Text

Summarizes 961009 Generic Fundamentals Exam Administered to 179 Candidates at 24 Facilities
	ML20134N421
Person / Time
Issue date:	11/21/1996
From:	Usova G; NRC (Affiliation Not Assigned)
To:	Richards S; NRC (Affiliation Not Assigned)
References
	NUDOCS 9611260334
	Download: ML20134N421 (55)
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November 21, 1996 4

MEMORANDUM TO: Stuart A. Richards, Chief Operator Licensing Branch Division of Reactor Controls and Human Factors, NRR N

FROM: George M. Usova (3' l Q5 1p-

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Operator Licensing Branch Division of Reactor Controls -

and Human Factors,

SUBJECT:

GFE RESULTS: OCTOBER 1996 The October 9,1996 Generic Fundamentals Examination (GFE) was administered to 179 candidates at 24 facilities. The examination operated smoothly and without incident. Of particular note, however, was a larger than normal number of candidates withdrawn from this test administration during the last several days leading up to the exam. Although facilities had originally registered 205 candidates to take the exam, there were 26 late withdrawals reducing the final group to 179. By contrast, there are usually no more than five to seven candidate withdrawals on any given examination. I intend to look into the matter since such large candidate withdrawals may distort the population for which the test intends to measure.

The October examination summary statistical results are as follows:

RWE B0/.8 Number of examineek 55 124 Number of failures 2 11 Mean score 90.9 87.3 High score 100 98 ,

Low score 74.7 72 Number of facilities 8 16 ,- (r-,).

Q) l Number of comments 2 (2 questions) 30 (10 questions) 96-nY 260059 Wt_o/

NRC RLE CBITER COPY ,

9611260334 961121 PDR ORO NRRA v, ,

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Facility Failures 79.00 Fitzpatrick 79.00 Catawba 74.00 Pilgrim 78.00 77.00 McGuire 78.00 Sequoyah j 74.00 1 79.00 St. Lucie 1 72.00 :

78.00 Surry 79.00 South Texas 79.00 Wolf Creek 79.00 Zion 1 Overall, the statistical results of this exam, e.g., mean scores, reliability, range scores, etc. are in line with recent past exam performance and are stable. Overall exam difficulty level (e.g., mean score) is targeted at 87.00 and actual exam mean difficulty levels of 87-91 are near targeted goals. -

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Of particular note: the PWR failure rate for this examination was nearly twice as high as J that of past average PWR failure rates. The number of failures per facility were fairly evenly distributed among those facilities having failures, indicating no single facility havirig any disproportionate effect on the overali failure rate. Nonetheless, I have asked the contractor to conduct a separate analysis as to why the PWR failure rate was as high {

l as it was. 1

' Facility Comments:

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'Two BWR facilities (Millstone 1 and Fitzpatrick) made comments on 54 and 82, ;

l respectively, contending that each item had two correct answers. Regarding the BWR facility comments, the contractor reviewed and researched each of the two comments.

Based upon its own analysis (See Attachment 1), the contractor recommended, for HOLB review and approval, that no changes be made to the answer key. Headquarter's staff reviewed and approved the contractor's BWR recommendation.

However, after thflinal grade reports were issued to the regions and facilities, the - #

l contractor contacted me to inform me that one facility had a continuing issue regarding -

l BWR item #82, arguing that answe ing the question required a high level of operator plant l knowledge, Upon analysis, the cor; tractor reviewed the item, contacted me, and we concurred that there was a higher taan normal amount of plant knowledge required and. :

expected of a GFE candidate to an. wor the question. This fact made the item invalid for GFE use and consequently, item 82 will be eliminated from the October 9th examination and from the GFE examination bank. The item deletion did not affect the pass / fail outcomes of any candidates. I subsequently contacted the facility representative l, responsible for the comment, thanked him for his contribution, and explained our i resolution to him.

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l The PWR examination received 30 facility on 10 items. Again, the contractor reviewed and researched each of the 10 comments. Based upon its analysis (See Attachment 2),

the contractor recommended, for HOLB review and approval, that the grading of items 42 and 55 be adjusted to accept two correct answers. The contractor, furthermore, recommended that no answer key changes be made to the remaining 8 items.

Headquarter's staff reviewed and approved the contractor's PWR recommendations.

The final answer keys reflect changes made for both examinations (See Attachment 3).

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l In keeping with our policy to provide comment resolution feedback to those facilities

! making comments, the contractor included, within the facility grade reports, 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 Facility Comments / Resolution

2. PWR Facility Comments / Resolution

3. Final Answer Keys 1

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o November 21, 1996 The PWR examination received 30 facility on 10 items. Again, the contractor reviewed and researched each of the 10 comments. Based upon its analysis f.See Attachment 2),

the contractor recommended, for HOLB review and approval, that the grading of items 42 :

and 55 be adjusted to accept two correct answers. The contractor, furthermore, I recommended that no answer key changes be made to the remaining 8 items.

l Headquarter's staff reviewed and approved the contractor's PWR recommendations.

The final answer keys reflect changes made for both examinations (See Attachment 3).

In keeping with our policy to provide comment resolution feedback to those facilities :

making comments, the contractor included, within the facility grade reports, 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:

l 1. BWR Facility Comments / Resolution +

2. PWR Facility Comments / Resolution

3. Final Answer Keys DISTRIBUTION:

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FACILITY - FITZPATRICK EXAM - BWR FORM A/B QUESTION: 82/10 i

The power range nuclear instruments have been adjusted to 100% based on a calculated heat balance. Which one of the following will result in indicated reactor power being lower than actual reactor power?

l l A. The feedwater temperature used in the heat balance calculation was lower than actual feedwater temperature.

B. The reactor recirculation pump heat input term was omitted from the heat balance calculation.

C. The feedwater flow rate used in the heat balance calculation was lower than actual j feedwater flow rate.

l l D. The steam pressure used in the heat balance calculation was lower than actual steam pressure.

i ANSWER: C.

COMMENT:

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Two correct answers.

Justification: Response "C" is correct if you consider that m. r is a factor in computing m..

Our training program (a portion of the lesson plan is attached) does not currently include this fact as a fundamentals concept. This training occurs later during the plant specific portion of l the training program. Knowledge is demonstrated by completion of several qualification card (attached) items.

l l Response "D" is also correct if, for example, the values of steam pressure under consideration l are 1000 psia and 100 psia. These pressures would compare enthalpy values of 1192.9 btu /lbm and 1187.2 btu /lbm respectively. ,

Candidate Responses: Based upon the above, it would be expected that the candidates would 1 select the "D" response. This is true for 5 of the 7 JAF candidates. The 2 that selected the "C" l response are well respected knowledgeable engineers with several years on-shift experience and may have possessed knowledge beyond the fundamentals training program.

Amplifying Information: An apparent predecessor (attached) to this question was deleted from the INPO GFE Test Item Catalog by Revision 3 to ACAD 93-002. !

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RESPONSE

Do not concur. This question does not require the examinee to know that mS is used to compute m.. Option C is still correct if the examinee considered only the effect of an erroneously low feedwater flow rate entry in the heat balance calculation.

Option D would be correct under the extreme conditions specified in the facility comment.

However, the examinee would have to assume an unrealistically low steam pressure value was used that was well out of the normal operating range of steam pressure--100 psia versus 1000 psia. There is no basis for making such an extreme assumption given the conditions postulated in the question. Therefore, option D is incorrect.

