ML20212L031

Exam Rept 50-254/OL-86-04 for Both Units of Exam Administered on 861014,1203-04 & 16-17.Exam Results:Four Senior Reactor Operator (SRO) & One Reactor Operator Candidates Passed & One SRO Failed Written Exam

ML20212L031
Person / Time
Site:	Quad Cities
Issue date:	01/16/1987
From:	Bishop M, Burdick T, Dave Hills, Morgan T, Sherman J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20212L007	List: ML20212L005 ML20212L031
References
50-254-OL-86-04, 50-254-OL-86-4, NUDOCS 8701290292
Download: ML20212L031 (108)

Text

. . , .

U.S. NUCLEAR REGULATORY COMMISSION REGION III -

Report No. 50-254/0L 86-04 Dockets flo. 50-254; 50-265 License Nos. DPR-29; DPR-30 Licensee: Commonwealth Edison Coropany Post Office Box 767-Chicago, IL 60690 Facility Name: Quad Cities Nuclear Power Station Examination Administered At: Quad Cities Nuclear Power Station Examinaticn Conducted: October 14, 1986; Dacember 3-4, 16-17, 1986 5- &

Examiners: D. Hills , _ _/,

Date T o I

b, ate M. Bishop Date

/ ([M g.7/My' J. Sherman ba=tc i

\ I lb %7 Approved By: T '. . Burdick, C ief 1 Operator Licensing Section D,a_te_

Examination Summary Examination administered cn October 14,,~1986,, December 3-4,,"16-1L~1986~~~

~ ~~ ~~

JReport Nos. Sb7M/DH6~ ~0'4)

Written and cral examinations were administered to five Senior Reacter Operators (SRO)candidatesandcneReactorOperator(RO) candidate.

Results: Four SR0 candidates and one R0 candidate passed these examinaticns.

D'ne' SRO candidate failed the written portion of the examination and therefore, was not administered the oral portion.

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REPORT DETAILS _

1. Examiners D. Hills, Chief Examiner M. Bishop T. Morgan J. Sherman

2. Examination Review Mee_t_ing Specific facility connents concerning written examination questions, followed by the NRC response, are enumerated in the Attachment.

3. Exit Meeting At the conclusion of the written examination, an exit meeting was conducted. The following personnel attended this exit meeting.

Facility Representative R. Bax, Station Manager T. Tamlyn, Production Superintendent R. Robey, Services Superintendent G. Tietl, Assistant Superintendent W. Graham, Principle Instructor M. Kooi, Regulatory Assurance R. Svaleson, Instructor, General Electric NRC Representatives C. Hehl, Chief, Operations Branch D. Hills, Operator Licensing Examiner R. Lanksbury, Operator Licensing Examiner Since requalification examinations were conducted concurrently with these replacement examinations, the discussions, for the most part, concerned requalification program evaluation guidelines and performance. In addition, the inconsistency in requirements for whole body counts for visiting NRC examiners was discussed. The licensee was advised that in the future the exarainer should be infonned of the requirements to be implemented prior to the examination trip. Thus, sufficient time can be appropriately figured into the examination schedule.

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ATTACHMENT QUAD CITIES 1 & 2 REACTOR OPERATOR AND SENIOR REACTOR OPERATOR REPLACEDENT EXAMINATION REACTOR OPERATOR AND SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION OCTOBER 14, 1986 1.12 Facility Comment:

Another answer that should be acceptable for this question is to maintain proper reactor water level to prevent carry under. The Reactor Vessel and Internals text lists carry under as a factor that will lower recirc pump NPSH. Pages 14 and 15 of this text are attached. (Also applicable to replacement question 5.11 and requalification questions 1.06 and 5.05).

Resolution:

Cross referenced to R0 1.12 & SRO 5.11 Replacement R0 1.06 & SR0 5.05 Requalification Comment noted but not accepted. The items requested to be included are already referred to in the answer, i.e. maintain proper water level to prevent carry under. This is addressed with the answer referring to low reactor vessel water level trip.

No change to the answer key will be made.

1.14 Facility Comment:

QCNPS is currently using the General Electric BWR Academic Series Heat Transfer and Fluid Flow for instruction in this area. This material includes " duration of power increase" as a factor for PCI failures and should also be considered an acceptable answer, as it is synonymous to the items listed on the key. A copy of page 9-47 is attached. (Also applicable to replacement question 5.13 and requalification question 1.08.)

Resolution:

Cross referenced to R0 1.14 & SRO 5.13 Replacement R0 1.08 Requalification Comment noted and accepted. " Duration of power increase" will be added to the answer key as an additional correct answer. Answer point breakdown remains unchanged. Page 9-47 will be added to the reference along with QC.LIC-0201-2, Nuclear Fuel Lesson Plan page 20.

2.03.c. Facility Comment:

This question is extremely confusing because it is not clear what the status of the flow converter upscale trip unit is. The question

states that the units are upscale, yet also states that the flow is less than 108%. The trip setting for this at QCNPS is 110%. If the student assumes an actual upscale trip is present, the answer key is correct in specifying a rod block exists, however, if the student merely assumes it is a high reading and a trip does not exist, since the setpoint has not been reached, then no trip is the correct response. Apparently several candidates questioned the examiners on this and were left with different impressions. In this case, either no trip or rod block should be considered correct, however, half scram should be considered erroneous. Page 4 of the APRM lesson plan is attached to describe the trip. (Also applicable to replacement question 6.01 requalification questions 2.02 and 6.01.)

Resolution:

Cross referenced R0 2.03 & SR0 6.01 Replacement R0 2.02 & SRO 6.01 Requalification Comment is accepted. The answer key will be changed to "no reactor protective system action" as would be the case if the flow units failed to the 108 or 107% position. When questioned during the exam it was stated the flow units fail upscale to 108%. Point breakdown remains unchanged.

2.06 Facility Comment:

Parts of this answer are in error and another auto start signal at 2000 R/hr drywell radiation should be added. The reactor building ventilation trip setting is 3 mR/hr, vice 10 mR/hr, and the low reactor water level setting is +8", vice -8". Pages 34 and 46 of the Primary and Secondary Containment text are attached.

