RWP, not including the Shift Supervisor?L(0.75)

L c.

Accoroing to' Technical Specifications 5.11.1,-Administrative controls, an. individual may be allowed to enter a VERY HIGH RADIATION AREA-provided.that 1.or all of 3 different methods or procedures.of protection are. utilized.

What are/the 3 methods or procedures of concern?

(0.75)

)

(*****

END OF CATEGORY 08 *****)

(*************

END OF EXAMINATI0t: ***************)

~5t__lBEQBl_QE_UMGLE68_EQWEB_EL6dl_QEE86I19Bt_ELMlQSt_60Q PAGE 20 IBEBdQQldedlQS ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE, J.

ANSWER 5.01 (1.00) c.

Rod position will be~the same.

(1.0)

REFERENCE Bcsic Nuclear Physics ANSWER 5.02 (2.00) 1.

High power 2.

Low RCS flow 3.

TMLP 4.

Axial Power Distribution Trip 5.

Assymetric Steam Generator Transient Trip (differential pressure trip)

(Any 4/5, 0.5 ea.]

(2.0)

REFERENCE Technical Specification 1.3 ANSWER 5.03 (1.00) b.

The second reactivity addition was smaller.

(1.0)

REFERENCE CE Trng. Ctr. Rx Theory Notes, p.139 l

ANSWER 5.04 (1.00) o.

Increase to a new equilibrium count rate.

REFERENCE CE Trng. Ctr. Rx. Theory Notes, Pp.139 - 144

L5t__IBEQBY_QE_NVQLE88_EQWEB_EL8NI_QEEBoI198t_ELUIDSt_6HD PAGE 21 IBEBdQQ18851DS ANSWERS'-- FT.:CALHOUN

-87/03/03-WHITTEMORE,1J.

. ANSWER 5.05=

(3.00) a.

Worth will increase with an increase in moderator temperature [0.,5) as. thermal-diffusion length will increase.~ [0.53 CONCEPT.

(Allow 0.25 credit for explanation of boron expanding out of core.)

(1.0)

-b..

No change. [0.5]

The: absolute value of neutron. flux will-.not change the worth.

A shift in flux distribution is required. [ 0.~ 5 ]

(1.0) c.

CEA size,-CEA absorber material, CEA location. [any 2, 0.5 ea)

(1.0)

. REFERENCE.

CE Trng C t r ' fbc Theory Notes, Pp. 181-183 ANSWER 5.06

-(2.50) n.

Nothing. [ Accept: Small dip and-return.to equilibrium.]

(0.75) b.

Xenon concentration initially dips (0.5] as increased flux results in increased burnout, [0.253 but the precursor (Iodine).does not immediately increase to the new level. [0.25] CONCEPT '

(1.0) c.

Approximately 0.0025 dK/K.-

Accept (0.0024 - 0.0026)

(0.75)-

-REFERENCE CE1Trng. Ctr. Rx. Theory Notes, Pp. 204, 214, 215-ANSWER 5.07 (1.50)

As RCS temperature increases indicated power reads higher.than actual. [0.75]

This is due to increased fast neutron leakage due to a decrease in coolant density. [0.75)

REFERENCE CE Trng Ctr R-Theory, p 166 4

e s

Ex _IBEQBl_QE_NMGLE68_EQWEB_EL8NI_QEEB6110Nt_ELVIDSi_eUD

_PAGE L'22 IBEBdQDINad101 LANSWERS -- FT.'CALHOUN

-87/03/03-WHITTEMORE,-J.

ANSW'ER 5.08 (3.00) a.'

Increase (0.25) More resonance peaks available to absorb neutrons' (0.'5) :

con'uctivity decreases, fuel temp increases, more b.

Decrease-(0.25) Gap d

-peak overlap'. So, Doppler' decreases. (0.5)

~

c.

Increase (0.25) Gap decreases, fuel temp = decreases, less peak overlap.

So,-Doppler increases'. (0.5)

Ld.

Decrease (0.25) Gap increases,' fuel temp increases,_more peak-overlap.

