ML20141B342

Discusses Results of 970409 Generic Fundamentals Exam, Administered to 155 Candidates at 25 Facilities.Bwr/Pwr Facility Comments/Resolutions,Contractor Explanation of Mislabled Options & Final Answer Key Encl

ML20141B342
Person / Time
Issue date:	05/13/1997
From:	Usova G NRC (Affiliation Not Assigned)
To:	Richards S NRC (Affiliation Not Assigned)
Shared Package
ML20141B348	List: ML20141B342
References
NUDOCS 9705150246
Download: ML20141B342 (39)

Text

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NUCLEAR REGULATORY COMMISSION f WASHINGTON. D.C 20666-0001-1 \

• • • • • l May 13, 1997 MEMORANDUM TO: Stuart Richards, Chief Operator Licensing Branch Division'of Reactor Controls and Human Factors, NRR FROM: George M. Usova h I d-Operator Licensing Branch Division of Reactor Contro' f and Human Factors, NRR -

SUBJECT:

GFE RESULTS: APRIL 1997 i

The April 9,1997 Generic Fundamentals ExWnation (GFE) was administered to {

155 candidates at 25 facilities. The examination operated smoothly and

.without incident.

The summary statistical results follow:

PWR BWR No. of' examinees 84 71 Mean score 91.07% 88.73%

Median score 91% 90%

High score )

98% 98%

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Low score 82% 71%  !

Number /% of failures 0/00% 4/5.6%

Failures by facility: Hope Creek - 1 (77%) 1 Peach Bottom - 1 (71%) l Clinton - 2 (77%, 76%) l 1

The statistical results of this exam, e.g., mean scores and range are generally in line and stable with past exam performance. Overall exam difficulty level (i.e., mean score) is targeted at 87.00 and actual exam

' difficulty levels of 91 and 89 are sufficiently close to targeted goals and consistent with past GFE performance.

Some slight rise in mean scores should reasonably be expected given facility "last day" withdrawals. I note that in this examination all facilities combined withdrew,.during the final week preceding the exam, a total of 22 candidates from their initial count of intended test takers. The initial

' count of intended test takers was 177 candidates; however,155 candidates took the examination. The most common reason given for last minute withdrawals was N

that the candidates withdrawn were not prepared for the NRC/GFE given their performance on the respective facility audit examinations.

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! Withdrawing a substantially large number of potentially poor performers from the GFE exam naturally excludes those potentially weak performers from affecting the GFE mean score since their inclusion in the exam would tend to

have.a depressing effect on the mean score. Conversely, the remaining examinees, who partook in the examination, likely represent a better prepared and more knowledgeable group of examinees. Hence, this more selective group

.of examinees will have the tendency to skew the mean score upward.

Facility Comments:'

In' summary, three BWR facilities made seven comments on five questions. 1 Similarly five PWR facilities made 18 comments on nine questions of which  !

Beaver Valley, alone, accounted for 8 of the 18 comments.  !

1 Regarding the BWR facility comments, the contractor reviewed and researched

. each of the comments affecting five items. Based upon its own analysis (see

-Attachment 1 for a complete discussion of comments), the contractor  ;

4 recommended, for HOLB review and approval, that the grading of two items (60 '

i- and 86) be adjusted to accept two correct answers as follows:

Question # Change  :

^

BWR 60/88 Accept two answers, B or D '

BWR 86/14 Accept two answers, A or B  !

1 The contractor further recommended that no answer key changes be made to the l remaining three BWR items.

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Regarding PWR items, the contractor again reviewed the comments affecting those nine items. Based upon its analysis (see Attachment 1), the contractor )

recommended, for HOLB review and approval, that the grading of four items (22, i 37, 44, and 64) be adjusted to accept two correct answers and recommended that i no answer key changes be made to the remaining five PWR items.

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PWR 22/50 Accept two answers, B or D 1 PWR 37 (form B) Accept two answers, A or D (form B only)

PWR 44/72 Change answer to B (for Beaver Valley and Vogtle only) j- PWR 64/92 Accept two answers, A or C (for Ginna only)

In particular, items 44 and 64 above, were facility-specific exceptions that warranted an-additionally correct answer for Beaver Valley, Vogtle, and Ginna only. On the Sand, item 37.of Form B contained a mislabelled distractor option seqw.:c where the letter "D" was in the "A" position. This unfortunate :currence inadvertently caused only one candidate at Beaver

. Valley to t,enscribe "D" on to the answer. sheet rather than his intended "A".

In fairnest, I agreed that the "D" response should be granted credit. As I noted, h N ver, this error only affected one candidate nationwide, Beaver Valley, and had no pass / fail consequence. See Attachment 2 for the 5

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contractor's explanation of this occurrence. The contractor has committed to stronger quality control measures to ensure that errors of this type do not  :

4 occur-again.

NRC Resolution of Comments:  :

Headquarter's staff reviewed the contractor's BWR and PWR recommendations and concurred with the recommendations. The final answer keys were revised  !

accordingly to accept answer key changes as recommended. See Attachment-3.

In keeping with our recent HOLB policy to provide comment resolution feedback ,

to those facilities making comments, I directed the contractor to include, within the facility grade report, a copy of the specific NRC resolution (s) to comments made by that individual facility.

.In summary, this examination administration was a successful one.

l Attachments: 1. BWR/PWR Facility Comments / Resolution

2. . Contractor explanation of mislabeled options

3. Final Answer Key -

i DISTRIBUTION:

Central Files PUBLIC

, HOLB RF GUsova DMcCain SSpessard/BBoger  ;

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Document Name: G:\US0VA\GFEAPR97 OFC HOLB:DRCH d NAME GUsova:rc M OATE 5/6/97 0FFICIAL RECORD COPY e

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S. Richards contractor's explanation of this occurrence. The contractor has committed to stronger quality control measures to ensure that errors of this type do not occur again.

NRC Resolution of Comments:

Headquarter's staff reviewed the contractor's BWR and PWR recomendations and concurred with the recomendations. The final answer keys were revised accordingly to accept answer key changes as recomended. See Attachment 3.

In keeping with our recent HOLB policy to provide coment resolution feedback to those facilities making comments, i directed the contractor to include, within the facility grade report, a copy of the specific NRC resolution (s) to coments made by that individual facility.

In summary, this examination administration was a successful one.

Attachments: 1. C'JR/PWR Facility Comments / Resolution

2. Contractor explanation of mislabeled options

3. Final Answer Key DISTR!BUTION:

Central Files PUBLIC HOLB RF GUsova DMcCain SSpessard/BBoger Document Name: G:\US0VA\GFEAPR97 0FC F10LB:DRCH d NAME GUsova:rc M DATE 5/l$/97 0FFICIAL RECORD COPY

t .e S. Richards contractor's explanation of this occurrence. The contractor has committed to stronger quality control measures to ensure that errors of this type do not occur again.