Resolution:

Comment is accepted. The answer key will be changed to reflect the answers in the material for the Primary and Secondary Containment vice the answers from the reference material for SBGT system lesson plan. The total number of possible answers will be increased from five to six and the setpoints will reflect the change. The point breakdown will not be effected. The question will be reworded to reflect the six possible answers. QC.LIC-1600-1, Primary-Secondary Containment lesson plan page 34 and 46 will be added to the reference.

The original reference will be deleted due to the conflicting information.

2.07.b. Facility Comument:

This question asks what may be calculated by turning the switch to COUNT. This may be answered by stating that the number of LPRMs inputting to the RBM may be calculated. Stating that 1 LPRH = 1 volt is not asked for and should not be required. (Also applicable to requalification question 2.03) l 2

Resolution:

Cross referenced to R0 2.07.b Replacement R0 2.03.b Requalification Comment accepted. Because the HOW was not asked for, "Can be calculated by using 1 volt per operating input" will not be required for full credit. The point value for the second portion of part 'b.' will be decreased by 0.5 points. Section and test values for both the R0 replacement and requalification will be adjusted accordingly.

2.09 Facility Comment:

This answer key assigns a point valve of 0.5 pts to the statement

" coolant and boron" when discussing improper mixing. If a car.didate discusses improper mixing, he should not be penalized this point value for not saying " coolant and boron", since this is implied in the discussion of improper mixing. (Also applicable to requalification question 2.05).

Resolution:

Cross referenced to R0 2.09 Replacement R0 2.05 Requalification Comment is not accepted. It cannot be assumed that the candidate knows what is being mixed 1.e., coolant and boron or hot and cold water etc. Therefore, the answer key will not be changed.

3.07.a. Facility Comment:

" Low speed limit" should not be required in this answer. The question asks for the component that is limiting and stating the

" Master Limiter" sufficiently answers this question. (Also applicable to replacement question 6.09 and requalification question 6.06.)

Resolution:

Cross referenced R0 3.07 & SR0 6.09 Replacement SR0 6.06 Requalification Comment is accepted. The answer key will be changed to indicate that " Low speed limit" is not required for full credit. The point value is unchanged.

3.10 Facility Comment:

PCIS Group I isolation" is listed in e. of this answer and should actually appear in f. of the answer instead where PCIS actuations are asked for. (Also applicable to requalification question 3.06).

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Resolution:

Cross referenced R0 3.10 Replacement R0 3.06 Requalification Connent is accepted. "PCIS Group Isolation will be moved to part f.

Also after further investigation it was noted that some of the ECCS systems had been omitted. LPCI Loop Select Logic will initiate.

EDG will start, and CS, after 8.5 minutes, will start at -59".

These three items will be added to the answer for part e. The question point value remains unchanged. The following will be added to the reference QC.LIC-1400 Core Spray page CS-13, QC.LIC-1000-1 RHR page 42.

3.11.b. Facility Comment:

The exact title of this alarm tile should not be required. If a student states that the indication is an alarm on his control panel that should be sufficient. (Also applicable to replacement question 6.10 and requalification questions 3.07 and 6.07)

Resolution:

Cross referenced to R0 3.11 & SR0 6.10 Replacement R0 3.07 & SR0 6.07 Requalification Comment not accepted due to the fact that there are several alarms associated with the Control Room H&V, including a major trouble alerm for the standby H&V system. Therefore, it is required the operator know this specific alarm.

4.04 Facility Comment:

This question asks for the immediate action steps that would be taken to control the reactivity addition. Parts 5 and 6 of the answer go beyond the required actions to control the reactivity.

(Also applicable to replacement question 7.04 and requalification questions 4.04 and 7.04)

Resolution:

Cross referenced R0 4.04 & SR0 7.04 Replacement R0 4.04 & SRO 7.04 Requalification Comment is accepted. Due to the way the question is worded, parts 5 and 6 will be deleted from the answer key. The .75 points assigned to the deleted answer will be redistributed throughout the remaining answers. 0.25 will be added to part 3. for "even if the sequence calls for stopping at an intermediate position" .0.5 will be added to part 4., 0.25 for "if a LSSS has not been exceeded and 0.25 for "if reactor power continues to increase."

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4.06 Facility Connent:

This question asks for three actions to be performed and one of the answers listed in the key is " refer to the abnormal procedure". The actions listed in this procedure should also be considered correct.

Therefore, insert the drive to its full in position and electrically disarm the drive in and drive out solenoids should be added to the answer key. A copy of Q0A 300-3 (INABILITY TO DRIVE A CONTROL ROD; UNC00 PLED CONTROL ROD) is attached. (Also applicable to requalification question 4.06).

Resolution:

Cross referenced R0 4.06 Replacement R0 4.06 Requalification Coment is accepted. Since the question did not specify actions required by QGP 1-2 and these two answers are Immediate Operator Actions in Q0A 300-3, these two answers will be added to the answer key for both the Replacement and Requalification exam. Point value remains unchanged.

4.08 Facility Comment:

To answer this question to the answer key requires, to a certain extent, the memorization of several form numbers or names that are in the rod sequence book provided to the operator. This memoriza-tion is not necessary or encouraged at QCNPS. In answering this question the following points are provided to allow candidate responses to better be evaluated and credit given appropriately.

Copies of QTP 1600-S2 and S3 are attached as well as a copy of QGP 4-1,

1. QTP 1600-S2 Control Rod Sequence Review Sheet is the first sheet of the rod sequence book. If candidates describe signature requirements on the first page of the book, this is the form that is being referred to. Answers 1 and 2 of the answer key are affected by this.

2. QTP 1600-S3 Control Rod Sequence Sheet is the form used for the bulk of the book. Each page describes the movement of a step. This form is referred to in answer 3.

3. A caution on page 6 and page 8 of QGP 4-1 states the NS0's responsibility in verifying proper movements and documentation.

This is accomplished on QTP 1600-S3, as described in QGP 4-1, and should also be considered as a correct response.