So, Doppler decreases (0.5)

(3.0)

REFERENCE CE Trng Ctr-Rx Theory, Pp. 159-170 JANSWER 5.09 (1.00)

REFERENCE Wast. Thermal Hydraulic Principles and-Applications.to the PWR II, p.

10-54 ANSWER 5.10-( 1.- 5 0 ) -

e.

False b.

False c.

True (0.5 ea.]

(1.5)

+

REFERENCE-Westinghouse Thermal Hydraulic principles, Pp. 10 - 36,37 2

b l

st__IBEQBl_QE_U9QLEeB_EQWEB_EL6NI_QEEBoI1QNt_EL91Q1t_68Q PAGE 23 IBEBdQQ1Nad1CS ANSWERS -_ FT. CALHOUN

-87/03/03-WHITTEMORE, J.

ANSWER 5.11 (1.50)

Required: No change [0.5) since required NPSH is a design criteria or foature of the pump. [0.25]

(0.75)

Actual NPSH: Increase [0.5) since temperature of the fluid is docreased.[0.25]

[0.75)

REFERENCE

. General Physics HT&FF p.

182 ANSWER 5.12 (2.00) c.

Coolant temperature Coolant pressure (0.25 ea.]

(0.5) b.

Rapid cooldown and depressurization (0.5] followed by rapid repressurization (0.5).

(Accept any viable sequence.)

(1.0) c.-

Neutron exposure (integrated) [0.25] makes the material more brittle (raises NDTT) [0.25).

(0.5)

REFERENCE C-E Concept of Fracture Analysis ANSWER 5.13 (3.00) c.

Decrease (0.25), smaller delta T required to transfer less energy to RCS [0.5).

(0.75) b.

Increase [0.25), center line temperature responds to RCS temperature in order to maintain constant delta T across cladding [0.5).

(0.75) c.

Decrease [0.25), fuel swelling and clad creep reduce clad gap which reduces delta T across the gap and lowers the center line temperature (0.5).

(0.75) d.

No change (0.25), pressure has little effect on heat transfer in subcooled fluids [0.5).

Accept increase if the assumption is stated that increasing pressure decreases nucleate boiling.

[0.75)

l 5t__IBEQBl_QE_NMCLEoB_EQWEB_ELeNI_QEEBeI10Nt_ ELM 10st_6ND PAGE 24 IBEBUQQ1Ned1GS ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE, J.

REFERENCE CE Trng Ctr Thermal Hydraulics, Pp. 18-27 ANSWER 5.14 (1.00) b.

Steam Quality increases.

REFERENCE Goneral Physics, Heat Transfer Thermodynamics and Fluid Flow, pp. 145 - 160.

'd r

+16t__5L6NI_111IEd1_QE11GNi_QQNIBQLt_68Q_IN11BudENI6IIQU PAGE-25-1 ANSWERS.- :FT.-CALHOUN

-87/03/03-WHITTEMORE,-J.

l 9

ANSWER' 6.01 (2.50)'

a..

1.

Removes zero. power bypass > 10 EE-4,% power.

Disable rate'of change output'< 10 EE-4% power.-

2.

Disable rate of change output ) 15 % power.

Enable APD trip at 15 % power.

Enable Loss-of Load' trip at 15 % power.

[0.25'ea.]

(Accept converse of above)

(1.25) b.

Power Range' Safety Channel.present output is compared to a' time de-layed previous output to' develop and output a "CEA Dropped" signal.

(Accept other viable NIS indications'for no credit.)

(0.75)-

c '..

Prevents any other signal from being affected when an incident

signal is shorted or connected to a power supply.

-(0.5)

. REFERENCE SD II-6, Pp 4-7 9

ANSWER 6.02 (3.00) 4 a.

-1.

Accident (or 480V) Load Shedding [0.253 sheds selected loads and.

results from SIAS' signal. [ 0. 5 ]J 2.

Undervoltage Shedding (0.53' sheds its loads when the bus relay senses-degraded voltage. [0.25]

3.

' Backup Load'Shed signal [0.5] occurs when all-feeder breakers to the affected bus are sensed to be open. [0.25]

(2.25) b.