The INPO question attached to the facility comment has two correct answers (b and d) and was

probably deleted for that reason. The concept of the INPO question is still valid as evidenced i

by NRC generic thermodynamics K/A 293007K113 which states " Calculate core thermal power using a simplified heat balance".

I Based on the interim answer key, this question was answered correctly by 29/55 examinees and yielded a relatively high positive discrimination index of +0.31. No answer key chance is recuired.

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l' FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE l FACILITY - MILLSTONE 1 EXAM - BWR FORM A/B i

QUESTION: 54/82 i

Neutron flux shaping in a reactor core reduces radial power pcaking:

A. in the center of the core caused by a high number density of fuel assemblies.

B. at the periphery of the core caused by moderator reflection of thermal leakage neutrons.

C. throughout the core caused by uneven burnout of control rod poison material.

D. throughout the core caused by loading fuel assemblies of various fuel enrichments.

1 ANSWER: A. l l

COMMENT:

1 Northeast Utilities believes that there are two (2) correct answers for question number 54 for l USNRC Generic Fundamentals Examination October 1996 BWR - Form A:

The interim answer key has given answer A as the correct answer and we find that this can be a correct answer. We also find answer D to be correct as supported by the following:

A. General Electric NEDO-108% May 1973 Reactor Fundamentals Training manual Volume 5 page 15-4 which states in part "There are basically three methods of accomplishing flux flattening by localized control of K .

1 3 selective poison loading;

2. nonuniform fuel loading; and 3.' programmed or group rod control.

In the first two methods, either extra poison (absorber of thermal neutrons) may be placed in the center of the core, or fuel loading may be reduced there, to lower the l thermal utilization factor (f) and thus K, there."

l l B. NEDE - 24810 - Volume 1 Chapter 20 Core Design page 20-5 section 20.3.3.6 Thermal Limits states in part:

"The radial power distribution is flattened by increasing the K, towards the periphery of the core relative to the center. ...By placing a ring of high-reactivity fueljust l

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1 FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE 1

I inside the periphery, the radial power distribution is flattened. Loading fuel in this manner is sometimes referred to as the ring-of-fire method of loading a core.

RESPONSE: I l

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Do not concur. Option D is incorrect because of the phrase " caused by loading fuel assemblies of various fuel enrichments". As stated in the references provided with the facility comment, these fuel assemblies are loaded in a manner that flattens the flux and reduces radial power peaking. They do not cause radial power peaking.

Based on the interim answer key, this question was answered correctly by 29/55 examinees and yielded a relatively high positive discrimination index of +0.27. No answer key change is reauired.

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FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE FACILITY - MAINE YANKEE EXAM - PWR FORM A/B QUESTION: 55/83 Which one of the following describes the absolute value of integral control rod worth (negative reactivity) during the complete withdrawal of a fully-inserted control rod?

A. Increases, then decreases B. Decreases, then increases C. Increases continuously D. Decreases continuously ANSWER: D.

COMMENT:

Question number 83 (Form B), question number 55 (Form A) asked what the absolute value of integral control rod worth (negative reactivity) does during a complete rod withdrawal of an inserted control rod. You provided the correct answer as being D. " Decreases continuously".

Maine Yankee Technical Data Book Figure 2.6.1, Integral Rod Worth curve, attached, shows that integral worth increases as rods are withdrawn from the bottom of the core. This curve shows that positive reactivity is added to the core as the rods are pulled out. Therefore the rods have more negative reactivity to add to the core if inserted by a plant trip or other means. At Maine Yankee the absolute value of integral control rod worth (negative reactivity) increases as control rods are withdrawn from the fully inserted position. i l

1, I approached our Reactor Engineering department for their interpretation of this question.. As I written, they also' teel that the absolute value of integral control rod worth (negative reactivity) l increases continuously in this situation. I The correct answer for Maine Yankee is C. " Increases Continuousiv"

RESPONSE

Concur. There are several different formats used to represent integral rod worth in the nuclear l industry--some use negative reactivity units and some use positive reactivity units. Inversion of the normal means of interpreting integral rod worth would cause significant confusion at some ,

! facilities. Therefore, either C or D could be considered correct depending on the facility. I 1

FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE l Rawd on the interim answer key, this question was answered correctly by 41/124 examinees and yielded a moderate positive discrimination index of +0.19. The answer key has been changed to accent either C or D for full credit.

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FACILITY - MCGUIRE and CATAWBA

EXAM - PWR FORM A/B l

QUESTION: 14/42 Refer to the drawing of a gas-filled detector characteristic curve (see figure below). !

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l What is the effect of operating a proportional neutron detector at a voltage near the high end of the proportional region?

l A. Gamma pulses will increase in size while neutron pulses remain essentially the same,

, causing some gamma pulses to be counted as neutron pulses and yielding a less accurate i l neutron count rate.

B. A high gamma radiation field will result in multiple small gamma pulses that combine to form larger pulses, which will be counted as neutron pulses, yielding a less accurate neutron count rate.

C. Neutron pulses will become so large that gamma pulse discrimination is no longer needed, yielding a more accurate neutron count rate.

D. The positive space charge effect will increase and prevent collection of both gamma and

neutron pulses, causing a less accurate neutron count rate.

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ANSWER: B.

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FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE COMMENT:

The question asks for a comparison between operating at the high end and the low end of the proportional region, but the question does not state this explicitly. Also, a competent operator should be able to distinguish that options "C & D" are incorrect, but may not be able to differentiate between "A & B". Therefore, this question appears to be beyond the scope of detail that an operator must commit to memory. The K/A (015.000 K5.03) is listed as calibration adjustments with an importance factor of 2.3, which is below the threshold for GFES exams. Though other K/As may apply, this appeared to be the most accurate one.

Either accept "A" or "B" as a correct response, -OR- delete question and re-evaluate the tie to the K/A for operators.

l RESPONSE: l l

Do not concur. This question is supported by NRC generic component K/A 191002Kil8,

" Theory and operation ofion chambers, ..." According to Westinghouse (Radiation, Chemistry, and Corrosion Considerations for Nuclear Power Plant Application,1983, p. 5-24), "...as the voltage increases, the height of all pulses increase [s]." Therefore, option A is clearly incorrect because of the phrase "while neutron pulses remain essentially the same" Option C also is clearly incorrect because gamma pulses will increase in magnitude along with the neutron pulses, causing additional gamma pulses to avoid discrimination and be counted along with the neutron pulses.

Westinghouse (p. 5-24) also states " At low operating voltages ... large numbers of positive ions remain in the vicinity of the [ positive] anode" This is called the positive space charge effect.

Westinghouse (p. 5-27) also states "As the voltage is increased, a point is reached where the positive ions no longer build up around the anode." Therefore, the positive space charge effect diminishes as operating voltage increases. This makes option D, which begins "The positive space charge effect will increase" clearly incorrect.

Option B is the only correct option and is supported by Westinghouse (p. 5-29) which states "In a high gamma field with high operating voltage, gamma pulse pile-up results in instrument output indicating a neutron flux much higher than actually exists."

Based on the interim answer key, this question was ariswered correctly by 23/124 examinees and yielded a very small positive discrimination index of +0.02. No answer key change is required.