4. QTP 1600-S1 through S5 have been attached so the examiners grading these exams may better understand QCNPS's control on rod sequencing and fairly evaluate candidate responses. (Also applicable to replacement question .7,07 and requalification question 7.05) 5

Resolution:

Cross referenced to R0 4.08 & SR0 7.07 Replacement SR0 7.05 Requalification Coment is partially accepted. The comment on the candidate memorizing from numbers is a valid point but was not required for credit. Portions of the answer not required for credit are inclosed in parenthesis, the form numbers are so marked. Item 1 is accepted. Credit will be given if the candidate specifies "the first sheet of the rod sequence book".

Item 2 is not accepted because the question is looking for the approvals not the description of the movements of a step. Item 3 is not accepted because the question is asking for verification before rod movement,not checks during rod movement. Item 4 is noted.

4.09 Facility Comment:

The answer key lists the six items from QAP 300-2 but fails to include the items addressed in the note immediately after the six items. This note concerns non-professional reading material and radios or television. These should be considered appropriate answers. A copy of page 7 of QAP 300-2 is attached. (Also applicable to requalification question 4.08)

Resolution:

Comment is partially valid since the referenced NOTE gives examples of actions which would distract the operators and is forbidden in item (3). Answer No. 3 will be modified to read as follows:

(3) Card playing, games, or other distractions from prescribed duties.

(Will also accept " reading Non-Professional reading material" or

" Utilizing unauthorized radios or TV" instead of "other distractions").

Credit will not be given for "other distractions" and the items in parenthesis .

4.12 Facility Comment:

Items 2, 6, and 7 of the answer to this question are portions of a group I isolation. A response of verifying a group I isolation should, therefore, be sufficient for these three items. Also, two other automatic actions occur which should be considered as correct

responses even though they are not listed in this QOA. These actions are

1) Off-gas sample vial box suction valve closes, and

2) off-gas line drain valves (5408's) close. Page 8 of the Process Radiation Monitor System text is attached. (Also applicable to replacement question 7.05.)

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Resolution:

Cross referenced to R0 4.12 & SR0 7.05 Replacement Comment is not accepted, because: One the procedure referenced does not list the automatic actions as verify a group 1 isolation and it is not assumed the candidate is aware of items 2, 6 and 7 when he/she state verify group 1 isolation. Two the alternate answers requested are not in the procedure and, therefore, will not be accepted and if list will be counted as wrong additional information.

5.07 Facility Comment:

In part a of this answer, the second sentence should not be required for full credit. The question is adequately answered with the first sentence.

In part b, the change may be attributed to an increase in slowing down length vice an increase in thermal diffusion length.

In part c, a discussion of increased leakage, due to the increased void content, is also acceptable. (Also applicable to replacement 4

question 1.08 and requalification questions 1.03 and 5.02)

Resolution:

Cross referenced to SR0 5.07 and R0 1.08 Replacement SR0 5.02 & R0 1.03 Requalification Conment is partially accepted. Comment for part a is valid and the answer key will be changed to indicate the second sentence is not required for full credit and the point value will be redistrib-uted to the remainder of the part a answer. The comment for part b is not valid, the control rods absorb thermal neutrons not intermediate neutrons as explained by the reference material. The comment for part c is noted but no change to the answer key is required. Paraphrasing is acceptable as long as the candidate understands the effects.

5.09.a. Facility Comment:

The term APLHGR is considered acceptable in place of MAPLHGR (actual) at QCNPS and credit should not be taken off if the term is used.

Apparently this is acceptable to the examiner since APLHGR is used in part b. of this question. (Also applicable to requalification question 5.03)

Resolution:

Cross referenced to SRO 5.09 Replacement SR0 5.03 Requalification Connent is accepted. APLHGR will be accepted as the same meaning as MAPLHGR actual.

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6.02 Facility Comment:

A lock-up of the pneumatic feed reg valve is affected by denergizing an air supply solenoid valve on its air supply. While a loss of power to this solenoid is not considered a lock-up signal, it will cause the valve to lock-up. Answers addressing this should not be considered incorrect. Q0A 600-2 is attached to describe this as well.

Resolution:

Comment is accepted. The answer supplied in the reference material will be added to the answer key as another required answer for full credit. The point value will be increased by 0.25 for the part of the answer and section and test point values will be adjusted accordingly.

6.04 Facility Comment:

Part a of this question should be removed from the test. The information for this question was found in the " Instructor's Personal Notes" section of the lesson plan and is presented to generate student interest in the subject only. Students are not provided with a copy of these pages. In our course we provide students with texts of the content / skills section. Here a reference is made of these, but even at that they are not listed and only mentioned that they may be found in the FSAR. This is in the background section of the text and is not information a student is required to know. Pages 2 and 3 of this lesson plan are attached.

In part c of this question the first two portions are erroneous since the station has implemented the E0P's. While the lesson plan still indicates material from the former QGA's, it is not an operational guidance and the correct answer should be as directed by the QGA 100 block. The QGA direction takes precedence over anything in a lesson plan. Page 3 of QGA 100-3 is attached.

Finally, the last part of c sounds as if we have only two ARI buttons.

In fact we have two pairs of buttons, either or both pairs may be utilized. Actions that occur when these buttons are actuated are described in the ATWS lesson plan as referenced, however, the fact that they remain energized for 30 seconds is less significant than the action they cause. Saying that the valves remain energized for 30 seconds should not be required. Instead this credit should be placed on what occurs when the buttons are actuated to effect rod insertion. Each channel (pair of buttons) will energize three solenoid operated valves to remove air from the scram air header causing the scram inlet and outlet valves to open and instrument volume vents and drains to close. (Also applicable to requalification question 6.03.)

RESOLUTION:

Cross reference to SR0 6.04 Replacement SRO 6.03 Requalification 8

Comments partially accepted. For part a. This comment is not accepted because it is stated in the lesson plan that attention will be focused on transient situations which have a relatively high expected frequency of occurrence. It then states the seven transients listed in the answer key are the ones that attention will be focused on. This is considered part of the design bases for the ATWS system. For part c. , regarding the first two portions this comment,is not accepted because QGA 100 states the same information. However, the point distribution will be changed due to examiners error in locating the partial credit. As to the comments on the last part of c, they are partially accepted. That portion referring to the implication of only two valves is wrong. The question did not ask how many pushbuttons are available but how it is initiated, and this is where two pushbuttons must be depressed, for a division, simultaneously. The portion regarding the " energized for 30 seconds" is accepted and will not be required for full credit. The portion regarding the expected actions is accepted. The following will be added to the answer key _for part c and will be required for full credit: "To remove air from the scram air header and cause the scram inlet [0.25] and outlet valves to open [0.25] and the instru-ment volume vent and drains to close [0.25]." The question and section point valves will be adjusted accordingly. Page 2 will also be added to the reference.