NRC Resolution of Comments:

lihadquarter's staff reviewed the contractor's BWR and PWR recommendations and concurred with the recommendations. The final answer keys were revised accordingly to accept answer key changes as recommended. See Attachment 3.

In keeping with our recent 110LB policy to provide comment resolution feedback to those facilities making comments, I directed the contractor to include,

-within the facility grade report, a copy of the specific NRC resolution (s) to comments made by that individual facility.

In summary, this examination administration was a successful one.

' Attachments: 1. BWR/PWR Facility Commerits/ Resolution

2. Contractor explanation of mislabeled options

3. Final Answer Key

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l ATTACHMENT 1

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FACILFIY-HOPE CREEK EXA.M-HWR FORM A/B '

QUESTION: 60/88 A reactor has been operating at 100% power for eight weeks when a reactor scram occurs. The~

reactor is critical 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> later and power is increased to 100% over the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Ilow is core Xe-135 concentration behaving when power reaches 100%7 A. Xe-135 is building in toward equilibrium.

B. Xe-135 is burning out toward equilibrium.

i C. Xe-135 is building in toward a peak value.

D. Xe-135 is burning out toward a minimum value.

ANSWER: 11.

COMMENT:

i Xenon concentration twelve hours after a scram with a subsequent stanup and power escalation i beginning six hours after the scram is diflicult to predict. Xenon concentration for such a scenario I would actually be determined by computer program. Ilope Creek has limited experience with stanups occurring within twenty hours of a full power scram. The student text (page attached) supports Selection "D" as the most correct answer. It also supports Selection "A" during a stanup which quickly establishes a power level of greater than 10%. Depending on the exact circumstances, we believe "A", "B", or "D" could be considered correct. Given the stated scenario, we ,vould not expect an operator to diff:remiate between selections at the recalllevel of knowledge.

RESPONSE

Panially concur. Compared to the nonnal core xenon-135 behavior following a reactor scram from long term 100% power, this scenario would resuh in the following important deviations Xenon would not reach its nonnal peak value following a scram Xenon would peak and tum sooner following the scram llecause of the rany time associated with the power increase, the imtial rate ofxenon bumout would be less than the burnout rate if reactor power had been quickly increased to 100%. As a resuh, xenon may or may not decrease below the 100% equilibrium level before fmally stabilizing.

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', i NRC RESPONSE TO FACILITY COMMENTS FOR TflE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION e

in addition to the above, xenon would decrease for 4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after reavor power reaches 100% due to the ' differences in halflives between I-135 and Xe 135. This mean, that xenon certainly will be burning out when power reaches 100%.

He knowledgeable examinee mayjustifiably conclude that, when power reaches 100%, xenon is either burning out toward a minimum value (if cmrently less than equilibrium) or burning out toward  ;

equilibrium (if currently greater than equilibrium). In either case, however, xenon will be buming out when power reaches 100%. Herefore, options B and D are both possible. On the other handi option A is not possible because it states that xenon will be building in..

Based on the interim answer key, this question was answered correctly by'45/71 examinees and yielded

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I a moderate positive discriminatiou index of+0.23. De answer key has been changed to accept both options B and D as correct answers. - I i

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' FACILifY-HOPE CREEK

. EXAM-RWR FORM A/B QUESTION: 66/94 1

A reactor is operating at 80% power near the end of a fuel cycle. Which one of the following lists the typical method (s) used to add positive reactisity during a normal power increase to 100%7 A. Withdrawal of deep control rods and increasing recirculation flow rate D. Withdrawal of deep control rods only C. Withdrawal of shallow control rods and increasing recirculation flow rate 4

D. Withdrawal of shallow control rods only ANSWER: A. '

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COMMENT:

1 Near the end of fuel cycle, there may be no deep rods remaining in the core. His occurs just prior to coast dowr Depending on how close the plant is to the end of cycle, shallow rods and i recirculation flow rate may be used to add positive reactivity. Selection "C" could be considered j

. as an alternate correct answer. '

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RESPONSE

, 4 Do not concur. End of fuel cycle power coast down is an atypical operation that would have been identified in the premise of the question. Making the assumption that the plant was operating in coast

' ' down would be unwarranted 'and incoirect Secondly the premise of the question states a normalthat

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power increase will occur. His is contrary to the assumption that coast down is occurring.

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUN" WENTALS EXAMINATION 2

Based on the interim answer key, this question was annvered correctly by 31/71 examinees and yielded a moderate positive discrimination index of+0.21. No answer key change is required.

FACILITY-HOPE CREEK EXAM-BWR FORM A/B QUESTION: 84/12 Which one of the following pairs of fluids undergoing heat transfer in identical heat exchangers will yield the greatest heat exchanger overall heat transfer coefficient?

A. Oil to water B. Air to water l

C. Steam to water

/ D. Water to water l

, ANSWER: C.

COMM ENT: i When steam condenses to water the convection heat transfer coemcient is quite large. When no phase change occurs the convection heat transfer coeflicient is very low. The answer on the examination key is "C"- steam to water. If the steam is saturated and rejects some ofits latent l heat ofvaporization, "C"is correct. If the steam is superheated, which it must be ifit is to reject energy and remain as steam, the correct answer is "D"- water to water. Selection "D" could be considered as an alternate correct answer.

RESPONSE

Do not concur. The facility comnwnt failed to provide an example of an application of a superheated steam to-water heat exchanger. The typical BWR has several applications that use steam-to-water heat e.tchangers, e.g. feedwater heaters and the main condenser. In these applications, steam condenution causes a large overall heat exchanger heat transfer coeflicient, thereby supporting option C as thr: correct answer. It would be unwarranted and incorrect for an examinee to assume a superheited steam heat exchanger application that is foreign to a BWR plant when several relevant applier.tions exist.

Based on the interim answer key, this question was answered correctly by 29/71 examinees and yielded a nearly zero discrimination index of-0.02. No answer key change is required.

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION '

FACILITY-LIMERICK

) EXAM-BWR FORM A/B QUESTION: 84/12 Which one of the following pairs of fluids undergoing heat transfer in identical heat exchangers will yield the greatest heat exchanger overall heat transfer coeflicient?

A. Oil to water B. Air to water C. Steam to water D. Water to water ANSWER: C.

COMMENT:

The question should be deleted for three reasons:

1.

There is insuflicient infbrmation given. To make the calculation to determine respective overall ifr coeflicients (U), the candidate would need the film thickness and conductance for each of the conditions given. That data was omitted, fbrcing the candidate to make an estimate.

2.

The answer "c" is not correct for all conditions. According to Granet (1980, pp 537) the estimated U for water to water can exceed the U for steam to water in the case of a tank heater. Since the type of heat exchanger was not specWd, the candidate may make a reasonable assumption (See attached copy of the U estimation table). '

3.

He question is beyond the scope of required knowledge. Even given the tools and the data to make the determination, the K/As in Section 293007, HEAT TRANSFER and HEAT EXCIIANGES do not support the question. The most closely related item, Kl.06, requires the operator to " discuss the factors which affect heat transfer in a heat exchanger."