Conditions:

Degraded voltage [0.15] AND-[0.15] SIAS signal. [0.15]

l Actions:

Opens incoming line breakers, [0.15] starts and loads diesel generators. [0.15]

(0.75)

REFERENCE SD II-7, Pp34-37 e

t a

-..wg y

.-e 7--,,,,.-w, yg,_

,w,,,

mym,

,.--m, e

,a

-, - ~,,,.,,,

,-,w..

can_,

g.e

.,,.r re,,-,-,-,-.,-m~ -,

kz__EL8BI_2XSIEUS_QE1196t_QQUIBQLt_800_IU11BudEUI611QN PAGE 26 ANSWERS

--'FT.

CALHOUN

-87/03/03-WHITTEMORE, J.

ANSWER 6.03 (3.00) c.

Normal is 125VDC from DC bus, [0.51 alterr. ate is AC via 480/120VAC regulated transformer. [0.25]

(0.75) b.

There are manual and static Cauto) transfer switches.

(0.5) c.

Two Safety related [0.5] and one non-safety related. [0.25]

C0.75) d.

Each inverter output is arranged so it can supply one bus besides the bus to which it is dedicated.

( 1. 0 )-

REFERENCE SD II-1, Pp. 36,37 ANSWER 6.04 (3.00) c.

Turbine tripped, FRV control in automatic, or AUT0/ INHIBIT (A/I)

Switch in AUTO.

[any 2, 0.5 ea.]

(0.5) b.

Depress the ramp stop button, reset the turbine, or position the A/I Switch to INHIBIT.

[any 2, 0.5 es.]

(1.0) c.

The valve will shift to manual and close. [0.5)

When level decreases to clear the high level condition, the valve will reopen to its previous position. [0.5]

Level will be controlled between the high level trip and reset points. [0.53 (1.5)

REFERENCE SD III-3, Pp. 43,44

6t__EL6NI_11SIEUS_DESIQut_QQUIBQLt_600_IN11BydENI611QN PAGE 27 ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE,-J.

ANSWER 6.05 (2.50) c.

The motor-driven pump runs at a constant speed and will maintain a (shutoff head of about 1200 psi).[0.25]

The turbine pump control system attempts to keep its discharge pressure about 40 psi above steam pressure.[0.25]

Therefore if only one pump flow or less is required the Turbine-Driven pump will be idling anyway.[0.5]

(1.0) b.

The auxiliary feed nozzles [0.25] and the normal feedwater header

[0.25].

.Necessary to ensure maintenance of heat sink in the event of aux. feed line rupture. [0.25]

(0.75) c.

(To keep the aux. feed nozzles submerged) minimizing thermal shock

[0.375] and avoiding water hammer. [0.375)

(0.75) l REFERENCE SD III-4, Pp.19, 20 l

l ANSWER 6.06 (2.50) c.

The normal range is 8 decades (0.1 to 10 EE 7 mr/hr)

[0.53 and the i

operator has the option of reading a scale of 8 decades or any 3 consecutive decades. [0.5]

(Accept setpoints for no credit)

(1.0) b.

To warn that the dose rate has reached an abnormal level that is still safe.

(0.5) c.

1.

Aux. bldg. vent. exhaust 2.

Cont. atmosphere purge exhaust.

3.

Cont. hydrogen purge exhaust.

4.

RWDS waste gas release.

[0.25 ea.]

(Allow full credit for any sources monitored prior to release)

(1.0)

REFERENCE SO II-12, Pp. 1-3 L

l

kz__PL6NI_$l$1EUS_QEg1QN _QQUIBQLt_6NQ_INSIBydENI611gN PAGE 28-t

)

ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE, J.

ANSWER 6.07 (2.50) a.

Solenoids are installed to require deenergizing both to open the interface valve. [0.25]

Primary and secondary solenoids are powered from opposite DC busses.

[0.25]

Secondary solenoids powered through.

pressure switches which open when CCW pressure decreases

[0.25) and controlled by switches on ASP.

[0.25]

The primary solenoids have manual control only from MCR. [0.25]

The operator obtains control of primary solenoid if ASP switch is in auto and CCW pressure below setpoint or the switch is in off. [0.25]

(1.5) b.