6.06 Facility Comment:

If this question is to be kept in an exam bank, the term " bypass" in part c should be changed to " equalizing" to be more consistent with QCNPS terminology and to prevent-future examinee confusion. (Also applicable to requalification question 6.04.)

Resolution:

Cross referenced to SR0 6.06 Replacement SR0 6.04 Requalification Comment noted. The word bypass will be replaced with equalizing to eliminate future confusion.

6.07.b. Facility Comment:

Most examinees will interpret this question as a normal power reduction of 10% power. In this case the EHC system will respond to produce the following correct answers:

1. 40%

2. 932 psig

3. 1800 RPM

4. 0%

5. 40% (Not applicable to requalification question 6.05.)

These answers are correct and should be accepted. Due, however, to interaction with the examiners, some examinees may have responded as described in the key. For this reason these answers should be 9

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acceptable for full credit with the exception of b.2. In-this case, the reactor pressure would remain at 935 psig. The close biase is not significant enough to retard the opening of the bypass valves to cause reactor pressure to increase as described in the answer key.

This scenario however, does not depict normal QCNPS operation since it assumes the plant is not paralleled into the Commonwealth Edison System, thereby allowing drastic turbine RPM changes. (Also applicable to requalification question 6.05.)

Resolution:

Cross referenced to SR0 6.07 Replacement SR0 6.05 Requalification Coment is accepted. Due to the confusion of the candidates on this question the question will be graded in accordance to the candidates assumptions. If the candidate answers assuming load rejection the existing answer key will be utilized, if the candidate assumes a power reduction than the facility answers will be used. If the candidate assumes the unit is still paralelled with another facility an additional answer will be generated and added to the key.

Additionally the word " reduction" will be replaced with " reject" and a statement of the only unit on the grid will be added to the question preface. The point value remains unchanged.

7.01.b. Facility Comment:

Some students may have responded that the first choice for drywell temperature is from the SPDS (Safety Parameter Display System) vice the process computer. This is an acceptable answer because the SPDS drywell temperature is from those two computer points. (Also applicable to requalification question 7.01.b.)

Resolution:

Cross referenced to SR0 7.01 Replacement SR0 7.01 Requalification Coment accepted. SPDS will be added to the answer key for an alternate answer to this question. No change to the point value is required.

7.09 Facility Comment:

Pressure should also be an acceptable answer to a.1. of this question. This exact scenario has been encountered at QCNPS before and the pressure increase has been noted. This is also addressed in the very paragraph of QOP 1000-5 that this question was taken from.

Page 2 of this procedure is attached to show this. (Also applicable to requalification question 7.06.)

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Resol tion:

Cross referenced SRO 7.09 Replacement SRO 7.06 Requalification Comment is not accepted. Pressure or re-pressurization will not work in the statenent made in the question. The procedure states after re-pressurization ha: occurred to monitor for metal temperature in-crease etc., then stratification is present.

In part a of this question, high pressure scram should be considered qO a correct response. This LSSS is listed directly next to the high 0,

pressure safety limit and does apply to it. This is shown on page 1.2/2.2-1 of the technical specifications, which is attached.

Concerning the relief valves, the relieve valves are provided to prevent the over pressurization of the vessel. These valves are sized to prevent lifting the safeties when the turbine trips from full power and the bypass valves fail to open. With this in mind, an examinee could easily be misguided by the wording in part b, of this question which asks for a " backup" to the protective actions in

a. For these reasons, relief valves should be an acceptable answer in a., if the safeties are identified in b. To verify this refer to page 8 of the Main Steam System text and page 1.2/2.2-2 of technical specifications, which are attached. ( Also applicable to requalifica-tion question 8.02.)

Resolution:

Cross referenced to SR0 8.02 Replacement SR0 8.02 Requalification Consnent is accepted. Any two of the three scrams now listed in the answer key will be accepted for the answer to part a. If high pressure scram is listed in part a, then credit will be given for relief valves in part b.

8.03.a. Facility Consnent:

With regard to the station management approval, this answer is correct with respect to technical specification 6.2.d. It should be noted, however, that QAP 1100-7 gives guidance in obtaining approval for a temporary procedure change and is more restrictive.

Either answer should be acceptable since no reference was given in the question. A copy of QAP 1110-7 and QAP 1100-T1 are attached.

Also concerning this question, two items are asked for. The approval discussed above is one, however, the review required by On-Site Review and the approval by the Station Manager should be treated as two separate items, with either or both being acceptable.

Again, see attached copy of QAP 1100-7.

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Resolution:

The comment is accepted. In lieu of the procedure supplied by the facility, that was not originally supplied to the examiners, the answer key need not be changed because the administrative procedure is more detailed than the answer key and T.S. as originally referenced.

The answer key will be changed, however, in that instead of two possible answers the answer key will be broken down into three possible answers separating On-Site review and Station Superintendent as two possible answers. No change to the point breakdown is needed.

8.04 Facility Comment:

This question does not specify a procedure or circumstance. In this light, several other answers should be acceptable as indicated below with their references. These references are attached.

1. Any GSEP condition: reference QAP 300-2 page 11

2. Flooding: reference Q0A 010-4 page 2

3. Fire: reference QEP 340-5 page 3

4. Scram: reference QAP 300-1 page 5

5. When relieving the SCRE during normal plant operation:

reference QAP 300-1 page 6.

(Also applicable to requalification question 8.03)

Resolution:

Cross referenced to SR0 8.04 Replacement SR0 8.03 Requalification Comment is accepted. The additional five conditions will be added to the answer key. The five procedures listed will be added to the reference. The point breakdown remains unchanged.