Given the need for supporting data, the lack of a clear tie to a K/A, and the reference information to the contrary of the answer, the question should be deleted.

Reference:

Gianet, I. (1980), Thermodynamics and heat power, Reston, VA: Prentice-Hall NRC RESPONSE TO FACILITY COMMENTS FOR TIIE APRi. 19P"' NRC GENERI FUNDAMENTALS EXAMINATION

RESPONSE

Do not concur for the following reasons:

In response to facility cornmem I, requiring examin!es to calculate exact values of heat transfer coeflicients is beyond the scope of the GFE. A general understanding of the heat transfer characteristics for the listed fluids is sufIicient to determine the correct answer.

In response to facility conunent 2, the question stated that identical heat excInngers were to be considered. 'Ihe reference provided by the facility does not show comparison values for identicai heat exchangers. Rather, the reference gives coeflicient values for different applications. Consequendy coeflicients resuhing from these applications cannot be directly compared. Instead, one can observe from the reference that steandto-water heat exchangers can resuh in maximum coeflicient values of 600 Iltuatr ft'- F,2 whereas watei-to-water heat exchangers can result in maximum coeflicient vahles only 275 BtuAu-fl - F. 'This is consistem with the text,1Icat Transfer and Fluid Flow, General Elec Febmany 1985, page 7-27, which shows a typical overall heat transfer coeflicient of 600 Btu /hr F.2,op for BWR condensers and feedwater heaters, both o.' .diich are steandto-water heat exchangers.

'Therefore, based on the relevant infonnation listed in d.. r:ference the answer key (option C) is supponed.

In response to facility comment 3, the referenced K/A does suppon the knowledge being tested question. This question is also supported by:

291006Kl.03, " Principle ofoperation ofcondensers" 291006K1.10, " Basic heat tramfer m a heat exchanger"

• l 293007K1.07, " Describe how a presence ofgases er steam can affect heat transfer andfluid flowin heat exchangers. "

1 Based on the interim answer key, this question was answered correctly by 29/71 examinees '

a nearly zero discrimination index of-0.02. No answer key change is required.

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FACILflY-LIMERICK I i

EXAM-BWR FORM A/11 QUESTION: 92/20 1

Which one of the follmving describes the fuel-to-coolant thermal conductivity for a fuel assembly '

at the beginning of core life (IlOL) as compared to the end of core life (EOL)?  ;

A. Larger at IlOL due to a higher fuel pellet density i

11.

Larger at IlOL due to lower contamination offuel rod fill gas with fission product gases C.

Smaller at 110L due to a larger gap between the fuel pellets and clad D.

Smaller at IlOL due to a smaller corrosion fdm on the surface ofthe fuel rods ANSWER: C.

l COMMENT: ,

' The correct answer is "II." not "C." This question is essentially identical to an existing INPO exam bank question (Thennal Limits #238) which was also incorrect until December,1996; the date of the issuance of the GF Ilank update (See attached copy).

The question was corrected in response to feedback (See attached copy of the feedback fonn dated 3/31/95). The technical justifica: ion for modifying the answer for both the exam bank question and the GFE question is identical; the fonner having already been corrected.

RESPONSE

Do not concur. The facility-provided reference material does not directly address how IlOL fuel-to-coolant themial conductivity compares to EOL thermal conductivity. 'i mentions factors that affect the MAPLilGR limit curve. '1he referen . pMut out that thennal cars .ictivity is a factor that influences the shape of the MAPLIIGR limit curve, I hwever, the curve .N Maienced by many factors, not ju the change in thermal conductivity. hefore, it cannot be pn..nmed that a change in thermal conductivity alone is responsible for the observed change in the MAPLHGR limit etuve. Likewise. it cannot be concluded that a decrease in the heat transfer coefDeient cf the fuel pin gases causes a reduction in the overall flicl-to-coolant themial conductivity. Hence, this argument does not justi accepting a second correct answer.

The text, Ileat Transfer and Fluid Flow, General Electric, February 1985, page 9-45 states "A short time after reactor startup the [new] fuel cracks taduy and redistributes out to the cladding." When the fuel pellets come into contact with the cladding, thermal conductivity increases dramatically.

The above text, page 9-?2, also states "The overall heat transfer coefIicientfor regions the pellet is in contact with the clad is given by its conductance through the gases." This excerpt comes from a NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION -

i section that is discussing the change in the MAI'LIIGR limit curve from MOL to EOL This quote l indicates that the fuel pellet maintains contact with the clad over core life. Because there is essentially l no direct contact between fhel pellets and clad at the BOL, the direct contact that exists at the EOL '

causes overall fuel-to-coolant thermal conducthity to be higher.

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Based on the interim answer key, this question was answered correctly by.46/71 examinees and yielded I a small positive discrimination index of+0.08. No answer key channe is required.-

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1 NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC i FUNDAMENTALS EXAMINATION  !

FACILITY-QUAD CITIES EXAM-BWR FORM A/B QUESTION: 86/14  !

l During full power operation, critical heat fh.x is most likely to occur in t o  !

A. centes fuel bundle wut flow restrictions.

B. center fuel bundle without flow restrictions.

1 C. outer fuel bundle with flow restrictic.

D. outer fuel bundle without flow restrictions. I

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ANSWER: A.

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COMMENT:

The provided answer ke" shows "A" as the correct answer. The facility requests that "B"also be  !

considered as a correct answer for the following reasons. The question references the likelihood  ;

of critical heat flux occurring. The term " flow restrictions"is confusing and can be interpreted two ways, first as core orificing and second as some other form of flow blockage, such as debris.

If " flow restriction" refers to additional flow blockage to an already orificed core then "A" is l

correct. If" flow restriction" refers to core orificing, then "B"is correct, i 1

Reference (s): General Physics Corporation '

BWR Generic Fundamentals l Hermodynamics  ;

Chapter 8, Hennal liydraulics, Rev.1 Page 22

RESPONSE

Concur. He text, lleat Transfer and Fluid How, General Electric, February 1985, page 8-40 refers to )

core odfices as " Local restrictions such as the inlet odfice.. ". Herefore, it is conceivable that an examinee might interpret the term " flow restrictions" as core orifices. If so, option B would be the correct answer.

r Based on the interim answer key, this question was answered correctly by 47/71 examinees and yielded a small positive discrimination index of +0.17. All of the remaining 24 examinees selected option B.

De answer key has been channed to accept both options A and B as correct answers.

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NRC RESPONSE TO FACILITY COMMENTS FOR TIIE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FACILITY-QUAD CITIES ,

EXAM-BWR FORM A/B QUESTION: 92/20 Which one of the following describes the fuel-to-coolant thermal conductivity for a fuel assembly at the beginning of core life (BOL) as compared to the end of core life (EOL)?