There are backup accumulators available [0.5) that can be used to operate the RW pump discharge valves and the discharge header cross-connect valves [0.25) and RW/CCW interface valves. [0.25) (1.0)

REFERENCE SD III-8, Pp. 20 23, ANSWER 6.08 (2.50) a.

The connection is used to test the operation of SI check valves, correct boron concentration in or fill the SI tanks, act as an alternate charging or inj ection path, Hot Leg Inj ect ion, or backup aux. pzr spray.

[any 3, 0.33 ea.]

(1.0) b.

To collect leakage from the SIT check valves; OR To accommodate flow during testing of SIT check valves. [ Accept either answer)

(0.5) c.

1.

To transfer refueling water between the SIRWT and the refueling cavity. (Allow 0.25 for: Cooling in the event of SDC loss)

[0.5) 2.

To allow the Shutdown Cooling Heat exchangers to be used as alternate or backup spent fuel pool cooling.

(0.5)

REFERENCE SD 1-9, P.

32

..=

i#

k 0

Lht__EL6HI_1111EdS_QE11GNt_QQUIBQLt_6NQ_INSIBydENI611QN PAGE 29 ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE, J.

ANSWER-6.'09

'(1.50) s

.a.

.The'-pumps willitrip on low suction pressure,C0.25] or lube oil pressure (0.25).but the trips are overridden'when the: pumps are started by the-sequencers. (0.25]

(0.75)

b.

1.

Degassification 2.

Filtration 3.

Demineralization CDe-ionization)

(0.25 ea.).

(0.75)-

ic REFERENCE SD'I-5, Pp. 21, 22 N

[

-ANSWER 6.10 (2.00) a.

_ FALSE (0.5]

This RTD can be used to compare actual steam temperature with saturation steam temperature to detect non-condensible gasses in the vapor space..CO.53 (1.0) i b.

'1.

Reactor trip on high pressure 2.

Open PORV 3.

Reactor trip on low pressure (IM/LP) 4.

Initiate SI (4 reg'd at 0.25 ea.)-

(1.0) i REFERENCE SD I-4, Pp 51,52 i

1 I

l l

t

Zi__PBQQEQyBER_ _NQBd8ki_8EUQBd6Lt_EdEBQENQ128NQ PAGE 30

'88Q19LQQ198L_GQNIBQL JANSWERSJ-- FT. CALHOUN

-87/03/03-WHITTEMORE, J.

~

ANSWER

'7.01 (2.50) 6.

. Perform a minimum of 2 (100 second) count rates (0.5] on 2 neutron

' indicating channels..[0.5]

(1.03-b.-

Channel 4 of. the Gaitronics system (0.'25] or the sound-powered phone system. [0.25]

(0.5) y c.'

1.

Area monitor 2.

.CTMT air - part iculate ' (RM-050)'

3.

-CTMT gas-(RM-051) 4.

Main stack gas (RM-62) 5.-

Stack _ gas iodine (RM-60) 6.

Stack gas particulate (RM-61)

-7.

Ventilation' isolation actuation- (VIAS) Cany 5.

O.'2'ea.)

8.

Stack rad. monitor (1.0)-

REFERENCE

~0P-11i Ppl3,14 & LICENSEE QUESTION BANK.

ANSWER 7.02-(2.50) e.

Turbine loading limits, maintenance of ASI, or fuel conditioning limits.

(Any 2, 0.5 es.)

(1.0) b.

POIL and ASI

[0.25 ea.]

(0.5)

c.. To maintain boron concentration low enough to preclude problems with-Moderator Temperature Coefficient. (CONCEPT)

(1.0)

REFERENCE TOB_III-14,.III-22,

~

,Zi__EBQQEQVBEE_ _NQBd6Lt_6ENQBd8Lt_EMEBGENQ1_8ND PAGE 31

-86DIQLQQIQ8L_QQUIBQL ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE, J.

l l

' ANSWER 7.03 (2'.50) a.