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I FACILITY - MCGUIRE and CATAWBA i EXAM - PWR FORM A/B QUESTION: 55/83 l Which one of the following describes the absolute value of integral control rod worth (negative ;

i reactivity) during the complete withdrawal of a fully-inserted control rod? I l A. Increases, then decreases l

l B. Decreases, then increases )

l C. Increases continuously l D. Decreases continuously ANSWER: D.

i COMMENT: )

i In the stem of the question the term " absolute value" is used. Interpreted mathematically, this implies a positive value would be derived. This would cause answer "C" to be correct.

Accept "C" or "D" as correct response.

RESPONSE

Concur. Question phrasing may have caused confusion. Additionally, several different formats are used to represent integral rod worth in the nuclear industry--some use negative reactivity units and some use positive reactivity units. Inversion of the normal means of interpreting integral rod worth would cause significant co.ifusion at some facilities. Therefore, both C and D could be considered correct depending on.the facility.

Based on the interim answer key, this question was answered correctly by 41/124 examinees and yielded a moderate positive discrimination index of +0.19. The answer key has been changed to accept either C or D for full credit.

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FACILITY - MCGUIRE and CATAWBA l EXAM - PWR FORM A/B 4

l QUESTION: 56/84 l Neutron flux shaping in a reactor core reduces radial power peaking:

l l A. in the center of the core caused by the high number density of fuel assemblies.

B. at the periphery of the core caused by moderator reflection of thermal leakage neutrons.

j C. throughout the core caused by uneven burnout of control rod poison material.

D. throughout the core caused by uneven burnout of fuel assemblies.

ANSWER: A.

i COMMENT:

Option "D" is correct also. We utilize a checkerboard fuel loading scheme that places burned l fuel assemblies next to fresh fuel assemblies in order to minimize the burnable poison requirements in the fresh fuel. The logic being that a neutron lost in a BP rod is not available for fission, therefore it is more efficient from a reactivity stand point to minimize the number of BP rods in the core. Because the fresh fuel is more reactive (even with BP rods inserted) than the burned fuel, the power sharing will be uneven between the fresh and burned fuel, thus causing uneven burnup between assemblies. Therefore, answer "D" is the best option provided.

Accept "D" or "A" as correct response.

RESPONSE: 1 Do not concur. Option D states that radial power peaking is caused by uneven burnout of fuel assemblies. Westinghouse (Reactor Core Control for Large PWRs,1983, p. 8-13) states "It should be noted that one of the major effects of burnup is to establish a flatter radial profile."

This indicates that fuel burnup does not cause increased radial peaking. In fact, burnup actually flattens the radial neutron flux distribution and reduces radial power peaking. Therefore, option j D cannot be correct.

I Based on the interim answer key, this question was answered correctly by 67/124 examinees and l yielded a moderate positive discrimination index of +0.20. No answer key chance is required.

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FACILITY - MCGUIRE and CATAWBA l

EXAM - PWR FORM A/B QUESTION: 60/88 A reactor is initially operating at 50% power with equilibrium core xenon-135. Power is increased to 100% over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 actions are taken. 1 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the power change?

1 A. Greater than 588 F and decreasing slowly l

l B. Greater than 588 F and increasing slowly l

C. Less than 588'F and decreasing slowly j D. Less than 588 F and increasing slowly ANSWER: A.

I COMMENT:

The question asks for students to determine if Xe is increasing or decreasing after six hours. I Based on our curves, Xe would still be decreasing six hours after the initial power change. '

Therefore, option "B" would be correct depending on which exact time was used.

According to the INPO Bank, the actual time would be approximately 5.65 hours7.523148e-4 days 0.0181 hours 1.074735e-4 weeks 2.47325e-5 months . With the question asking for six hours of time, there is no; enough delineation of evaluation.

Accept "A" or "B" as correct response.

Also, it would be better in the future to use 3 or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the event to clearly differentiate the answer.

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RESPONSE

Do not concur. The question states the reactor is at equilibrium when a power increase occurs. l Neither of the cases cited by the facility meets the conditions in the question. One case involves l

transient xenon; the other case involves a power decrease. However, on one of the figures ;

provided by the facility (OP-MC-RT-RP-4), a similar transient is shown. At time 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months , I power is increased from 50% (near equilibrium xenon) to 75 % over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period. According to the figure, core xenon reaches minimum and begins to increase approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the power change is initiated, or 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after the power change is completed. This provides evidence that xenon will be increasing under the conditions in the question, making A the correct answer. The higher power le" ' (100%) stated in the question will result in the lowest point of the xenon dip being reached e1 mner than shown on the figure. Therefore,6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the I l power change, core xenon clearly will be increasing. !

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 ke_v change is reauired. l l

FACILITY -- MCGUIRE and CATAWBA EXAM -- PWR FORM A/B QUESTION: 78/6 l A plant is operating at 90% of rated power. Main condenser pressure is 1.7 psia and hotwell condensate temperature is 120 F.

Which one of the following describes the effect of a 5% decrease in cooling water flow rate through the main condenser?

A. Overall steam cycle efficiency will increase because the turbine will be operating more efficiently.

B, Overall steam cvgle efficiency will increase because condensate depression will decrease.

C. Overall steam cycle efficiency will decrease because the turbine will be operating less efficiently.

D. Overall steam cycle efficiency will decrease because condensate depression will increase.

ANSWER: C.

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FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE COMMENT:

Question has NO CORRECT ANSWER. Overall steam cycle efficiency will decrease, but the l turbine will be operating MORE efficiently. As turbine delta-h decreases, it approaches an ideal turbine. Therefore, turbine efficiency increases.

l Delete question.

RESPONSE

Do not concur. As condenser pressure increases, the enthalpy change of the steam as it passes j through the turbine will decrease, causing turbine output to decrease with an overall decrease in steam plant efficiency. A decrease in turbine output does not necessarily cause turbine efficiency to increase. Without additional information (e.g., before and after values of enthalpy and moisture content of inlet steam and turbine exhaust), it is impossible to determine the actual j direction of change in turbine efficiency (although steam cycle efficiency has certainly l decreased). Therefore, option C cannot be eliminated as the correct answer. Options A and B, .

l on the other hand, can be eliminated because they state that steam cycle efficiency will increase. I Option D, also can be eliminated because a 5% decrease in cooling water flow rate will n21 l

increase condensate depression. Therefore, the knowledgeable examinee will be able to eliminate three of the four options and select C as the correct answer.

1 Based on the interim answer key, this question was answered correctly by 91/124 examinees and ;

yielded a relatively high positive discrimination index of +0.35. No answer key chance is required.

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. 1 FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE FACILITY - MCGUIRE and CATAWBA EXAM - PWR FORM A/B QUESTION: 94/22 ,

A reactor is operating at 80% of rated thermal power with power distribution peaked both radially and axially in the center of the core. Reactor power is then increased to 100% over the next two hours using only reactor coolant boron adjustments for reactivity control.

Neglecting any effect from reactor poisons, when power is stabilized at 100%, the radial peaking factor will be _ and the axial peaking factor will be .

A. higher; lower B. higher; higher C. the same; lower D. the same; higher ANSWER: D.

COMMENT:

As read, question intended to ask for the value of axial peaking factor. However, it was read by some examinees as asking for " higher or lower" relative to core location. Therefore, "C" would be correct.

Accept "C" or "D" as correct responses.