A. Larger at BOL due to a higher fuel pellet density

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l B. La*;ger at BOL due to i>wer contamination offuel rod fill gas with fission product gases l

C. Smaller at BOL due to a larger gap between the fuel pellets and clad l 1

D. Smaller at BOL due to a smaller corrosion film on the surface of the fuel rods ANSWER: C.

COMMENT:

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He provided answer key shows "C" as the correct answer. The facility requests that "B" also be I considered as a correct answer fbr the following reasons. The question references fuel to coolant thennal conductivity at BOL compared to EOL. The question refers to core life and not bundle life. Since a fuel bundle can remain in the core for several operating cycles, this can be confusing, causing answer to vary. If the time span referred to is bundle life, then "B"is correct since fission product gas contamination affects the fuel themiallimits at the end of bundle life. For a new fuel bundle "C"is correct. As the fuel ratchets and expands from neutron exposure it may come into contact with the clad. His would cause thermal conductivity to increase, but this occurs only early in bundle life.

Reference (s): General Physics Corporation BWR Generic Fundamentals Thermodynamics Chapter 9, Core Thennal Limits, Rev.1 Page 41

RESPONSE

Do not concur. Regardless ofwhich phrase is used (core life or bundle life), the intent of the question premise remains the same; to compare the fuel-to-coolant thermal conductivity of a fuel bundle before and aller long tenn power production. Secondly, the comparison is made between only two specific ieference points, BOL and EOL ne parameters that affect thermal conductivity during core burnup (e.g., contamination of fuel rod gases) are not relevant unless they support the fmal conclusion, that thermal conductivity is smaller at BOL Option B states that fuel thermal conductivity is larger at BOL and , therefore,is incorrect.

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NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION The references provided by the facility point out that thermal conductivity is a factor that influences the shape of the MAPLIIGR limit curve. However, the curve is influenced by many factors, not just the change in thermal conductivity. Therefore, it cannot be presumed that a change in thermal conducthity alone is responsible for the observed change in the MAPLHGR limit curve. Likewise, it cannot be concluded that a decrease in the heat transfer coeflicient of the fuel pin gases causes a reduction in the overall fuel-to-coolant thermal conductivity. Hence, this argument does not justify accepting a second conect answer.

The text, Heat Transfer and Fluid Flow, General Electric, Febmary 1985, page 9-45 states "A short time after reactor startup the [new] fuel cracks radially and redistributes out to the cladding." When the fuel pellets come into contact with the cladding, thermal conducthity increases dramatically.

The above text, page 9-22, also states "Ilie overall heat transfer coeflicientfor regions the pellet is not in contact with the cladis given by its conductance through the gases." This excerpt comes from a section that is discussing the change in the MAPLHGR limit curve from MOL to EOL. This quote indicates that the fuel pellet manitains contact with the clad over core life. Because there is essentially no direct contact between linel pellets and clad at the 11GL, the direct contact that exists at the EOL causes overall fuel-to-coolant themial conductivity to be hiples.

Ilased on the interim answer key, this question was answered correctly by 46/71 examinees and yielded a small positive discrimination index~of +ROS. No answer key _ change is required.

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NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 N'RC GENERIC FUNDAMENTALS EXAMINATION FACILffY-BEAVER VALLEY EXAM-PWR FORM A/B QUESTION: 14/42 Which one o"the following will cause an upscale failure of a boron-trifluoride (BF3) failed fuel detector operating in the proportional region?

A. The detector electrode high voltage power supply output ha,s decreased 5% due to setpoint drift.

B. The detector chamber has become flooded with water due to leakage around the electrodes.

C. A power supply fuse in the amplifier circuit for the neutron monitoring instmment has opened.

D.

A temperature rise has caused the gas pressure inside the detector to increase to within 5 psi  ;

of design pressure.

ANSWER: 11.

COMMENT:

Most candidates had to answer this question by eliminating the distracters. No reference could be located that supports the key choice ofB..

Most of the staff concur that a 11F3 proportional counter will fail if flooded with water but most would expect the instrument power fuses to open, causing a downscale meter response.

Recommendation: Delete from this exam. Modify for future use. Change choice B to read "The l

discriminator voltage level has decreased to 25% of its required level." OR "the detector develops a short between the electrodes. l Reference (s): Westinghouse, Radiation, Chemistry and Corrosion Considerations for Nuclear l Power Plant Application, Page 5-22.

i RESPONSE- l Do not concur. The proposed replacement options would be potentially correct answers for this I question. Ilowever, each replacement option also has the potential of opening the instrument power fuses.

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Second, this question should not be assumed to be fauhed simply because the process of elim' nation was used by a number of examinees to identify the correct answer. Eliminating obviously wong options can be an effective method for problem sohing, especially for this question if the exannnee  !

c

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION lacked the knowledge that borated water has higher conductivity than the inert gas used in the BF3 detector.

Based on the interim ans.ver key, this question wn answered correctly by 63/84 examinees ar.33 ielded a moderate positive discrimination index of+0.25. No answer key change is required.

FACILITY-BEAVER VALLEY EXAM-PWR FORM A/B  !

, QUESTION: 22/50 l

A centrifugal pump is operating at rated conuitions in an open system with all valves fully open. l If the pump suction valve is throttled to 50% closed, pump discharge pressure will and pump differential pressure will

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A. remain the same; remain the same

B. decrease; remain the same C. remain the same, decrease D. decrease; decrease 4

ANSWER: D.

COMMENT:

Too many assumptions are required to determine the correct answer. Specifying an open system without any system description presents choice "B" as possibly correct.

By assuming that the system determines pump discharge pressure, as it would discharging into the

, bottom of a large tank choice B would be correct. To determine that the pump differential decreases the candidate must assume suction pressure drops below the minimum NPSH.

Recommendation: Accept B and D for this exant Avoid asking candidate to consider actions which are counter to " good operating practices" or include in the question" ... erroneously throttled...". Accompany the question wid: a diagram that better defines the pump / system arrangement. Possibly add initial suction pressures and temperatures that show a reasonable close proximity to saturation.

Reference (s): BV, Nuclear Operator Training, Thermodynamics, Chapter 4, System, INmps and Valves.

t nrm RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC ,

FUNDAMENTALS EXAMINATION RESPONGE:

Concur. Depending on the assumption by an examinee both B and D could be correct. If the exam'mee assumed that partially closing the suction valve maintained pump net positive suction head (NPSH), then B is correct. If the examinee assumet that partially closing the suction valve reduced pump NPSH below requiied NPSH, then cavitation would occur and the answer would be D.

Therefore, both B and D could be considered as correct answers.

Facility recommendations for question improvement are all valid.

Based on the interim answer key, this question was answered correctly by 29/84 examinees and yielded a near zero discrimination index of +0.01. The answer key has been changed to accept both B and D i as correct answers.  !

1 FACILITY-BEAVER VALLEY EXAM-PWR FORM B ONLY QUESTION: 37 A centrifugal pump is taking suction from the bottom of a vented cylindrical storage tank that contains 100,000 gallons ofwater at 60 F. A pressure gauge at the inlet to the pump indicates 40 psig. Over tl e next several days storage tank temperature increases to 90 F with no change in tank water level and no change in head loss in the pump suction line.