During cooldown, indicated pressurizer level may differ from actual due-to changing density conditions in the pressurizer (as well as the-varying weight of the vapor space). [0.75]

During an accident, changing ambient conditions.may cause density changes in the detector reference leg, causing indicated level to very from actual level. [0.75]

(1.5)

'b.

Protection signals for SI, Cont'Isol., etc. will be' initiated by.

Containment pressure signals.

[0.5] _ Operator _will normally place

~

PPLS Blo'ck Sw. to Emer. Reset to allow auto initiation. [0.5]

(1.0)

' REFERENCE-

-T.D.B.

III-1, 01-RC-4, Pp 4,6 ANSWER.

7.04~

(2.50) s.

Head installed:

Establish a flow ~ path from the SIRWT, [0.25] through the HPSI pumps,

[0.25] into the primary system, [0.25] and out of the PORVsgor loop drains to the RCDT.[0.5]

(1.25) b.

Head removed:

Align fuel transfer canal drain-pumps to take suction from SIRWT ((0.5]

and discharge to SFP. [0.25]

Align SFP pumps to take suction from SFP

[0.25] and discharge to SIRWT. [0.25]

(1.25)

REFERENCE AOP-19, P.3

Z1__EBQQEQUBE1_:_NQBdakt_6BNQBdaLt_EMEBQENQX_680' PAGE 32

'88Q10LQQ108L_QQNIBQL ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE,.J.

ANSWER 7.05

.(2.50)

-e..

Because an RAS signal will close SI pump recirc. valves,-[0.53' causing the pumps to deadhead. [0.53

(1.0) b.-

1.

.At least 1 RCP running 2.

Core delta T < 10 oF 3.-

RCS-subcooling > or'= '20 oF.

(any 2, 0.3 es.]'

(0.6) c.

1.'

RCS subcooling > or = 20 oF 2.

PZR lavel > 45%, and not decreasing 3.

At least 1 S/G available for heat removal 4.

RVLMS indicates core covered.

(any 3, 0.~3 es.]

(0.9)

I REFERENCE t

AOP-23, Pp2,3 l-k' I

ANSWER:

7.06~

(2.00) e.

Pressurizer--------------

quench tank Pressurizer--------------

containment Rx.. vessel--------

• ----- quench tank x

Rx.

vessel---------------

containment (0.25 ea.]

(1.0)

'b.

Pressurizer: The procedure. specifies duration-dependant on size of

. bubble in pzr. (Determined by use of graphs.]

[0.5]

,R x. v e s s e l : Venting duration determination will be by trial and error.

OR. Use of the vessel level monitoring system is.

(Accept either answer.)

[0.53 (1.0)

REFERENCE 01-RC-12, P.

4

Z1__EBQQEQUBE1_:_UQBueLi_eauQBuett_EUEBGEUQX_6NQ PAGE 33 BeDIDLQQ10eL_QQUIBQL ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE, J.

ANSWER 7.07 (3.00) c.

That the particular step may be performed whenever applicable.

(0.75) b.

May pursue other urgent Safety Functions, [0.5) but must continue to attempt to establish reactivity control. [0.53 (1.0) c.

1.

Recovery of AC busses 2.

Recovery of DC busses 3.

Recovery of instrument air 4.

Recovery of component cooling

[0.25 ea.,

0.25 for correct order)

(1.25)

REFERENCE E0P 20,PP. 25, 38 ANSWER 7.08 (1.00) c.

Proceed to long term action.

(1.0)

REFERENCE FORT CALHOUN E0Ps ANSWER 7.09 (2.50) a.

(Locally) operate the Steam Driven AFW pump, [0.5] while manually operating HCV 11078 and 1108B [0.25] or allow MS safeties to maintain S/G pressure. [0.25]

(1.0) b.

There is no make up available and voids will form when pressurizer level decreases due to RCS temperature decrease.

(0.5) c.

1.

Deenergize computer inverters 2.

Stop DC Emergency Lube Oil Pump 3.

Deenergize 2 120VAC vital bus inverters.

4.

Stop DC Emergency Seal Oil Pump 5.

Secure selected emergency lighting (Any 3, 0.33 ea.]