RESPONSE: , ,j Do not concur. The question referred to the qualitative change in the axial and radial peaking factors, not the change in core location of the axial and radial power peaks. Even if the question was misread, the knowledgeable examinee would dismiss the assumption that the terms " higher" and " lower" referred to core location because the location of the radial power peak cannot be l described in this manner.

l Based on the interim answer key, this question was answered correctly by 47/124 examinees and yielded a moderate positive discrimination index of +0.19. No answer key change is required.

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! FACILITY - ST. LUCIE 1

l EXAM - PWR FORM A/B QUESTION: 14/42 Refer to the drawing of a gas-filled detector characteristic curve (see figure below).

What is the effect of operating a proportional neutron detector at a voltage near the high end of the proportional region?

A. Gamma pulses will increase in size while neutron pulses remain essentially the same, causing some gamma pulses to be counted as neutron pulses and yielding a less accurate neutron count rate.

B. A high gamma radiation field will result in multiple small gamma pulses that combine to form larger pulses, which will be counted as neutron pulses, yielding a less accurate neutron count rate.

C. Neutron pulses will become so large that gamma pulse discrimination is no longer needed, yielding a more accurate neutron count rate.

D. The positive space charge effect will increase and prevent collection of both gamma and neutron pulses, causing a less accurate neutron count rate.

ANSWER: B.

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d 4 FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE COMMENT:

Figure for alpha and beta radiation given with question, question pertains to neutron and gamma radiation.

The response of radiation detectors within the proportional region are consistent, thus making that region acceptable for commercial power plant use.

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As the Limited Proportional Region is entered, pulse height discrimination and radiation measurement becomes ineffective due to pulse sizes no longer being proportional, thus making that region unacceptable for commercial power plant use.

The concept that gamma pulses combine to form larger pulses in a voltage range where l commercial power plant proportional detectors do not operate, is not an applicable GFES question.

Recommendation: . Deletion of question from exam.

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References:

Applied Engineering Principles (USN),2-51 through 2-53 l General Physics Training Manual, Gas-Filled Detectors An 1 Advanced Course,2-15 through 2-19 Eberline Neutron rem Detector Technical Manual Nuclear Reactor Engineering, Glasstone and Sesonske, 5.237-5.267 i

RESPONSE

Do not concur. The figure is a representation of the response of a gas-filled radiation detector-in a radiation field to a enange in detector operating voltage. The knowledgeable examinee should know that gamma and neutron radiation will cause the same detector response as alpha and beta.

This question is supported by NRC generic component K/A 191002K118, " Theory and operation ofion chambers, ..." This knowledge also is supported by Westinghouse (Radiation, Chemistry, )

! and Corrosion Considerations for Nuclear Power Plant Application,1983, p. 5-29) which states "In a high gamma field with high operating voltage, gamma pulse pile-up results in instrument ;

output indicating a neutron flux much higher than actually exists."

Based on the interim answer key, this question was answered correctly by 23/124 examinees and yielded a very small positive discrimination index ol +0.02. No answer key change is required.

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FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE FACILITY - ST. LUCIE EXAM - PWR FORM A/B QUESTION: 55/83 Which one of the following describes the absolute value of integral control rod worth (negative reactivity) during the complete withdrawal of a fully-inserted control rod?

A. Increases, then decreases B. Decreases, then increases C. Increases continuously D. Decreases continuously ANSWER: D.

COMMENT:

The change in the Absolute Magnitude of the integral (negative reactivity) rod worth as rods are withdrawn is a matter of perspective. The concept of integral vs. differential is important and should be tested.

If looking for the total worth change from the CEAs with the known reactivity configuration being shutdown with all CEAs inserted, then the Absolute Magnitude would increase as CEAs are withdrawn. This would make the response of "C" (answer key).

Most plant physics curve books represent CEA worth increasing as CEA height increases.

This would make the response of "C" (answer key).

If looking for the integral negative reactivity worth change from the CEAs, as the CEAs are withdrawn the following conclusions can be made:

With a CEA fully inserted it is adding the maximum amount of negative reactivity it can add.

As the CEA is withdrawn the amount mf negative reactivig it is adding decreases up to and including the full out position where it is adding zero negative reactivity.

This would make the response "D".

I

1 FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOdER 1996 GFE j Evaluating the CEA worth from an operating standpoint, particularly from a LDM standpoint, as the CEAs inscrt on a trip they add negative reactivity that offsets the positive reactivity of Power Defect.

This would make the response of "D".

Recommendation: Accept both "C" and "D".

Technical

Reference:

General Physics Reactor Theory Course l PSL Plant Physics Curves ,

L Nuclear Reactor Engineering, Glasstone and Sesonske, l

5.224-5.229-Nuclear Engineering, Lamarsh, pg 302 through 304

RESPONSE

Concur. Question phrasing may have caused ce Jon. Additionally, several different formats are used to represent integral rod worth in the. " mr industry--some use negative reactivity units and some use positive reactivity units. iv.esion of the normal means of interpreting integral rod worth would caus signiticant confusion at some facilities. Therefore, both C and D could be considered correct depending on the facility.

Based on the interim answer key, this question was answered correctly by 41/124 examinees and yielded a moderate positive discrimination index of +0.19. The answer key has been changed to accent either C or D for full credit. w-FACILn Y - ST. LUCIE

EXAM - PWR FORM A/B l

QUESTION: 88/16 Which onc;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 it tiection capability i

D. Steam generator steaming capability ANSWER: D.

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. FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE COMMENT: l This question is beyond the scope of GFES fundamentals. It requires knowledge ofintegrated plant operations along with knowledge of plant and operator response to accidents. Students at the end of the fundamentals stage have not been trained in Core Damage prevention or mitigation techniques.

In order for a student to correctly answer this question, he or she must be able to differentiate between a small break Loss of Cooling Accident and understand the mitigation methods and core heat removal methods available and used for each. For instance:

Small break LOCAs have insufficient break flow from the Emergency Core Cooling l

'~

System to remove all core decay heat; therefore, secondary heat removal capability (steam generators) is required.

Large break LOCAs have sufficiently large ECCS flow to remove all core decay heat and the steam generators are essentially uncoupled from the RCS and core and have no effect on decay heat removal.

These concepts are taught in the Integrated Plant Operations phase of training and during introduction to analyzed accidents and accident mitigation.

Recommendation: Delete question from exam.

Technical

References:

NUREG 1021, ES-205

.NUREG 1021, ES-201 NUREG 1122 K & As

RESPONSE

Partially concur. Although the conditions in the quesdon provide an accident situation, an examinee knowledgeable in heat transfer and thermal 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 danged to j accept either A or D for full credit.

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FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE FACILITY - ST. LUCIE EXAM - PWR FORM A/B i QUESTION: 94/22 i

A reactor is operating at 80% of rated thermal power with power distribution peaked both ;

radially and axially in the center of the core. Reactor power is then increased to 100% over the next two hours using only reactor coolant boron adjustments for reactivity control.

Neglecting any effect from reactor poisons, when power is stabilized at 100%, the radial peaking factor will be and the axial peaking factor will be .

A. higher; lower B. higher; higher C. the same; lower D. the same; higher ANSWER: D.

COMMENT:

Question does not specify whether location or magnitude of peaking factors is desired. Location is referenced in the root of the question.

If magnitude is desired then an increase in the Axial Peaking Factor is a possible answer.

This would make the response "D" (answer key).

If location is desired tben a decrease in the Axial Peaking Factor would-be correct. Axial Peaking Factor is not only referred to in magnitude, but also according to which horizontal slice of the core it occurs in.