8.06 Facility Comment:

This question references QAP 300-3 which in turn references Technical Specifications figure 6.1-3, both of which are attached.

While these differ slightly either should be acceptable. (Also applicable to requalification question 8.04)

Resolution:

Cross referenced SR0 8.06 Replacement SRO 8.04 Requalification Comment is not accepted, because the Technical Specifications do not address the position as required and the question ask for "at normal power conditions" not with "there is fuel in both reactors."

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4 e e 8.08 Facility Comment:

Initial license candidates are not given specific instruction on 10 CFR 55. At QCNPS, individual licensees do not control these aspects of their licenses. Instead, this is considered an administrative requirement that is enforced by plant management when such a situation arises. A great deal of similar detailed requirements exist on licenses and renewals that are not considered individual licensee responsibilities at QCNPS, instead, the training, administrative and operating department management ensure all requirements are met. This should be taken into consideration when gradin (Also applicable to requalification question 8.06) g this question.

Resolution:

Cross referenced to SR0 8.08 Replacement SR0 8.06 Requalification Comment is accepted. Since the facility does not hold the candidates responsible for this license condition this question will be deleted from both the replacement and requalification exams and point values adjusted to reflect the deleti'on.

8.11.a. Facility Comment:

Full credit should be given to an answer that adequately describes the difference between the cards. Detailed color and border schemes should not be required.

Resolution:

Comment is accepted, because paraphrasing is acceptable as long as the candidate describes the card in sufficient detail to answer the question.

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

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U. S. NUCLEAR REGULATORY COMMISS ) l 0 REACTOR OPERATOR LICENSE EXAMINATION FACILITY: QUAD CITIES 1&2 REACTOR TYPE: BWR-GE3 DATE ADMINSTERED: 86/10/14 EXAMINER: BISHOP. M.

CANDIDATE INSTRUCTIONS TO CANDIDATE:

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

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

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 24.50 24.hh 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW

. .i a go

-00 25.et 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 -2 4 . ^0 3. INSTRUMENTS AND CONTROLS 25.00 2+-6N 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL C

100.h Totals All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature g

• Og

s. ,

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.

9 ., Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and' place finished answer sheets face down on your desk or table.

12. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

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14. Show all calculations, methods, or assumptions used to obtain an answer

. to mathematical problems whether indicated in the question or not.  :

l

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16. If parts of the examination are not clear as to intent, ask questions of I the examiner only. ,

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17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has  !

been completed.

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18. When you compl$ete your examination, you shall: l

a. Assemble your examination as follows:

(1) Exam questions on top.

(2) Exam aids - figures, tables, etc.

(3) Answer pages including figures which are part of the answer.

b. Turn in your copy of the examination and all pages used to answer the examination questions.

c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.

d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while.the examination is still in progress, your license may be denied or revoked. l l

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1. PRINCIPMc OF NUCM AR POWER PLANT OPERATION. Paco 4 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLQW QUESTION 1.01 (1.00)

The ratio of Pu-239 and Pu-240 atoms to U-235 atoms changes over core life. Which one of the pairs of parameters below are most affected by this change?

a. moderator temperature coefficient and doppler coefficient

b. Doppler coefficient and beta bar

c. beta bar and moderator temperature coefficient

d. moderator temperature coefficient and void coefficient QUESTION 1.02 (1.00)

Choose which of the following statements about Sa-149 is TRUE7

a. It is removed from an operating reactor by burnout and radioactive decay.

b. When a reactor, at MOL, is restarted after a temporary shutdown Sa-149 concentration increases for several days.

c. It has less effect on reactor operation than Xe-135 due to its smaller fission yield and smaller microscopic neutron cross section.

d. The equilbrium concentration of Sm-149 at 50% FP is about two thirds of the equilibrium concentration at 100% FP.

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(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

8

. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pcco 5 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.03 (1.00)

A middle of life reactor startup is in progress. The reactor engineer has estimated that the reactor should so critical at notch 26 on a particular control rod if the reactor operator reaches that notch at 0800.

HOW WOULD EACH OF THE FOLLOWING conditions or events AFFECT the ACTUAL CRITICAL ROD POSITION (more rod withdrawal, less rod withdrawal, or no significant effect)?

Assume that if no conditions changed, the reactor would have been critical at notch 26 of the indicated rod.

a. The reactor has been shutdown from extended full power operation for 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> and due to circumstances beyond your control you will not have the indicated rod at notch 26 until approximately 0900. (0.5)

b. Shutdown cooling is terminated. (0.5)

QUESTION 1.04 (1.50)

Indicate how fuel pin centerline temperature will change (INCREASE, DECREASE or REMAIN THE SAME) for each of the following conditions.

a. A 0.001 inch thick layer of corrosion product deposits on the clad surface.

b. The Pressure Set on EHC is lowered by 10 psig.

c. A fuel bundle reaches DNB l l l.

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. 1.. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pass 6 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.05 (1.00)

The fission process in a commercial reactor requires the neutrons that are " born" by fission to be "thermalized". The interaction in the reactor core which is most efficient on thermalizing neutrons for fission occurs with the ... (CBOOSE ONE)

a. OXYGEN atoms in the water molecules

b. BORON atoms in the control rods

c. ZIRCbHIUM atoms in the fuel cladding

d. HYDROGEN atoms in the water molecules QUESTION 1.06 (1.50)

For the following conditions, state if the magnitude of the FUEL TEMPERATURE COEFFICIENT becomes more or less negative.

a. Increase in fuel temperature

b. Increase in moderator temperature

c. Increase in void fraction QUESTION 1.07 (1.00)

[TRUE OR FALSE]

a. MAPRAT maintained within limits ensures that transition boiling will not occur in 99 percent of the fuel bundles

b. Maintaining MAPRAT limits ensures that the APLHGR limits are not violated.