(1.0) l REFERENCE l

E0P-02, Pp 4, 10 l

l

\\

'l F m

.T /

t

,4

4. <

g.

s Z t _ _ E B Q Q E Q L!B E E _:_N Q B d o k t_8 B N Q BU 6 L4_EU M Q E U R Y _ o N Q PAGE 34 88Q10LQQ196L_GQUIBQL J'

T

\\,'

s A

s V

i ANSWERS -- FT. CALHOUN

.<87/03/03-WHITTEMORE, J.

s, 1:

h g

s s

s 1.

t,

-s s

l A )

ANSWER 7.10 (2.50)

I

' [k '

l.

c.

1.

RCP NPSH 2.

RCS Subcooling 1

S/G prek)ur,e) 3.

Minimizing leak rate ('RCS <x100 psig >

(any 2, 0.5 eat-)'

W(

(130)

\\

n +

( 0. 5 ) \\,

b.

Accept either* Drain to RWDS OR. Dump steam.

,x'

\\t.

g. 3 7 s,

c.

Feed and bleed (or drain ahd refill) usi(og' feedwater m.'d Stpam

'f t, \\

[0.5)

Generator blowdown.

d.

Voiding in RCS (Head)

(0.5) s-REFERENCE EOP-4, Pp 7, 10, 12

(

4 v

y g

)g

)

1 ANSWER.

7.11 (1.50)

(

f y

i U

+

While recirculating the tank using the SI pump recire lines.>Ct.5) adg

{

boric acid via the SIRWT acceso plug, (0.5] and confirm increosp with

semple. (0,5) p.

7_

(1.5) g t

i'

~

~h y '%

s REFERENCE AOP-4, P.1

)

'O) f%

U C

4, p

i

'4

.,-f

}

r v,

(?

s,.

is k ',' 1 r

't t

, ;N

')

?,

4

.?",

te 1

h s

0

}

,, '; i n

~l I

\\

r I

'f \\.

k

,y g

k?

L N

s v

az__eQU18111$611VE_EBQQEQUBESt_QQUQ1IIQUlt_eUQ_Lld11eI19NS PAGE 35 N...

ANSVERS -- FT. CALMOUN

-87/03/03-WHITTEMORE, J.

t.

i

((

T

\\'6 one s

.kNSWER

.\\;je.01 (3.00'][

g-Wh s. Y leis In compliance, unidentified limit is 1 GPM.

[0.5]

[0,

• ?lN/ Accept either answer below:

[0.5]

• j "d:Out of compliance,' individual S/G 1eakage exceeds 0.3 GPM, by S.O

,s In compliance, Total leakage less than 1.0 GPM, by T.S.

_y 3.

In compliance, identified leakage spec. is 10 GPM

[0.25]

,3 K

4.

In compliance, identified leakage spec. is 10 GPM

[0.25] (1.5) t Vo.

Not in compliance (0.5] as total leakage is 10.1 GPM. [0.25]

(0.75)

V D

c.

' Containment: Cont. Dewpoint inst.

p Cont. Rad m'onitor Cont. Sump Level inst.

CVCS VCT inventory inst.

[Any 3, 0.15 es.]

nW Secondary:

S/G 8/D rad $ inst Cond. Exh.y PL, analysis.

[Any 2, 0.15 ea.]

(0.75) y, monitor Periodic sample

}

t.

REFERENCE FC TS 2.1.4, Special Order # 35 a

s ANSWER 8.02 (2.50) 5.s l' Channel calibration (0.5) i 0

'3 b.

iio provide adequate power tilt detection. [0.5)

Reduce power to 70%

[0.25] to prevent undetected flux tilt from exceeding design peaking factors. [0.25] OR;' Place i channel in trip within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> [0.25] to provide m'inqumtdegree, of redundancy. [0.25]

y

-c.