This would make the response "C".

Recommendation: Accept both "C" and "D".

Technical

Reference:

PSL Technical Specification.

l l

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FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE 4

RESPONSE

I Do not concur. The question referred to the quantitative change in the axial and radial peaking i factors, not the change in core location of the axial and radial power peaks. Even if the question was misread, the knowledgeable examinee would dismiss the assumption that the terms " higher" and " lower" referred to core location because the location of the radial power peak cannot be described in this manner. 1 Based on the interim answer key, this question was answered correctly by 47/124 examinees and yielded a moderate positive discrimination index of +0.19. No answer key chance is reauired.

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. i FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE FACILITY - SEQUOYAII EXAM - PWR FORM A/B QUESTION: 55/83 Which one of the following describes the absolute value of integral control rod worth (negative reactivity) during the complete withdrawal of a fully-inserted control rod?

A. Increases, then decreases B. Decreases, then increases C. Increases continuously D. Decreases continuously I

ANSWER: D. l l

COMMENT:

Accept C or D. )

Question requires interpretation. " Absolute value" of the " negative reactivity" provides confusion as to the intent of the direction of the enange.

C- As a rod is pulled, each step removes negative reactivity; therefore, the effect on the core of each rod step results in an integral rod worth available to the core that increases continuously during withdraw.

D- With a rod fully out, its worth is 0. The rod has maximum worth when it is fully inserted; therefore, integral rod worth in the core decreases continuously as the rod is withdrawn. The curves that are used at Sequoyah, see attached BOL Hot FulJ Power Curve, show that the ihtegral worth of a bank of rods is zero when they are fully withdrawn.

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RESPONSE

Concur. Question phrasing may have caused confusion. Additionally, several different formats are used to represent integral rod worth in the nuclear industry--some use negative reactivity units and some use positive reactivity units. Inversion of the normal means of interpreting integral rod worth would cause significant confusion at some facilities. Therefore, both C and D could be considered correct depending on the facility.

l - - _ _ _ _ _ - _ _ _ - _ . - -_.

FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE l

Based on the interim answer key, this question was answered correctly by 41/124 examinees and yielded a moderate positive discrimination index of +0.19. The answer key has been changed to accent either C or D for full credit.

FACILITY - SEQUOYAH EXAM - PWR FORM A/B l

l QUESTION: 56/84 Neutron flux shaping in a reactor core reduces radial power peaking:

A. in the center of the core caused by the high number density of fuel assemblies.

B. at the periphery of the core caused by moderator reflection of thermal leakage neutrons.

C. throughout the core caused by uneven burnout of control rod poison material. i 1

D. throughout the core caused by uneven burnout of fuel assemblies. l ANSWER: A.

COMMENT:

Accept A or D.

A- Neutron flux shaping in a core results in radial peaking in the center of the core; however, "high number density" of fuel assemblies is undefined.

D- Core design, or neutron flux shaping, promotes even core burn up which reduces radial peaking.

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RESPONSE

Do not concur. Option D states that radial power peaking is caused by uneven burnout of fuel assemblies. Westinghouse (Reactor Core Control for Large PWRs,1983, p. 8-13) states "It should be noted that one of the major effects of bumup is to establish a flatter radial profile."

This indicates that fuel bumup does not cause increased radial peaking. In fact, burnup actually flatten; the radial neutron flux distribution and reduces radial power peaking. Therefore, option D cannot be correct.

Based on the interim answer key, this question was answered correctly by 67/124 examinees and yielded a moderate positive discrimination index of +0.20. No answer key change is reauired.

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FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE FACILITY - SEQUOYAH EXAM - PWR FORM A/B QUESTION: 69/88 i

A reactor is initially operating at 50% power with equilibrium core xenon-135. Power is increased to 100% over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 actions are taken.

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

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

COMMENT:

Not enough information is provided to answer the question. Without actual .e this' question -

cannot be answered. The general guidelines for xenon transients are that the riinimum value of xenon after a power increase will occur at 4-6 hours from the time the chanr,e is started. For l a step change, this occurs closer to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; for a ramp, the minimum is closer to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The !

question deals with a ramp change; therefore, the time frame shorld be longer than the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time frame. If the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months referred to in the question is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months afte # power has reached 100%,

the peak has occurred about 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the transient started which is 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after reaching !

100 %. In another 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (which would be 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after reaching 100%), the xenon value should be very close to the value at the time the plant reached 100%. Using the plant computer which calculates xenon values, the xenon value when the plant reached 100% and the value 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months later differed by 43 pcm. Using the general guidelines for xenon transients, it is not possible for this to be determined. If the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months referred to in ne question is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from the time the power increase was initiated, the minimum value of' xenon should have been reached.

1 . However, the power increase was a ramp instead of a step power change; therefore, whether or not the minimum had been reached is difficult, if not impossible, to determine. If the minimum

value had not been reached then temperature would still be increasing, if it had been reached l then it would be decreasing. The only answer that can be ruled out is answer "D".

O FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE I

RESPONSE

Do not concur. Following a power increase, minimum core xenon will occur after 4 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from the time the final power level is reached, not from the beginning of the power change. For l a ramp change, this occurs closer to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; for a step change, this occurs closer to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

For a 50% step power increase, minimum core xenon will be reached in approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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, within 4 to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after power reaches 100%,

minimum core xenon will be achieved.

l 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 change is required.

FACILITY - SEQUOYAII 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:

Question is beyond the scope of GFES training. It involves system knowledge and plant response to transients.

RESPONSE

Partially concur. Although the conditions in the question provide an accident situation, an l examinee knowledgeable in heat transfer and thermal hydraulics should be able to readily

~ '

FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE f 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 l option that directly affects heat transfer in the core. Therefore, options A and D will be accepted.

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

i l

FACILITY - SEQUOYAH l EXAM - PWR FORM A/B l QUESTION: 94/22 ;

l A reactor is operating at 80% of rated thermal power with power distribution peaked both !

radially and axially in the center of the core. Reactor power is then increased to 100% over the I next two hours using only reactor coolant boron adjustments for reactivity control.

Neglecting any effect from reactor poisons, when power is stabilized at 100%, the radial peaking 1 I factor will be and the axial peaking factor will be .

A. higher; lower j B. higher; higher l C. the same; lower I

D. the same; higher ANSWER: D.

COMMENT:

l Accept C or D.

! Question requires interpretation. Different perspectives result in different answers.

C- " Higher" and " lower" typically imply core location; " increase" and " decrease" typically imply magnitude of a value. The axial peaking factor will increase at a lower core location.

D- The magnitude of the axial peaking factor will be " higher".

1 4

FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE

RESPONSE

Do not concur. The question referred to the qualitative change in the axial and radial peaking factors, not the change in core location of the axial and radial power peaks. Even if the question was misread, the knowledgeable examinee would dismiss the assumption that the terms " higher" and " lower" referred to core location because the location of the radial power peak cannot be described in this manner.

Based on the interim answer key, this question was answered correctly by 47/124 examinees and yielded a moderate positive discrimination index of +0.19. No answer key change is reauired.

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e e FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE FACILITY - SOUTH TEXAS PROJECT EXAM - PWR FORM A/B QUESTION: 14/42 Refer to the drawing of a gas-filled detector characteristic curve (see figure below).

What is the effect of operating a proportional neutron detector at a voltage near the high end of the proportional region?