Which one of the following is the current approximate pressure at the inlet to the pump?

D. 39.8 psig l

B. 37.4 psig C. 34.6 psig D. 31.2 psig ANSWER: A.

COMMENT:

Obvious Typographical Error Choices lettered: D D

C D

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC -

FUNDAMENTALS EXAMINATION Recommendation: Change first choice to "A"  !

RESI ONSE:

Concur. This typographical error occurred only on the form B exam. It had the potential ofcausing an examinee to mark option D on the answer sheet when option A was being chosen. Therefore, because +

s the answer is option A, both options A and D should be considered as correct answers.

Based on the intenm answer key, this question was answered correctly by 74/84 examinees and yielded a high positive discrimination irvi-x of +0.39.

• Six of the ten examinees who answered incorrectly selected option D on form B. The answer key for fom; D has been chariged to accept both options A '

and D as correct armvers.

FACILIIWHEAVER VALLEY EXAM-PWR FORM A/B QUESTION: 44/72 While remotely investigating the condition of a typical normally-open motor control center  ;

(MCC) feeder breaker, an operator observes the following indications:

l Green breaker position indicating light is lit.

Red breaker position indicating light is out.

MCC voltmeter indicates zero volts.

MCC ammeter indicates zero amperes.

llased on these indications, the operator can accurately report that the breaker is open and racked to position.

A. the OUT 1 J

l B. the IN C. the TEST D. an unknown ANSWER: D.

COMMENT:

1 Two problems:

Defining "Remately" '

Defining fypical" 34.-

ew r "

i NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION If" Remotely" means control room, the MCC voltmeter and ammeter readings are not available.

j For some Breakers the test position will remove all remote (control room) light indication and l

others the light indication is maintained. Depending on assumptions concerning " Typical" and j "Rernote" choice "C" or "D" could be correct.

Recommendation: Drop from this exam. Recommend this question be removed from the examination bank, due to difliculty in derming " Typical" Various breaker types and control power schemes in use throughout the industry provide operators a widely varied view of what l

" Typical" means.

I i

RESPONSE

Panially concur. De term " remotely"is commonly used in the nuclear industry when referring to a location some distance away from the equipment being operated (such as the control room or another remote location). Regardless ofwhere the remote station (s)ivare, the question provided the necessary information and should not have caused any confusion as to the source of the information.

I i

Also, it is not clear how option C might be considered a correct answer. De comment states that, for i some breakers, remote position indication will not be available if the breaker is racked to the TEST position. His would seem to indicate that the breaker must be racked IN. His would cause option B ,

to be correct if the typical MCC feeder breaker at Beaver Valley lases remote breaker position indication when in the TEST position.

Upon request, the facility submitted electrical drawings for a sample MCC feeden breaker. R ese drawings show that remote breaker position indication is removed when the breaker is in the TEST position. Upon follow-up, the facility representative stated that the sample breaker position indication circuit is representative of most MCC feeder breakers at Beaver Valley. In addition, the facility provided follow-up documentation to clarify their position on this question. He follow-up documentation states "For Beaver Valley choice 'B' appears to be the correct answer."

Based on comments from this and another facility, it appears that this question is facility-specific.

Herefore, because Beaver Valley has shown that their MCC feeder breakers lose remote position indication when in the TEST position, the answer to this question for Beas er Valley should be B.

Finally, the term " typical"is used to direct the examinee away from rare or uncommon examples ofthis breaker application. Without its occasional use, there could be cases where an exception would prevent testing some important knowledges.

Based on the interim answer key, this question was anmered correctly by 36/84 exammees and yielded a very high positive discrun' m ation index of+0.43. He answer key will be channed to accept option B as the correct answer for Beaver Valley.

. 5.

NRC RESPONSE TC "ACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FACILITY-BEAVER VALLEY I EXAM-PWR FORM A/B '

QUESTION: 60/88 l

A reactor had been operating at 50% power for two weeks when power was increased to 100%

over a 3-hour period. In arder to maintain reactor power stable during the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, which one of the following incremental control rod manipulations will be required?

i A. Withdraw rods slowly during the entire period +

B. Withdraw rods slowly at first, then

... a rods slowly C. - Insert rods slowly during the entire period

D. Insert rods slowly at first, then withdraw rods slowly ANSWER

D. l COMMENT:

Although the candidates answers this question by considering the required control rod ,

manipulations, most later stated that no rod motion was required to maintain rezetor oower  ;

stable.

Above the POAll steam demand controls reactor power and rods or boron control reactor coolant temperature.

Recommend 1 tion: Modify question for future use. Modify question to read, "In order to i maintain Reactor Coolant temperature stable.

l l

Reference (s): Westinghouse, Reactor Core for Large 15ssurized Water Reactors, Pages 9-21 through 9-23, '

RFSPONSE:

Concur. The facility recommendation for question revision is valid.

i Based in the interim answer key, this question was answered correctly by 82/84 examinees and yielded a small positive discrimination index of+0.12. No answer key change is reauired.

l i

i i

i t , .

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FACILITY-BEAVER VALLEY l EXAM-PWR FORM A/B QUESTION: 64/92 Why are bumable poisons installed in a reactor core?

A. To shield reactor fuel from thermal neutron flux until later in core life IL To compensate for control rod depletion that occurs over core life  !

C. To flatten the axial thermal neutron flux distribution early in core life D.

To ensure that the control rods will be above the rod insertion limit when the reactor is  !

critical ANSWER: A.

COMMENT:

Choice D appears to be an equally correct answer.

Purpose of Bumable poisons, includes ofTset of 16, when considered with soluble boron (limited by positive MTC) and control rods (limited by SDM and hot channel factor limits).

Without poisons the reactor could be made critical with rods below insertion lh Reconunendation: Accept 'A' or 'D' on this examination and replace distracter D for future l examinations.

Reference (s): Westinghouse, Reactor Core Control for Large Pressurized Water Reactors, Pages 8-9 through 8-13.

RESPONSE

Do not concur. It is tme that bumable poisons pemiit a lower RCS boron concentration and consequently a more negative MTC. 110 wever, placing bumable poisons in the core does not ensure control rods will be above the rod insertion limit (RIL) when the reactor is critical Rod position at criticality is a function of many core parameters, e.g., RCS boron concentration, core fission product poisoning, core bumup, residual bumable poisons, and others. He critical position of the control rods is estimated prior to a startup. If criticality is predicted with control rods below the RIL, then the operator will adjust RCS boron concentration to attain a higher estimated critical rod position.

Herefore, RCS boron concentration is the only parameter that can be controlled to ensure that control rods are above the RIL when the reactor is critical Based on the interim answer key, this question was answered correctly by 20/84 examinees and yielde a moderate positive discrimination index of 40.24. No answer key change is reouired.