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

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. 1m PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Paga 6 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW 1

QUESTION 1.05 (1.00)

The fission process in a commercial reactor requires the neutrons that are " born" by fission to be "thermalized". The interaction in the reactor core which is most efficient on thermalizing neutrons for ,

fisaion occurs with the ... (CHOOSE ONE) l I

a. OXYGEN atoms in the water molecules l l

b. BORON atoms in the control rods  ;

c. ZIRCONIUM atoms in the fuel cladding

d. HYDROGEN atoms in the water molecules l QUESTION 1.06 (1.50)

For the following conditions, state if the magnitude of the FUEL TEMPERATURE COEFFICIENT becomes more or less negative.

a. Increase in fuel temperature

b. Increase in moderator temperature

c. Increase in void fraction QUESTION 1.07 (1.00)

[TRUE OR FALSE]

a. MAPRAT maintained within limits ensures that transition boiling will not occur in 99 percent of the fuel bundles

b. Maintaining MAPRAT limits ensures that the APLHGR limits are not violated.

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

, 1. PRINCIP N OF NUCLEAR POWER PLANT OPERATION. Pose 7 i THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.08 (3.M)

Explain HOW and WHY Rod Worth changes for the following conditions.

a. Rod Worth of a center rod compared to a peripheral rod.

b. Rod Worth when plant. conditions change from cold to hot at 1 percent power.

c. Rod Worth when plant conditions change from hot at 1 percent power to hot at 100 percent power.

QUESTION 1.09 (3.00)

Describe HOW and WHY a centrifugal pumps discharge head is affected for each of the following. (Consider each condition seperatly and assume NPSE is maintained in all cases.)

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a. Suction pressure increases.

b. The discharge valve is throttled closed.

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c. The temperature of the fluid being pumped increases. -

QUESTION 1.10 (3.00)

a. Approximatly what percentage of neutrons from U-235 are born delayed? [0.5)

I b. How does the percentage of delayed neutrons produced in the CORE vary over core life and WHY. [1.5]

c. How do delayed neutrons contribute to the control capability of a commercial reactor? (1.00) l l (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

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. 1.. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Pc';a 8 l THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l

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QUESTION 1.11 (2.00) i The reactor is exactly critical LOW in the intermediate range.  ;

A control rod is withdrawn one notch. '

A. Describe WHAT happens to indicated neutron level AND WHY7 i (Continue your discussion until a steady state condition l is reached. Assume no futher operator action other than ranging the IRM meters. (No other parameters are changed.)

B. Describe BOW reactor period would respond AND WHY7 QUESTION 1.12 (1.50)

What are THREE of the design or operational factors that insure adequate Net Positive Suction Head (NPSH) for the '

recirculation pumps? l l

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(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) l l

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, 1.. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Paga 9 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW I

QUESTION 1.13 (1.50)

MATCH the appropriate THERMAL LIMIT (a-c), to each FAILURE MECHANISH AND to each LIMITING CONDITION given below. l THERMAL LIMIT l

a. Linear Heat Generation Rate (LHGR)

b. Average Planar Linear Heat Generation Rate (APLHGR)

c. Minimum Critical Power Ratio (MCPR)

FAILURE MECHANISM LIMITING CONDITION F1. Clad melting caused by L1. Coolant transition decay heat & stored heat boiling following a LOCA F2. Clad cracking from the surface L2. Clad plastic strain becoming vapor " blanketed" < 1%

F3. Clad cracking caused by L3. Maximum clad temp-high stress from pellet erature of 2200 des F expansion QUESTION 1.14 (1.50)

The Pellet-Cladding Interaction (PCI) failure mechanism is dependent upon many factors. List three (3) of these factors.

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. P o 10 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.15 (1.00),

How is condensate depression affected (INCREASED or DECREASED) by the following changes in the circ water flowing through the condenser?

a. Flow decreases.

b. Temperature increases.

(***** END OF CATEGORY 1 *****)

' FW NN Wo*4 *Nww wwNS dW"* pen- sa-r n w w'er---@ rC'* ---

et w + - ee- g---m e ---w--+--- w -

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Peca 11 SYSTEMS QUESTION 2.01 (3.00) l Answer the following in reguard to the Recire System:

a. In the event both recirc pump seals fall, WHAT componment would limit the Reactor Coolant System flow from the seal assembly?

b. How many consecutive starts are allowed if the Recirc Pump MOTOR is at Ambient temperature and if it is at Rated temperature?

c. What system supplies cooling to the recire pump motors and seals and what components are cooled. ,

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QUESTION 2.02 (2.00)

Give THREE of the four reasons bottom entry control rods are used on l BWRs. '

l QUESTION 2.03 (2.00) i For each of the following, state whether a ROD BLOCK, HALF SCRAM, or NO REACTOR PROTECTION SYSTEM ACTION is generated for that condition.

NOTE: If two or more actions are generated, i.e. a rod block and a half scram, state the most severe, i e. , half scram,

s. APRM 'B' downscale, Mode Switch in Run .

b. 12 LPRM inputs to APRM 'C', Mode Switch in Startup.

l c. Flow units 'A' and 'B' Upscale (less than 108 percent flow), Mode Switch in Run.

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d. Reactor water level +55 inches, Reactor Power 18 percent, Mode Switch in Run.

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

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. 2 .. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 12 SYSTEMS QUESTION 2.04 (2.00)

• During normal control rod insertion, the flow to the CRD is approximately 4 gym and flow during withdrawal is approximately 2 gym.

Since the overpiston area is approximately one-third that of the underpiston area, WHY isn't the withdrawal flow only one-third of the insert flow?

QUESTION 2.05 (3.00) l The ATWS system may be initiated either automatically.or manually. ,

a. What signals will automatically initiate the ATWS system? l (Setpoints required for full credit.)

b. What mitigating action occurs on an automatic initiation '

that does not occur on an manual initiation?