A. Gamma pulses will increase in size while neutron pulses remain essentially the same, causing some gamma pulses to be counted as neutron pulses and yielding a less accurate neutron count rote.

B. A high gamma radiation field will result in multiple small gamma pulses that combine to form larger pulses, which will be counted as neutron pulses, yielding a less accurate neutron count rate.

C. Neutron pulses will become so large that gamma pulse discrimination is no longer ;

needed, yielding a more accurate neutron count rate. l D. The positive space charge effect will increase and prevent collection of both gamma and neutron pulses, causing a less accurate neutron count rate. j i

ANSWER: B.

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CAS FILLED DETECTOR CHARACTERISTIC CURVE 5

FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE I

l COMMENT: l The figure provided with the question does not accurately represent a Gas Filled Detector l Characteristic Curve in that it shows a disproportionality in Region III, the Proportional Region. l The detector response in Region III should show e direct proportion for each of the two I characteristic curves shown and they should parallel one another throughout this region. These I attributes are shown in the Sourcebook on Atomic Energy, Third Edition (Glasstone), page 200 and Nuclear Reactor Engineering, Third Edition (Glasstone & Sesonske), page 311. The figure

provided with the question depicts the curves as being disproportionate with respect to each other.

Because a more direct proportionality actually exists than was depicted on the exam question, operation of a detector near the high end of the proportional region would be no different than i one operating at a lower point, thus we feel the question has no correct answer.

RESPONSE

Do not concur. According to Westinghouse (Radiation. Chemistry, and Corrosion Considerations for Nuclear Power Plant Application,1983, p. 5-29), "In a high gamma field with high operating voltage, gamma pulse pile-up results in instrument output indicating a l neutron flux much higher than actually exists." This makes option B the correct answer.

Based on the interim answer key, this question was answered correctly by 23/124 examinees and yielded a very small positive discrimination index of +0.02. No answer key change is required.

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

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! FACILITY - SOUTH TEXAS PROJECT l EXAM - PWR FORM A/B !

QUESTION: 56/84 Neutron flux shaping in a reactor core reduces radial power peaking: l A.. in the center of the core caused by the high number density of fuel assemblies.

l B. at the periphery of the core caused by moderator reflection of thermal leakage neutrons.' ;

C. throughout the core caused by uneven burnout of control rod poison material. l D. throughout the core caused by uneven burnout of fuel assemblies.

i ANSWER: A.

.l COMMENT:

The answer key cited choice "A" as the correct answer: "in the center of the core caused by j the high number density of fuel assemblies." However, the " number density" of fuel assemblies is a somewhat ambiguous term depending on the area of the core or the number of fuel l assemblies being evaluated. The " number density" could be viewed as a constant for the core providing only the area actually occupied by fuel assemblies is taken into account. This was l essentially the interpretation used at our facility which was the basis for rejecting choice "A".

We feel an equally correct answer is choice "D": "throughout the core caused by uneven ,

burnout of fuel assemblies." Following the initial core load, subsequent refuelings involve I replacement of approximately one third of the core. In recognition of the difference on fuel inventory between new fuel assemblies and those remaining in the core from the previous cycle, placement within the core is chosen to promote optimum flux shaping so as to reduce radial power peaking. Thus, the fuel content (i.e., uneven burnout of fuel assemblies) is in, fact, a i

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consideration for optimizing core ^ power distribution.

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RESPONSE

Do not concur. If all fuel assemblies contained the same fuel enrichment, the center of a i refueled core would have the highest neutron flux of any location in the core. This is because L fission rate would increase geometrically from the perimeter of the core toward the center due

! to the cylindrical geometry of the core. This geometry would cause the center fuel assemblies to receive fission neutrons from the greatest number of fuel assemblies, and thereby generate the highest power and produce the highest radial peak. Neutron flux shaping using fuel loading

! patterns prevent this high peak in the center of the core.

FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE i Westinghouse (Reactor Core Control for Large PWRs,1983, p. 8-13) also states "It should be noted that one of the major effects of burnup is to establish a flatter radial profile." This indicates that fuel burnup does not cause increased radial peaking. In fact, burnup actually flattens the radial neutron flux distribution and reduces radial power peaking. Therefore, option l D cannot be correct.

I Based on the interim answer key, this question was answered correctly by 67/124 examinees and yielded a moderate positive discrimination index of +0.20. No answer key change is required.

l FACILITY - SOUTH TEXAS PROJECT l

EXAM - PWR FORM A/B l

QUESTION: 88/16 l

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 l B. Pressurizer level in the indicating range I

C. Emergency core cooling injection capability i

D. Steam generator steaming capability ANSWER: D.

COMMENT:

We feel this question is beyond ,the scope of the generic fundamentals area. This question deals with Small Break Loss of C6olant Accident (SBLOCA) analysis and is very specific in its application within that analysis. We address SBLOCA analysis in the Transient Accident Analysis and Mitigating Core Damage Courses which are presented much later in our Initial Licensed Operator Training Program.

l l RESPONSE:

! Partially concur. Although the conditions in the question provide an accident situation, an

( examinee knowledgeable in heat transfer and thermal 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 i _ -. _ - ___ __

FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE 1

' 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.

1

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 changed to accept either A or D for full credit.

1 FACILITY - SOUTH TEXAS PROJECT l EXAM - PWR FORM A/B I

QUESTION: %/24 Which one of the following will prevent brittle fracture failure of a reactor vessel?

A. Manufacturing the reactor vessel from low carbon steel l B. Maintaining reactor vessel heatup/cooldown rates within limits C. Maintaining the number of reactor vessel heatup/cooldown cycles within limits l

D. Operating above the reference temperature for nil-ductility transition (RTuor) l ANSWER: D. ,

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COMMENT: ,

i The answer key cited choice "D" as the correct answer: " Operating above the reference temperature for nil-ductility transition (RTuor).

We feel choice "B" is equally correct: " Maintaining reactor vessel heatup/cooldown rates within I limits." The limits of heatup/cooldown rates and their accompanying operating curves are also based on the prevention of brittle fracture. These limits are derived through Fracture Mechanics analysis which uses the RTuor as the basis for determining allowable operating regions of pressure and temperature. Although remaining above RTsor will certainly prevent brittle failure, it is not operationally feasible. In order to enable operation of the plant over the entire spectrum of temperature, a method must be employed to ensure brittle failure does not occur when below the RTuor. This method is the Fracture Analysis nientioned earlier and is an extension of the ;

l original analysis and testing that yielded the RTuor results. Thus, staying within the heatup/cooldown limits is the operational method employed to ensure brittle fracture does not occur.

FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE

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l RESPONSE: l 1

l Do not concur. Option B does not refer to pressure-temperature curves for heatup and l

cooldown. It refers to heatup and cooldown Ialsa (*F/hr). Simply maintaining heatup and '

cooldown rates within limits will not prevent brittle fracture. That is why the 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 chance is required.

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FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE l

FACILITY - WATERFORD EXAM - PWR FORM A/B l

QUESTION: 14/42 l

Refer to the drawing of a gas-filled detector characteristic curve (see figure below). ;

1 What is the effect of operating a proportional neutron detector at a voltage near the high end of the proportional region?

! A. Gamma pulses will increase in size while neutron pulses remain essentidly the same, l causing some gamma pulses to be counted as neutron pulses and yielding a less accurate !

neutron count rate.

l B. A high gamma radiation field will result in multiple small gamma pulses that combine to form larger pulses, which will bc ecunted as neutron pulses, yielding a less accurate neutron count rate.