-17

i NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC

~

FUNDAMENTALS EXAMINATION

• FACILITY-BEAVER VALLEY EXAM-PWR FOP.M A/B QUESTION: 74/2 Consider a pressurizer containing a saturated water / steam mixture at 636 F with a quality of I 50*/o.. If an outsurge removes 10% of the liquid volume from the pressurizer, the temperature of the mixture will' (

and the quality of the mixture will .

(Assume the mixture remains saturated.) '

A. decrease; decrease i

11. decrease; increase C. remain the same; decrease D. remain the same; increase o

ANSWER: 13.

COMMENT:

Not clear to all candidates that the pressurizer was half full ofwater and half full of steam, both at saturation. Some attempted to apply the stated 50% Quality to the steam space alone.

l Afler assuming the steam had significant moisture content, they logically concluded that pressure  !

would not drop until all moisture was gone (100% stm quality). Others that focused on just the i quality of the steam space overlooked the obvious, "Outsurge removes liquid from the  ;

pressurizer." '

l l

Recommendation: Modify question for fliture use. See attached rewrite of question that avoids using the tenn Quality which is most commonly used to describe the amount of moisture (or lack of)in steam.

Proposed Replacement

^

Consider a pressurizer containing a steam bubble and a volume of water. Both the water and steam are uniformly at saturated conditions for 636*F. The volume of steam and water in the pressurizer are equal. If an outsurge removes 10% of the liquid volume from the pressurizer, the temperature of the pressurizer will , and the volume of the steam will be than the volume of the water.

A. decrease,less l

.B. decrease, greater

-18

l l

l NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC 1 FUNDAMENTALS EXAMINATION 1

C. remain the same, less D. remain the same, greater

RESPONSE

i Concur. He proposed replacement contains some enhancements that will improve the clarity of the 1

. question.-

Based on the interim answer key, iis question was answered correctly by 52/84 examinees and yielded a small positive discrimination index of+0.1 Ic No answer key change is required.

I FACILITY-BEAVER VALLEY '

EXAM-PWR FORM A/B QUFSTION: 84/12 Which one of the following pairs of fluids undergoing heat transfer through a heat exchanger will yield the greatest heat exchanger overall heat transfer coeflicient? i l

.A. Oil to water B. Air to water i

C. Steam to water D. Water to water  !

ANSWER: C.

COMMENT: '

\

This question asks about "Overall Heat Transfer Coeflicient" but the " Preliminary Key" answer, "C" seems to indicate the question is asking about the heat transfer rate.

Overall IIeat Transfer Coeflicient for a Tube:

\

v. 1 s l

u. .5.

.l& #

. J.

h,

\

. i 2 i 2; ,

Note that all components of this formula for heat transfer coefficient describes the physical

{

arrangement and the film coefficients of the two fluids involved. This fonnula does not consider  !

I

_ _a

I NRC RESPONSE To FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION '

whether the latent heat of evaporation is involved. Air-to-water could be the correct answer considering the very small film layer on the air side of the heat transfer boundary. j i

Recommendation: Modify question for future u3e. Modify the question lead to read. " yield the greatest heat exchanger heat transfer rate?" I l

Reference (s): BV, Nuclear Operator Training, Thermodynamics, Chap. 3 - Connection and Fluid Flow, Page 3. l

RESPONSE

Partially concur. While it would be correct to use the wording suggested in the comment, it is not necessary to change the current wording. The comment uses the equation for the overall heat transfer I coeflicient for a heat exchanger tube asjustification. However, the question refers to the entire heat exchanger. 'Re thermal conductivity of the heat exchanger tubes is not the only factor to be considered when evaluating the overall heat transfer coeflicient of a heat exchanger in which fluids are flowing. More importantly, the convective heat transfer coeflicients for the flowing fluids must be considered.

Ilased on the interim answer key, this question was answered correctly by 30/84 examinees and yie '

a small positive discrimination index of +0.12. No answer key change is required.

1 i

l l

ll i

l I

l l

l l

l

-2 0-i

3,

• NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 N FUNDAMENTALS EXAMINATION FACILITY-COMANCIIE PEAK EXAM-PWR FORM B ONLY QUESTION: 37 A centrifugal pump is taking suction from the bottom of a vented cylindrical storage tank th contains 100,000 gallons ofwater at 60 F. A presure gauge at the inlet to the pump indicates 40 psig. Over the next several days storage tank temperature increases to 90 F with n i tank water level and no change in head loss in the pump suction hne.

Which one of the following is the current approximate pressure at the inlet to the pump?

D. 39.8 psig B, 37.4 psig_

C. 34.6 psig D. 31.2 psig ANSWER: A.

COMMENT:

Question number 37 of the Form B examination has a typographical error.

Three of the four candidates using Fonn B correctly answered the question, circling the l by the correct response. When they copied their answer to the answer sheet, they fille See the attached pages from the student's examination ;.

I Recommend that you accept both A and D for the Form B of this examination.

RESPONSE:  !

Concur. This typographical error occurred only on the form B exam. It had the po  !

examinee to mark option D on the answer sheet when option A was being chosen. He '

the answer is option A, both options A and D should be considered as correct answers.

1 i

Based on the interim answer key, this question was answered correctly by 74/84 a high positive discrimination index of +0.39.

selected option D on fonn B. He answer key for form B has been c and D as conect answers.

l 1

-21

NRC RESPONSE TO FACILITY COMMENTS FOR THE APluL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FA CILITY-COMANCIIE PEAK EXAM-PWR FORM A/B QUESTION: 84/12 Which one of ti.e following pairs of fluids undergoing heat transfer through a heat exchanger will yield the greatest heat exchanger overall heat transfer coeflicient?

A. Oil to water B. Air to water C. Steam to water D. Water to water ANSWER: C.

COMMENT:

This question asks for the combination of heat exchanger lluids yielding the greatest overall heat transfer coefIicient. The key answer is steam to water.

The thermal conductivity of water is superior to steam, oil, and air (see attached Figure E.1). If the fluid on one side of the heat exchanger is steam, the overall thennal conductivity will have decreased from the same heat exchanger with water for the fluid. The drastic reduction of the heat transfer coeflicient when Departure from Nucleate Boiling occurs is based on this phenomenon.

If the water on one side of the heat exchanger is undergoing a phase change (nucleate boiling, for example), then the overall heat transfer coeflicient improves. He question does not give enough infonnation for a candidate to consistently make that interpretation for steam to water.

Recommend that you change the answer key to accept either "C. Steam to water" or "D.

Water to water."

RESPONSE

Do not concur. He thennal conductivity values for steam and water are not the principle factors to be considered when evaluating the overall heat transfer coeflicient of a heat exchanger in which fluids are flowing. More importantly, the thennal conductivity of the heat exchanger tubes and the convective heat transfer coeflicients for steam and water should be considered.

It has been documented (Granet, I., Thermodynamics and heat power,1980, page 537) that stearrw to-water heat exchangers can result in maximum coefficient vahies exceeding 600 Btu /hr-ft'- F, whereas water-to-water heat exchangers can resuh in maximum coeflicient values of only about 275

- l e ,

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION  !

