ML20134G590

Forwards Results of Gfes of Written Operator Licensing Exam, Administered on 961009,to Nominated Employees of Facility

ML20134G590
Person / Time
Site:	Wolf Creek
Issue date:	11/07/1996
From:	Brockman K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Carns N WOLF CREEK NUCLEAR OPERATING CORP.
References
NUDOCS 9611130286
Download: ML20134G590 (7)

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/h .1 NUCLEAR REGULATORY COMMISSION REGION IV 611 RY AN PLAZA DRIVE. SUliE 400 AR LINGToN, T E X AS 760118064 1

9....+J NOV T 1996 Neil S. Carns, President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, Kansas 66839

SUBJECT:

GENERIC FUNDAMENTALS EXAMINATION RESULTS This letter forwards the results of the Generic Fundamentals Examination Section (CFES) of the written operator licensing examination that was administered on October 9,1996, to nominated employees of your facility. We are forwarding the following items:

o the examinations, including answer keys; o the results for your nominated employees; and o copies of the individual answer sheets completed by your nominated employees.

We request that your training department forward the individual answer sheets and results to.the appropriate individuals. It should be noted that the examination was administered in two forms, which were identical except for the sequence of questions.

In accordance with the Commission's regulations, 10 CFR 2.790, a copy of this letter and the examination and answer key will be placed in the NRC's Public j Document Room (PDR). The individual results and answer sheets are exempt from  !

public disclosure and, therefore, will not be placed in the PDR. j Questions concerning this examination should be directed to Dr. George Usova at l (301) 415-1064.

Sincerely, va enneth ro an, Acting Director Division of Reactor Safety Docket: 50-482 Licenses NPF-42 Enclosures As stated cci (see next page) l l

9611130286 961107 PDR ADOCK 05000482 y

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Wolf Creek Nuclear Operatin8 Corporation cc:

Gary Boyer, Manager Nuclear Training Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 l

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DRP Branch Chief Resident Inspector  !

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11/04/96 (f 11/(/96 11/ f/96 / /96 / /96 OFFICIAL RECORD COPY

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, FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE FACILITY -- WOLF CREEK

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4 EXAM -- PWR FORM A/B QUESTION: 60/88 A reactor is initially operating at 50% power with equilibrium core xenon-135. Power is l increased to 100% over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period and average reactor coolant temperature is adjusted to 588 F using manual rod control. Rod control is left in Manual and no subsequent operator j actions are taken. 1 I

, Considering only the reactivity effects of core xenon-135 changes, which one of the following describes the average reactor coolant temperature 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the power change?

j A. Greater than 588 F and decreasing slowly  ;

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B. Greater than 588*F and increasing slowly i C. Less than 588 F and decreasing slowly l D. Less than 588 F and increasing slowly ANSWER

A. I COMMENT:

This question deals with understanding when xenon will dip on a power increase from 50% to 100%. Answer "A" is stated as the correct answer; answer "B" should also be accepted based on the following:

There are two thumbrules used to estimate the time of the xenon dip. Either:

.8 x / Power Change from the INPO GFES bank (attached), or

/ Power Change from Wolf Creek's training material (attached) ..

Using the first thumbrule has xenon dipping at 5.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, and using the second thumbrule has xenon dipping at 7.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. it is not reasonable to have examinees determine the time of the xenon dip at the accuracy mquired to answer this question using a thumbrule. Estimating the effects of xenon either 4 or 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the power change would be more reasonable.

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< l FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE i i

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RESPONSE

Do not concur. Following a power increase minimum core xenon will occur after 4 io 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

For a 50% increase the required time is approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> based on graphs provided in Westinghouse (Reactor Core Control for Large PWRs,1983, p. 4-26) and General Electric (BWR  !

Academic Series, Reactor Theory,1984, p. 6-10a) Therefore, for the power change listed in the question, only option A is correct.

The facility comment stated that 7.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> should be accepted as the time to minimum core xenon based on a thumbrule in the facility lesson plan. However, the thumbrules in the facility training material refer to the number of hours to reach peak xenon following a power decrease.

In fact, these thumbrules are most accurate when estimating the time to peak xenon following  !

a reactor shutdown (or trip). They are not accurate when used to estimate the time to minimum xenon following a power increase.

Based on the interim answer key, this question was answered correctly by 61/124 examinees and yielded a moderate positive discrimination index of +0.21. No answer key chance is reauired.

l FACILITY -- WOLF CREEK EXAM -- PWR FORM A/B QUESTION: 88/16 Which one of the following must be present to prevent departure from nucleate boiling from occurring in a reactor core following a pressurizer vapor space instrument line rupture if the leak rate is less than normal makeup capability?

A. Reactor coolant pump flow capability B. Pressurizer level in the indicating range C. Emergency core cooling injection capability D. Steam generator steaming capability ANSWER: D.

COMMENT:

This question deals with determining what is necessary to prevent departure from nucleate boiling following a Pressurizer vapor space leak within the normal makeup capacity. Distractor "A" should also be considered correct because having RCP flow will increase the margin to DNB.

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FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE The attached training material states how flow oscillations, due to no forced flow, can lower the Critical Heat Flux required for DNB by as much as 40%. Also, Technical Specifications tout flow as being important when considering DNB. -

RESPONSE

Concur. An examinee knowledgeable in heat transfer and thennal hydraulics should be able to readily eliminate options B and C. Option B can be eliminated because it does not directly affect heat transfer conditions in the core. Option C can be eliminated because the leak rate is less than the normal makeup capability. Option D is the correct answer. Option A is the only remaining option that directly affects heat transfer in the core. Therefore, options A and D will be accepted.

Based on the interim answer key, this question was answered correctly by 48/124 examinees and yielded a small positive discrimination index of +0.09. The answer key has been chanced to accept either A or D for full credit.

FACILITY -- WOLF CREEK EXAM -- PWR FORM A/B QUESTION: 96/24 Which one of the - '; wing will prevent brittle fracture failure of a reactor vessel?

A. Manufacturing the reactor vessel from low carbon steel B. Maintaining reactor vessel heatup/cooldown rates within limits C. Maintaining the number of reactor vessel heatup/cooldown cycles within limits D. Operating above the reference temperature for nil-ductility transition (RTm)

ANSWER: D.

COMMENT:

This question deals with the prevention of brittle fracture. Answer "D" is state.d as the correct answer; answer "B" should also be accepted as correct because the training material implies brittle fracture is prevented by maintaining pressure and temperature to the right of curves based on heatup and cooldown rates (attached).

1 FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE

RESPONSE

Do not concur. Option B does not mention maintaining pressure and temperature in decordance with the pressure-temperature curves for various heatup and cooldown rates. It refers to only heatup and cooldown rates (*F/hr). Simply maintaining heatup and cooldown rates within limits will not prevent brittle fracture. That is why pressure-temperature curves and overpressure protection systems were developed.

Based on the interim answer key, this question was answered correctly by 92/124 examinees and yielded a small positive discrimination index of +0.13. No answer key change is reauired.

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