C. Neutron pulses will become so large that gamma pulse discrimination is no longer needed, yielding a more accurate neutron count rate.

D. The positive space charge effect will increase and prevent collection of both gamma and neutron pulses, causing a less accurate neutron count rate. l ANSWER: B. j i

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FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE 1

I COMMENT:

l Question 14, in our opinion, is testing at a level equivalent to that which would be expected l knowledge for a maintenance technician specializing in radiation instrumentation.

! RESPONSE:

l Do not concur. This question is supported by NRC generic component K/A 191002K118,

" Theory and operation ef!on chambers,..." According to Westinghouse (Radiation, Chemistry, and Corrosion Considerations for Nuclear Power Plant Application,1983, p. 5-29), "In a high gamma field with high operating voltage, gamma pulse pile-up results in instrument output indicating a neutron flux much higher than actually exists." This makes option B the correct answer.

Based on the interim answer key, this question was answered correctly by 23/124 examinees and yielded a very small positive discrimination index of +0.02. No answer key change is required.

FACILITY - WATERFORD EXAM - PWR FORM A/B l 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 +

m D. Steam generator steaming capability ANSWER: D.

! COMMENT:

l Question 88, in essence, is asking whether or not the candidate can identify the mitigating strategy for a small break LOCA event. In order to answer this question, the candidate must know the relationship between break flow, injection flow and core cooling during a small break LOCA. Although fundamentals training provides the understanding of the relationship between

(

FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE pressure, temperature and DNBR, these are only the building blocks needed to advance to further training to understand break flow cooling mechanics. This level of knowledge would not be obtained until the candidate has taken Mitigating Core Damage or Transient and Accident Analysis training.

RESPONSE

1

! Partially concur. Although the conditions in the question provide an accident situation, an examinee knowledgeable in heat transfer and thermal 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 i

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 l accepted.

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

l l

,. , L l

I i l

FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE FACILITY - WOLF CREEK EXAM - PWR FORM A/B l

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

l l 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the power change?

l i

f A. Greater than 588'F and decreasing slowly l

B. Greater than 588 F and increasing slowly l

C. Less than 588'F and decreasing slowly 1

D. Less than 588*F and increasing slowly l ANSWER: A.

COMMENT: )

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

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

.8 x / Power Cliange 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 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , and using the second thumbrute has xenon dipping at 7.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . It is not reasonable to have examinees determine the time of the xenon dip at the accuracy required to answer this question using a thumbrule. Estimating the effects of xenon either 4 or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the power change would be more reasonable.

I.

l

l FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE l

RESPONSE

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

For a 50% increase the required time is approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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 i listed in the question, only option A is correct.

The facility comment stated that 7.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months should be accepted as the time to minimum core

! xenon based on a thumbrule in the facility lesson plan. However, the ht umbrules in the facility training mater 41 refer to the number of hours to reach 2:ak 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 l xenon following a power increase.

l 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 change is required.

FACILITY -- WOLF CREEK EXAM - PWR FORM A/B l 1

l QUESTION: 88/16 i l

Which one of the following must be present to prevent departure frem nucleate boiling from occurring in a reactor core following a pressurizer vapor space instru'nent line rupture if the leak i rate is less than normal makeup capability?

l A. Reactor coolant pump flow capability B. Pressurizer level in the indicating range l

C. Emergency core cooling injection capability j D. Steam generator steaming capability ANSWER: D.

l 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 l "A" should also be considered correct because having RCP flow will increase the margin to DNB.

2-

j '. FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE i

l The attached training material states how flow oscillations, due to no forced flow, can lower the l Critical Heat Flux required for DNB by as much as 40%. Also, Technical Specifications tout l flow as being important when considering DNB.

I j RESPONSE:

Concur. An examinee knowledgeable in heat transfer and thermal 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 i 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 i yielded a small positive discrimination index of +0.09. The answer key has been changed to accept either A or D for full credit. I 1

FACILITY - WOLF CREEK t

1 EXAM - PWR FORM A/B l l

QUFSTION: %/24 Which one of the following 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,s refe.rence temperature for nil-ductility transition (RTay)

ANSWER: D.

COMMENT:

This question deals with the prevention of brittle fracture. Answer "D" is stated 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).

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

FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE

RESPONSE

Do not concur. Option B does not mention maintaining pressure and temperature in accordance 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 required.

l l

'l '

FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE FACILITY - ZION EXAM - PWR FORM A/B QUESTION: 42/70 An operator must never open or close a high voltage (greater than 750 Vac) air break disconnect l unless:

A. a parallel path exists for current flow.

B. the circuit it is in is already deenergized. :

C. the current flowing through it is approximately zero.

D. the current flowing through it is less than its design current carrying capability.

ANSWER: B.

COMMENT:

Concerning the operation of disconnects. We feel that distracter C is also correct. At Comed, there are disconnects that are operated when energized from one side (with little or no current flow). These are typically disconnects that isolate OCBs (switchyard breakers). With the OCBs open, an isolating disconnect may be energized from the downstream side and isolated on the upstream side with the OCB itself. This condition matches distracter C.

RESPONSE

Concur. There are situations in which disconnects may be operated when energized, but with ,

very small currents (approximately zero) flowing through them. Therefore, option C is also !

correct.

Based on the interim answer key, this question was answered correctly by 109/124 examinees and yielded a small negative discrimination index of -0.06. The answer key has been changed j to accept either B or C for full credit.

i 1

4 l

FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE O

j FACILITY - ZION 1

l EXAM - PWR FORM A/B QUESTION: 60/88 A reactor is initially operating at 50% power with equilibrium core xenon-135. Power is increased to 100% over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 actions are taken.

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the power change?

A. Greater than 588 F and decreasing slowly B. Greater than 588 F and increasing slowly C. Less than 588'F and decreasing slowly 1

D. Less than 588'F and increasing slowly 1

ANSWER: A. I l

COMMENT: I l

i This question concerns Xenon-135 peaking during an up-power transient. The maximum decrease (dip) in Xenon concentration occurs about 4 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the transient (Ref. 1 Pressurized Water Reactor Core Control; Westinghouse, Chapter 4, page 4-26). At the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months point, it's difficult to determine exactly what Xenon is doing. It may just be reaching the ;

maximum dip, thus makino, distracter B correct.

RESPONSE':

Do not concur. As stated in the facility comment, following a power increase, minimum core xenon will occur after 4 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . But this 4 to 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months time period begins at the time the final power level is reached, not from the beginning of the power change. For a ramp change, this occurs closer to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; for r. step change, this occurs closer to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . For a 50% step power increase, minimum core xenon will be reac;1:d in approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 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 j change listed in the question, at a time less than 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after power reaches 100%, minimum core xenon will be achieved, making only option A correct.

l l

? ,,

's -

FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE t 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 change is required.

1 FACILITY - ZION EXAM - PWR FORM A/B QUESTION: 70/98 A reactor is operating with the following initial conditions:

Power level = 100%

Coolant boron = 620 ppm Average coolant temperature = 587 F After a load decrease reactor conditions are as follows:

Power level = 80%

Coolant boron = 630 ppm Average coolant temperature = 577 F Given the following values, how much reactivity was added by control rod movement during the load decrease? (Assume fission product poison reactivity does not change.)