)

Bru/br-A' *F.

i This is consistent with the text, IIeat Transfer and Fluid Flow, General Electric, l February 1985, page 7-27, widch shows a typical overall heat transfer coefficient of 600 Bru/hr-ft ,op 2

for a BWR condenser and feedwater heaters, both of which are steam-to-water heat exchangers.

'Iherefore, based on the maximum coefficients listed for various steam-to-water and water-to-water applications, option C is the only correct answer.

A phase change is an atypical occmrence for most water-to-water heat exchangers (an exception be a PWR steam generator). Such an occurrence would then give the heat exchanger the characteristics of a water to-steam heat exchanger and would not result in a valid conparison.

]

flased on the interim answer key, this question was answered correctly by 30/84 examinees and yielded a small positive discrimination index of+0.12. No answer key channe is requird.

I l

1

. l i

d

. l 1

4

,Y 1

l

NRC RESPONSE TC FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FACILITY-GINNA EXAM-PWR FORM A/B QUESTION: 49/77 After the first fuel cycle, suberitical multiphcation can produce a visfoie neutron level indication on the source range -'aar instrumentation following a reactor shutdown without installed neutron sources. This is because a sufficient source ofneutrons is being produced by:

A. spontaneous neutron emission from control rods.

B. photo-neutron reactions in the moderawi.

C. Iow level thermal fission in the fuel.

D. alpha-neutron reactions in the fuel.

ANSWER: II.

COMMENT:

Reconunend deleting question. Dealt with source neutrons without installed sources. Installed sources were removed at Ginna for our Cycle 20 refueling (currently on Cycle 25) and the Westinghouse WCAP 14290 for fuel analysis states that the primary source of neutrons is spontaneous fission of CM-242 and CM-244. Therefore, no correct answer is given.

RESPONSE

i Do not Conc *it.

He question does not ask for thepnmary source of source neutrons ne knowledgeable examinee should know that option 11 (photo-neutron reactions) is the only significant source ofneutrons. listed in the answer options.

Based on the interim answer key, this question was answered correctly by 51/84 examinees and yielded a small positive discrimination index of+0.12. No answer key change is required.

i e

6

-24 ,

NRC RESPONSE TO FACILTfY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FACILITY-GINNA EXAM-PWR FORM A/B

' QUESTION: 64/92 Why are burnable poisons installed in a reactor core?

A.

To shield reactor fuel from thermal neutron flux until later in core life B. To compensate for control rod depletion that occurs over core life C. To flatten the axial thermal neutron flux distribution early in core life D.

To ensure that the control rods will be above the rod insertion limit when the reactor is critical ANSWER: A.

COMMENT:

Recommend answer C as a correct choice. Dealt with reasons for installing bumable poisons.

Choice A, given as correct answer, describes what they do, not why they are installed. Ginna's Cycle 25 WCAP 14290 for fuel reload states that one of the reasons is to reduce axial power peaking, which is answer C. We use integral fuel burnable assemblies and only the center section of the fuel rod is coated. The primary reason is to reduce the amount of soluble boron needed to account for K excess at BOL, which was not one of the choices.

RESPONSE

Concur. The facility-specific reference provided by the facility supports the facility recommendation to accept option C as a correct answer. Because Girma was the only facility to provide justification for accepting option C, it appears that this function of bumable poisons is unique to Ginna.

Based on the interim answer key, this question was answered correctly by 20/84 examinees and ielded 3

a small positive discrimination index of +0.24. The answer key has been channed to accept both options A and C as correct answers for Girma.

-25

i NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC FUNDAMENTALS EXAMINATION FACILITY-SALEM I EXAM-PWR FORM B ONLY QUESTION: 37 i

A centrifugal pump is tak.ing suction from the bottom of a vented cylindrical storage tank that  !

contains 100,000 gallons ofwater at 60 F. A pressure gauge at the inlet to the pump indicates 40 psig. Over the next several days storage tank temperature increases to 90 F with no chan tank water level and no change in head loss in the pump suction line. j i

I Which one of the following is the current approximate pressure at the inlet to the pump?

D. 39.8 psig i

D. 37.4 psig C. 34.6 psig I

\

D. 31.2 psig  !

4 ANSWER: A. I COMMENT: i "A"is not shown as a possible selection. There are two "D's". Our proctors told candidates (who pointed out this) that the first "D" should be marked as "A" on the SCAN'lRON Form.

The proctors did not make a general announcement so that we would not indicate someone els had "A. " arrived at the answer. Candidates should have determined or asked if

RESPONSE

Concur. His typoyaphical error occurred only on the fonn B exam. It had the potential o examinee to mark option D on the answer sheet when option A was being chosen. Herefo the answer is option A, both options A and D should be considered as correct answers.

Based on the interim answer key, this question was answered correctly by 74/84 exa a high positive discrimination index of +0.39.

Six of the ten exanunees who answered incorrectly selected option D on fonn B. De answer key for form B has been changed to accept both opti and D as correct answers.

• t' NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC- '

FUNDAMENTALS EXAMINATION FACILITY-SALEM EXAM-PWR FORM A/B QUESTION: 64/92 Why are burnable poisons installed in a reactor core?

A. To shield reactor fuel from thermal neutron flux ua:il later in core life

. B. To compensate for control rod depletion that occurs over core life C. To flatten the axial thermal neutron flux distribution early in core life D. '

To ensure that the control rods will be above the rod insertion limit when the reactor is critical ANSWER: A.

., COMMENT:  !

The answer on the examinatio'n key is "A" and it is correct. Selection "D"is a correct statement when one considers the long-life cores and the technical specification for moderator temperature coeflicient (MTC). Burnable poison compensates for Kexcess at BOL, allowing a soluble boron concentration that supports MTC within technical specification limits. Without burnable poison, the control rod height at criticality would have to be very low to allow a boron concentration that supports MTC within technical specil; cations. We suggest that this question be revised prior to next use.

RESPONSE

Do not concur. Placing bumable poisons in the core does not ensure control rods will be above the rod insertion limit (RIL) when the reactor is critical Rod position at criticality is a function of many core parameters, e.g., RCS boron concentration, core fission product poisoning, core bumup, residual 1

burnable poisons, and others. The critical position ofthe control rods is estunated prior to a startup.

criticality is predicted with control rods below the RIL, then the operator will adjust RCS boron concentration to attain a higher estimated critical rod position. Therefore, RCS boron concentration is the only parameter that can be controlled to ensure that control rods are above the RIL when the reactoris critical Based on the interim answer key, this question was answered correctly by 20/84 exambiees and yi a moderate positive discrimination index of+0.24. No answer key change is reauired.

27-

s '

NRC RESPONSE ~ . FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC  :

FUNDAMENTALS EXAMINATION FAClLITY--SALEM EXAM-PWR FORM A/B -

QUESTION: 84/12 Which one of the following pairs of fluids undergoing heat transfer through a heat exchanger will yield the greatest heat e ' anger overall heat transfer coeflicient?