All rods mpsi tbh, withdrawn to' the full out position [0.5] because the limiting incideni (Single CEA withdrawal) cannot occur. [0.53 (1.0) i REFERENCE FC TS DEF's, & TS 2.14, 2.15

.u

(

r

(

Y I

i t.-

!i a

s 4,

,.]y

~

~

~ ~~

~

v ; R;-'

3:;'

.)' '(-

.y "j

n.G #

t)

.' Ai-80%N11IB8IIVE EBQQEQWBEli_QQNQlIl0N1t aND_ Lit 3116Il0N1

.PAGE.36 4

1 c

nr ?_q ry s

.-: ANSW$RS D,, 'FT. ' CALHOUN

-87/03/03-WHITTEMORE,-J.

y

.c.

1 4

-g 1

T

.i.

if;.

b, r5 j,

V

.,i w

x

'A 4

? ANSWER-8.03.

1 (2.00)

['

f

~

, I _.

s*:

J-i G. ;,- - - - - - - - - - - - - - ld - ; 3. -

9N

~-

t

)

3 f,5 -(f p

)

/g-J-

s.,

6,~; - - - J, Az.

}'

_-------- 2.

[

+

x,,

(;

%;c. ----------------- 4.

c 1.,

e.

m..

~

j d. --- s l '. -

(0.5 ea.)

(,2 ' 0 )

~

.c.. s, v..

1 REFERENCE u

1 Licensee - Quest ion. Ba'nk, '2 3 ('4. 2 3 )

Ju 1

K e

n...

n ir(

j!.)

[

?j-p+,

i i.y 8.04

(?.,50) c

'.AN SWE R,

+

t 1a.

- STA l-( 0 ~. 5 )

'bi.

$TA.= and Or ig inator ~-

(1.0) a-

c.. ~ Reactor 3 Engineer,'PRC, SARC. (.0. 2 5 e a.

rev. an'd 0.25 for correct'

.W %

s,-

order)-

n(1.0).

y e

y jy-

g if yi

'c REFERENCE' W

r S.

O.

R-4-3, 5.0

/ 1 l-

.h 4

'(ANSWER' 8.05 (2.00).

+

i Plant Endi.nehr

.(0.5) c.

b.-

Maintenance Order.'.

i-(0.5)

c '.

Maintenance Order. and PRC approved procedure.

(1.0) l(Accept-~any valid example of PRC approved procedure, eg; Surv.

~

. test, Design Change Order, Special' Maintenance ~ Procedure, etc.)

~

-REFERENCE' T

S. : 0.

G-23,.1 3 '. 0,

i' i

3 3

4 ti.

• l&

t[rQt a r,p.

4.-

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

/

}

?At_18DMINISIB611VE_PBQQEQUBEft_QQNQ111QN1t_6NQ_ Lid 11611QNS PAGE 37 a

-ANSWERS.-

FT. CALHOUN

-87/03/03-WHITTEMORE,'J.

4 IANSWER 8.06 (3.00):

.a.-

Assistant ~ Reactor Operator (0.5) b.

Licensed Equipment Operator (0;5).

c.;

Take charge of-the_MCR operating crew.

(0.5)

.d.

' Fort Calhoun. store room.-(Warehouse)

(0.5)

I a.~

G e'it r o n i c s line 1 is used for-fire fighting communications,.J(0.25) while line 3 is used for. operations. [0.25]

Sound powered phones (0.25)..or portable radios (0.25) are'used for backup.

(1.0).

REFERENCE 4

S.'

O.

G-28,.Pp 7;8,12 ANSWER.

8.07 (2.50) a.

Not required if a pressurizer. safety /PORV' removed,.the reactor vessel head is removed, or no charging pump available to borate.

(Accept any 2 of'3, 0.5 ea.]

(1.0)

Eb..

1.

3 2.

0 3.

'2 N

4.

3 5.

1 (0.3 ea.]

(1.5) i-REFERENCE i -

TS 2.3 l

1

~

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

At__6Dd1N11IB611YE_EBQQEQMBEst_QQUQ1IIQUlt_aNQ_LidlI8IIQNS PAGE - 38

= ANSWERS'-

FT.'CALHOUN

-87/03/03-WHITTEMORE, J.

ANSWER 8.08'

( 2. 50 )'

a.

Fuel' Cladding

1. ^ Coolant activity 2.

. Core exit temperature 3.

Core subcooling-4.