Total power coefficient = -1.5 x 10 2% AK/K/%

Moderator temperature coefficient = -2.0 x 10-2% AK/K/ F Differential boron worth = -1.0 x 10-2% AK/K/ ppm A. -0.2% AK/K I 1

i B. +0.2% AK/K C. -0.4% AK/K s s

D. +0.4% AK/K ANSWER: A.

I COMMENT:

This is a reactivity balance question. Changes in moderator temperature occur with any power ramp due to changes in power defect. If moderator temperature is maintained at the program value, then the changes in temperature produce no reactivity effect outside of that considered for j Total Power Defect. This question does not indicate whether the temperature change is FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE consistent with the program for Tave or whether this is an isothermal temperature change in addition to that caused by Total Power Defect. If moderator temperature change was an isothermal change in addition to that caused by power defect, then distracter C is correct.

RESPONSE

Do not concur. The normal method for performing a power change requires RCS average coolant temperature to stay "on program". Power coefficient includes the reactivity effect of the power change with the assumption that average coolant temperature remains on the program.

It would be unwarranted and incorrect for an examinee to assume that average coolant temperature had varied from the program.

l Based on the interim answer key, this question was answered correctly by 111/124 examinees ]

and yielded a small positive discrimination index of +0.15. No answer key change is reauired. l l

FACILITY - ZION 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 l B. Pressurizer level in the indicating range 1 l

C. Emergency core cooling injection capability D. Steam generator steaming capability

.s . .

ANSWER: D.

COMMENT:

This question is outside the scope of Generic Fundamentals. This more appropriately belongs t to a Mitigating Core Damage course, which for ns is taught after the students have learned systems. This question requires the student to be familiar with the operation of the different systems in the plant - in particular the conditions necessary to provide adequate core cooling i with Emergency Core Cooling Systems.

i t

I

=

'i -

FACILITY COMMENTS AND NRC RESPONSES FOR TIIE OCTOBER 1996 GFE

RESPONSE

Partially concur. Although the conditions in the question provide an accident situation, an examinee knowledgeable in heat transfer and thermal 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 tafer 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 changed to accept either A or D for full credit.

FACILITY - ZION EXAM - PWR FORM A/B QUESTION: 94/22 A reactor is operating at 80% of rated thermal power with power distribution peaked both radially and axially in the center of the core. Reactor power is then increased to 100% over the next two hours using only reactor coolant boron adjustments for reactivity control.

Neglecting any effect from reactor poisons, when power is stabilized at 100%, the radial peaking factor will be and the axial peaking factor will be .

A. higher; lower l B. higher; higher

! C. the same; lower t .

' D. the same; l6gher -

ANSWER: D.

COMMENT:

The condition postulated in the question is unusual in that the flux is centered axially at 80%

power. A logical assumption may be made that the above condition is due in part to greater fuel depletion in the core bottom. Therefore, the final 20% power increase could in fact reduce the axial peak-to-average ratio making answer C correct.

l

'l FACILITY COMMENTS AND NRC RESPONSES FOR THE OCTOBER 1996 GFE

RESPONSE

i Do not concur. Even if the fuel in the lower half of the core was more depleted than the fuel ;

in the upper half of the core, the power increase would result in an upper core coolant l temperature increase that is greater than the temperature increase in the lower portion of the core. This would add comparatively more negative reactivity to the upper portion of the core, ;

resulting in a shift in power prodt.ction from the upper and central portions of the core into the lower portion of the core. Concentrating more power production into a smaller portion of the ]

core will result in a larger axial peak and axial peaking factor. l Based on the interim answer key, this question was answered correctly by 47/124 examinees and yielded a moderate positive discrimination index of +0.19. No answer key change is required.

1 1

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d P

1 l

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Y 5-

V e ., ,

l

ATTACHMENT 3 y A4 I

f l

1 1

\

i,, , l S *00 FINAL ANSWER KEY *00 OCTOBER 1996 GENERIC FUNDAMENTALS EXAM BOILING WATER REACTOR - ANSWER KEY FORM ANS FORM ANS FORM ANS FORM ANS A B A B A B A B 1 29 B 26 54 D 51 79 C 76 4 D 2 30 B 27 55 C 52 80 B 77 5 C 3 31 C 28 56 A 53 81 C 78 6 B 4 32 A 29 57 C 54 82 A 79 7 D 5 33 D 30 58 C 55 83 A 80 8 A 6 34 C 31 59 D 36 84 C 81 9 B 7 35 B 32 60 B 57 85 B 82 10 DELETED 8 36 B 33 61 C 58 86 D 83 11 B 9 37 A 34 62 D 59 87 D 84 12 A 10 38 D 35 63 C 60 88 D 85 13 D 11 39 C 36 64 D 61 89 C 86 14 A 12 40 B 37 65 C 62 90 D 87 15 A 13 41 D 38 66 D 63 91 B 88 16 B 14 42 A 39 67 A 64 92 A 89 17 A 15 43 B 40 68 C 65 93 D 90 18 D 16 44 3 41 69 A 66 94 D 91 19 A l 17 45 B 42 70 A 67 95 B 92 20 D 18 46 D 43 71 B 68 96 C 93 21 C 19 47 A 44 72 A 69 97 C 94 22 C 20 48 A 45 73 C 70 98 D 95 23 A l 21 49 D 46 74 A 71 99 A 96 24 B 22 50 D 47 75 D 72 100 B 97 25 A 23 51 C 48 76 C 73 1 D 98 26 A )

l 24 52 B 49 77 A 74 2 B 99 27 B l

25 53 C 50 78 B 75 3 A 100 28 B l

l l

l?

- - FINAL ANSWER KEY *** i

! l l OCTOBER 1996 GENERIC FUNDAMENTALS EXAM i PRESSURIZED WATER REACTOR - ANSWER KEY FORM $NS FORM ANS FORM ANS FORM ANS A B A B A B A B 1 29 B 26 54 A 51 79 D 76 4 C i 2 30 D 27 55 A 52 80 C 77 5 A 3 31 A 28 56 C 53 81 A 78 6 C l

4 32 A 29 57 A 54 82 B 79 7 B 5 33 C 30 58 A 55 83 C/D 80 8 A l 6 34 C 31 59 D 56 84 A 81 9 D 7 35 C 32 60 B 57 85 D 82 10 D 8 36 B 33 61 B 58 86 C 83 11 D l 9 37 A 34 62 P 59 87 C 84 12 B 10 38 D 35 63 A 60 88 A 85 13 A l 11 39 A 36 64 C 61 89 B 86 14 B 12 40 D 37 65 B 62 90 B 87 15 D l

13 41 C 38 66 C 63 91 C 88 16 A/D 14 42 B 39 67 C 64 92 D 89 17 D ;

15 43 C 40 68 A 65 93 D 90 18 A 16 44 8 41 69 A 66 94 A 91 19 C.

17 45 C 42 70 9 'C 67 95 D 92 20 C 18 46 B 43 71 D 68 96 B 93 21 A 19 47 B 44 72 B 69 97 8 24 22 D 20 48 C 45 73 C 70 98 A 95 23 B 21 49 C 46 74 B 71 99 D 96 24 0 22 50 D 47 75 D 72 100 B 97 25 B 23 51 A 48 76 D 73 1 D 98 26 B 24 52 D 49 77 B 74 2 B 99 27 C 25 53 A 50 78 B 75 3 A 100 28 A

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