A. Oil to water B. Air to water C. Steam to water D. Water to water ANSWER: C. '

COMMENT

When steam condenses to water the convection heat transfer coeflicient is quite large. When no phase change occurs the convection heat transfer coeflicient is very low. The answer on the examir,ation key is "C"- steam to water. If the steam is saturated and rejects some ofits latent

heat ofvaporization, "C"is correct. If the steam is superheated, which it must be ifit is to reject energy and remain as steam, the correct answer is "D"- water to water. Selection "D" could be considered as an altemate correct answei. I I

RESPONSE

Do not concur. He facility comment failed to provide an example of an application of a superheated steam-to-water heat exchanger. He typical PWR has several applications that use steam-to-water heat exchangers, e.g. feedwater heaters and the main condenser, in these applications, steam condensation causes a large overall heat transfer coeflicient, thereby supporting option C as the correct answer. It would be unwarranted and incorrect for an examinee to assume a heat exchanger application that is I foreign to a PWR plant when several relevant applications exist.

Based on the interim answer key, this question was answered correctly by 30/84 exanunees and yielded a small positive discrimination index of+0.12. No answer key change is reauired.

I NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENER FUNDAMENTALS EXAMINATION FACILITY-VOGTLE EXAM-PWR FORM A/B QUESTION: 22/50 A centrifugal pump is operating at rated conditions in an open sys em with all valves fully open.

If the pump suction valve is throttled to 50% closed, pump discharge pressure will and pump differential pressure will A. remain the same; remain the same fl. decrease; remain the same C. remain the same; decrease D. decrease; decrease ANSWER: D. l l

COMMENT: l We duplicated these conditions on our loop flow training device, which consists of an open tank as the suction source and the discharge collection point. All valves in the discharge and suction flowpaths of the centrifugal pump were fully opened after the pump was started. Discharge pressure and pump D/P were measured under full flow conditions. These parameters were measured again as a suction valve was partially throttled shut. The results were that pump D/P remained constant and pump discharge pressure decreased.

11ased on this data, we believe that the correct choice on this question to be "B."

RESPONSE

Partially concur. Depending on the assumption by an examinee both options B and D could be correct.

If the examinee assumed that partially closing the suction valve maintained pump net positive suctio head (NPSII), then B is correct. If the examinee assumed that partially closing the suction valve reduced pump NPSH below required NPSI1, then cavitation would occur and the answer would be D.

Herefore, both B and D could be considcred to be correct answers.

Based on the interim answer key, this question was answered correctly by 29/84 examinees 3 and ielded a near zero discrimination index of +0.01. He answer key has been changed to accept both B and D as correct answers.

• * ., a,. .

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENER ITJNDAMENTALS EXAMINATION FACILITY-VOGTLE EXAM-PWR FORM B ONLY QUESTION: 37

' A centrifugal pump is taking suction from the bottom of a vented cylindrical storage tank that contains ,

100,000 gallons ofwater at 60 F. A pressure gauge at the inlet to the pump indicates 40 psig. Over the next several days storage tank temperature increases to 90 *F with no change in tank water level and no change in head loss in the pump suction line.

Which one of the following is the runent approximate pressure at tile inlet to the pump?

D. 39.8 psig T

B. 37.4 psig C. 34,6 psig D. 31.2 psig ANSWER: A.

COMM ENT: ,

The correct answer was the first choice (39.8 psig) vhich apparently was mislabeled as choice D.

RESPONSE

Concur. His typographical error occurred only on the form B exant It had the potential ofcausing examinee to mark option D on the answer sheet when option A was being chosen. Therefore, because the answer is option A, both options A and D should be considered as correct answers.

Based on the interim answer key, this question was answered correctly by 74/84 exammees an a high positive discrimination index of +0.39.

Six of the ten examinees who answered incorrectly selected option D on form B. De answer key for fann B has been changed to accept both option and D as correct answers.

h o

-30

NRC RESPONSE TO FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION FACILITY-VOGTLE EXA.M-PWR FORM A/B QUESTION: 44/72

-While remotely investigating the condition of a ' typical normally-open motor control center (MCC) feeder breaker, an operator observes the following indications:

Green breaker position indicating light is lit.

Red breaker position indicating light is out.

MCC voltmeter indicatec =ro s olts.

MCC ammeter indicates zero amperes.

Based on these indications, the operator can accurately report that the breaker is open and racked to position.

A. the OUT B. the IN C. the TEST

, D. an unknown i ANSWER:. D.

1 COMMENT:

' Question states that an operator is " remotely" investigating the condition of a typically normally-open MCC feeder breaker. The operator observes the green light is lit, red light is out, and MCC voltmeter and ammeter both read zero.

This question has two correct answers depending on how you interpret what the tenn " remotel means.

If remotely means at the breakers' remotely controlled station (i.e., control roora) then that fact that you have a green indicating light shows that the breaker has control power to the remote control circuit. This indicates that the I reaker must be racked to the "in" position, which would make choice D correct. l Ifyou interpret " remotely"to be at the breaker in question then choice D would be correct since indicating lights on the breaker cubical function in both the " test" and "in" positions.

l Accept choices B and D as correct on this examination.  !

Other factors that provide additional confusion are:

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NRC RESPONSE m FACILITY COMMENTS FOR THE APRIL 1997 NRC GENERIC FUNDAMENTALS EXAMINATION MCC voltage and ammeter indications-where are you reading these from, locally at the MCC or remotely somewhere else?

MCC feeder breaker-implies the supply breaker to that MCC which is typically a 180 Vac switchgear breaker, which may be in a different room compared to the MCC.

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RESPONSE

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Partially concur. He term " remotely"is commonly used in the nuclear industry when referring to a location some distance away from the equipt. . being operated (such as the control room). The tenn l

" local" would have been used if the indications were located at the breaker cabinet. 1 i

The comment states that the lit green indicating light indicates that the breaker must be racked to the l

IN position. Ilowever, no documentation was provided to support the comment. Upon request, the facility submitted electrical drawings for a sample 480 Vac MCC feeder breaker. Rese drawings show l that remote breaker pouion is removed when the breaker is in the TEST position. Upon follow-up, )

the facility representative stated that the sample beaker position indication circuit is representative of l

most MCC feeder breaken at plant Vogtle i

Based on comments from this and another facility, it appears that this question is facility-specific. l

'lherefore, because Vogtle has shown that their typical MCC feeder breakers lose remote position indication when in the TEST position, the answer to this question for Vogtle should be B.

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Regarding the other factors which caused confission, it is not certain that the examinee would benefit l from additional wording to address these issues. On the contrary, additional wording actually may reduce the quality of this question by introducing information that some examinees would find distracting.

Based on the interim annver key, this question was answered correctly by 36/84 examinees and yielded a very high positive discrimination index of +0.43. He answer key has been changed to accept option ILas the correct answer fbr Vogtle.

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