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QUESTION 2.06 ( 2'. 00) l m 1 What are FOUR of thes#4ve automatic initiation signals for the Standby Gas Treatment System? Setpoints required for full credit. l I

(1. 5)

QUESTION 2.07 (t:10)

For the Rod Block Monitor (RBM), provide answers to the following questions :

I

a. What is the system designed to prevent? (0.5) I

b. When the Meter Function Switch on the 90X-37 Back Panel Meter Section is in the " Count" position, what are the " units" of the indication on the meter and what can be calculated by utilizing the indicated value? ( t-+)

(10)

. (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 13 SYSTEMS QUESTION 2.08 (3.00)

Attached is a single line diagram of the RHR System (Figure 1). List the labeled components shown in the flow path for each of the following modes of the RER System:

a. Fuel Pool Cooling Assist

b. Suppression Chamber Spray

c. Drywell Spray QUESTION 2.09 (2.00)

Why is the Standby Liquid Control System (SLC) designed to require at least ninety (90) minutes to pump the SLC storage tank contents into the RPV7 QUESTION 2.10 (2.00)

What are FOUR of the FIVE possible causes for an automatic trip of the Reactor Water Cleanup Pump? (Setpoints not required for full credit)

(2.0)

QUESTION 2.11 (1.00) j Fill in the blank. l l

HPCI is designed to provide emergency core cooling during a l

. (Three words) (1.0) ,

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QUESTION 2.12 (2.00)

What are EIGHT components that would lose cooling if TBCCW were lost?

(***** END OF CATEGORY 2 *****)

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. 3, INSTRUMENTS AND CONTROLS Paca 14 QUESTION 3.01 (1.50)

Refer to Figure 2, Region of Detector Operating Characteristics Gas Conductivity Curve.

Identify which Region the SRM's, IRM's and LPRM's operate in.

. QUESTION 3.02 (1.00)

The IRM's are withdrawn from the core as soon as they are no longer needed.

Why is it important that this be done?

QUESTION 3.03 (1.50)

With the reactor operating at power, you are instructed by the SRO to place the mode switch in SHUTDOWN, which results in a SCRAM, and to reset the SCRAM as soon as possible.

a. When can this SCRAM be reset? [0.5]

b. Why does the circuitry not allow the SCRAM to be reset immediately?

[1.0]

QUESTION 3.04 (2.50) .

a. What THREE other systems are provided signals from the LPRM system?

[1.5]

b. What indication does an operator have of a high LPRM level? [0.5]

c. What indication does an operator have of a low LPRM level? [0.5]

QUESTION 3.05 (1.50)

There are six level switches associated with the scram discharge volume.

List the THREE functions they perform and their setpoints.

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

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t 3, INSTRUIENTS AND CONTROLS Peca 15 QUESTION 3.06 (3.50)

In regard to the Control Rod Drive Hydraulic System (CRDH):

a. What are the THREE functions performed by water discharged from the CRD pump? [1.5]  !

b. What is the purpose of the stabilizing valves? [1.0]

c. How do the stabilizing valves accomplish their purpose? [1.0]

QUESTION 3.07 (3.00)

Refer to attached Figure 3 Recirculation Speed Control Network, for the following:

a. The plant is operating at 30% power with both recire pump M/A transfer stations in MANUAL. FOR EACH of the following instances.

INDICATE HOW the speed of Reciro Pump 'A' would change (increase, decrease, or remain the same) AND WHICH component (s) of the control system is(are) limiting.

1. Recire Pump 'A' M/A transfer station is placed in ' AUTO'.

2. The generator speed tachometer output feedback signal fails low due to a loss of continuity through the field breaker contacts.

b. WHAT action must be taken by the control room operator prior to resetting a 'LOCEED OUT' scoop tube? WHY?

QUESTION 3.08 (3.50) >

Consider the Rod Block Monitor system (RBM):

a. WHAT are TWO reasons why the gain of the RBM channel is increased?

[2.0)

b. WHAT are THREE ways the RBM trips are BYPASSED? [1.5]

i I (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

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. 3, INSTRUMENTS AND CONTROLS Paco 16 QUESTION 3.09 (2.00)

a. What TWO parameters will initiate the loop select logic for the I LPCI mode of RHR? Include setpoints. l i

b. Explain how a recirc loop is chosen for the injection flowpath by the LPCI select logic. Assume loop Delta-P < 2 paid at initiation signal.

QUESTION 3.10 (3.00)

The reactor is operating at 40% power with the Feedwater Control System in single element control, and level Channel "A" selected for input. The  !

reference leg isolation valve to the Channel "A" NR GEMAC develops a i significant packing leak and the associated reference leg starts to  ;

gradually decrease.

Describe the effects on and of the following, assuming no operator action. l Include any setpoints.

a. Actual RPV water level

b. Indicated RPV water level on Channel "A"

c. FWLC

d. RPS

e. ECCS )

f. PCIS i

)

l QUESTION 3.11 (2.00) I The control room emergency sone is protected against some toxic gases by a control room HVAC system automatic isolation.  !

a. What gases are monitored by this isolation system. [1.0]

b. What indication is available to alert the operators that the isolation system has actuated? [0.5]

c. How is the control room emergency zone protected against toxic gases not monitored by the automatic isolation system? [0.5]

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(***** END OF CATEGORY 3 *****)

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. 4, PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pra 17 l AND RADIOLOGICAL CONTROL l l

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QUESTION 4.01 (2.00)  !

I What are FOUR of the five entry conditions for QGA 200-1, SUPPRESSION POOL TEMPERATURE CONTROL? l QUESTION 4.02 (2.00)

According to QOA 202-1 " Jet Pump Failure", there are two indications )

or symptoms where simultaneously occurring would indicate a failure of a jet pump.

WHAT are these two indications? l (Be specific, include setpoints if needed.)

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QUESTION 4.03 (1.00)

According to QGP 1-1 Normal Unit Startup, if a procedure is terminated at any time during its execution, several administrative requirements must be met.

List four of these requirements that must be performed.

4 QUESTION 4.04 (3.50)

If while operating at 95% power, with EGC in operation, an 'A' heater high level alarm and a step increase in APRM readings are received.  ;

i WHAT Immediate Action Steps would be taken to control reactivity l addition? (Limit your answer to the immediate action steps required ,

by QOA 400-1 " Reactivity Addition") (An action step may contain more  !

than one action item.)

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) j

. 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Paga 18 AND RADIOLOGICAL CONTROL QUESTION 4.05 (0.50) ,

[TRUE OR FALSE]

Once an Emergency Operating Procedure has been exited, it need not be re-entered even if an entry condition is reached.