. Power. OR;: Safety setpoint

['any 3, 0.25 ea;)

Reactor Coolant System

'1.

RCS l'eak rate

~

2.

RCS pressure 3.

Containment pressure AND radiation level-4.

Pressure / Temperature limits'.

[any 3, 0.25 ea.)

(1 ~ 5).

b.

-ALERT----------------- 1 BARRIER FAILED F

SITE EMERGENCY---------2 BARRIERS FAILED GENERAL EMERGENCY------3 BARRIERS-FAILED

[0.33 EA.)

(1.0)

REFERENCE

EPIP OSC-1,' Pp5,6

- ANSWER 8.09 (2.50) a.

Shift. Supervisor or Plant manager.-

(0.6) b ',

Nebraska State Patrol, Washington County sheriff, and Blair Fire Department.

(0.9) i c.

(Copy' original message)and handle as little as possible (0.5) in order to preserve evidence for investigators. [0.5)

(1.0) 4 O

REFERENCE FC QUESTION BANK & S.O.

G-16 t

1 4

i a.m,'

~

y, g np4,

.,--n.,,,-

w p.---,w

,v

--g-,,,--wn.,,.e m<-

w-r-.n--,

v-m

-w-e< w -

~

e-A

)

At__aQd1NISIBoIIVE_EBQQEQUBEft_QQU01I1991t_eU0_L10116I1983 PAGE 39 ANSWERS -- FT. CALHOUN

-87/03/03-WHITTEMORE, J.

ANSWER 8.10 (2.50) c.

100 mrem / day-300 mrem / week 1250 mrem /qtr 5000 mrem /yr (0.25 ea.]

(1.0) b.

Plant Health Physicist, Supervisor C/RP, Plant Mgr., HP Tech, ALARA Coordinator, or Group Supervisor. (Any 3, 0.25 ea.]

(0.75) c.

A device which continually indicates the dose rate.

A device which integrates the dose and alarms at the limit.

Accompanied by a qualified individual to exercise positive control of activities in the area.

(0.25 ea.]

(0.75)

REFERENCE T.S 5.11.1 & RAD. PROT. MAN., Pp VII-2-4,9 l

e. -

TEST CROSS REFERENCE PAGE 1

QUESTION VALUE REFERENCE 05.01-1.00 WJE0001085

.05.02 2.00 WJE0001086 05.03

-1.00 WJE0001087 05.04-1.00 WJE0001088 05.05 3.00

'WJE0001069 05.06

'2.50 WJE0001090 05.07 1.50 WJE0001091 05.08 3.00 WJE0001092 05.09' 1.00 WJE0001093 05.10 1.50 WJE0001094

<05.11 1.50 WJE0001095 05.12 2.00 WJE0001096 H05.13 3.00 WJE0001097

'05.14 1.00 WJE0001098 25.00 06.01 2.50 WJE0001099 06.02 3.00 WJE0001100 06.03 3.00 WJE0001101 06.04 3.00 WJE0001102 06.05 2.50 WJE0001103 06.06 2.50 WJE0001104 06.07 2.50 WJE0001105 06.08 2.50 WJE0001106 06.09 1.50 WJE0001107 06.10 2.00 WJE0001108 25.00 07.01 2.50 WJE0001119 07.02 2.50 WJE0001120 07.03 2.50 WJE0001121 07.04 2.50 WJE0001122 07.05 2.50 WJE0001123 07.06 2.00 WJE0001124

--07.07 3.00 WJE0001126 07.08 1.00 WJE0001127 07.09 2.50 WJE0001128 07.10 2.50 WJE0001129 07.11 1.50 WJE0001130 25.00 08.01 3.00 WJE0001109 08.02 2.50 WJE0001110 08.03 2.00 WJE0001111 08.04 2.50 WJE0001112 08.05 2.00 WJE0001113 08.06 3.00 WJE0001114

TEST CROSS REFERENCE PAGE 2

QUESTION VALUE REFERENCE

-03.07 2.50 WJE0001115 08.08 2.50 WJE0001116 08.09 2.50 WJE0001117 08.10 2.50 WJE0001118 25.00 100.00 l

lr-l,

_________