QUESTION 4.06 (1.50)

According to Procedure QGP 1-2, Unit Startup to Hot Standby, a coupling check is performed each time a rod reaches position 48. What are THREE actions you must perform if uncoupling is detected during this check?

QUESTION 4.07 (2.00)

During the performance of QGP 1-2, Unit Startup to Hot Standby, when Reactor Pressure reaches approximately 875 psig rods are notched in to stop the pressure rise and pressure'is maintained between 850 psig and 920 psig.

What are the FOUR methods allowed to maintain RPV pressure within the specified band.

QUESTION 4.08 (2.00)

QGP 4-1, Control Rod Hovements and Control Rod Sequences, specifies the NSO's responsibility in regard to use of the Control Rod Sequence Package ,

prior to normal in-sequence control rod movements.

What are IOUR things the NSO must verify prior to commencing normal '

in-sequence control rod movements? )

l QUESTION 4.09 (2.50) j In order to assure control room operators are attentive to their panels at all times, six actions or performances are forbidden by QAP 300-2, Conduct of Shift Operations. What are FIVE of these actions or performances?

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(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

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. 4i PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Paga 19 AND RADIOLOGICAL CONTROL QUESTION 4.10 (2.50)

Procedure QOA 201-3, Inadvertent Actuation of One Main Steam Relief l Valve, requires six immediate operator actions in the event an SRV l inadvertently opens.

What are FIVE of these actions?

QUESTION 4.11 (2.00)

WHAT are TWO (2) exceptions to the requirement that " Individuals are prohibited from entering a high radiation area unless they notify the control room OR have a safety man present"?

QUESTION 4.12 (2.00)

Answer the following questions according to QOA 1700-5, Main Steam Line High Radiation, immediate operator action steps for High Radiation on both channels,

a. What are six of the automatic actions verified by step 1.b.

" Verify Automatic Actions ...". [1.5]

b. Step 1.c. has the operator place the Off-Gas isolation valve AO-5406 control switch to close. EXPLAIN the BASIS for taking this action? [0.5]

4 QUESTION 4.13 (1.50)

Concerning the use of radiation dose meters, answer the following i

TRUE or FALSE.

a. Earphones are only required in areas of low visability conditions.

b. A GM detector is preferred for setting dose rates.

c. Before entering a suspected radiation area, the meter selector

switch should be turned to the highest range.

(***** END OF CATEGORY 4 *****)

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

. EQUATION SHEET f = ma v = s/c v.g ,,ye+ g,g 2 Cycle efficiency = *I '

E = aC2 .

, , gy , y ,)jt II = hav vg=v + at A = kN A = A,e' E PE = agh u = e/t 1 = in 2/tg = 0.693/tg ,

W = v&P-

. tg (aff) = (t,,)(q) ..

AE = 931Am .

( 4 )

Il=[mcAT P I.I,*

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-Pur = W'g a" = I.I o to / M -x P=P 10 M (t). TVL = 1.3/u y.y t e /T o

EVL a 0.693/u

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SUR = 26.06/T T = 1.44 DT sCR = s/(1 - x,,g) f1'gro) set = 26 3_ CR,=s/i1-r,gg,)

T = p*/, ) + [(f ,)/1,,,,3 C"1Cl - *eff)1

• C"2CI ~~ Keff}'2 T,= t*/ (, _ r; M = 1/(1 - K.gg) = CR g/CR0 T = (I - p)/ 1,gg o g , Cg ~ geff)0/II ~ Eeff)1 8 " ( eff'I) eff " A eff eff E SDM = (1 - K,gg)/K,gg p= [1*/TKygg.} + [I/(1 + A,ggt )] 1* = 1 x 10" seconds

~I P = If7/(3 x 10 ) 1,ggA= 0.1 seconds E = No -

Idgg=Id22 UATER PARAMETERS Id g =I02 1 gal. = 8.345 lba R/hr = (0.5 CE)/d 2 g,,g,,,)

, 1 gal. = 3.78 liters R/hr = 6 CE/d2 gg,,,) ,

$ 1 ft3 = 7.48 gal. MISCELLANEOUS CONVERSIONS ,

Density = 62.4 lbs/fc 3 1 Curia = 3.7 x 10 dys 10 Density = L gn/cm 1 kg = 2.21 lba Heat of vag orization = 970 Etu/lba 1 hp = 2.54 4 10 BTU /hr 1

Heat of fusica = 144 Btu /lba 1 Hw = 3.41 x 106 Btu /hr l 1 Ata = 14.7 Psi = 29.9 in. I g. 1 Btu = 778 f t-lbf i 1 ft. H 2O = 0.4333 lbf/in 2 g inch = 2.54 cm F = 9/5'C + 32

• C = 3/9 ('T - 32)

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1. PRINCIPMM OF NUCLEAR POWER PLANT OPERATION. Pega 20 THERMODYNAMICS. MAT TRANSFER AND FLUID FLOW

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MASTER C0?Y ANSWER 1.01 (1.00)

b. doppler coefficient and beta REFERENCE G. E. Reactor Theory, pg. 70 WNP-II Version Quad Cities Rx Theory Coefficients of Reactivity pg 54 of 85 and Reactor Kinetics pg 40 ANSWER 1.02 (1.00)

c. It has less effect on reactor operation than Xe-135 due to its smaller fission yield and smaller microscopic neutron cross section.

REFERENCE G. E. Reactor Theory, pg. 87 WNP-II Version -

Quad Cities Rx Theory - Fission Product Poisons pg 72 ANSWER 1.03 (1.00)

a. Less rod withdrawal (due to Xenon burnout)

b. More withdrawal (due to heatup)

REFERENCE G. E. Reactor Theory, pas. 85, 66 l Quad Cities Rx Theory Fission Product Poisons. pg 70 figure 60  :

Coefficients of Reactivity pg 48 '

ANSWER 1.04 (1.50) i

a. increase

b. decrease i

c. increase l

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

1. PRINCIPfRR OF NUCfRAR POWER PLANT OPERATION. Pcco 21 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE G. E. BWR Training Center Thermodynamic Heat Transfer and Fluid Flow Quad Cities Thermodynamic Heat Transfer and Fluid Flow pg 7-55 thry 7-58 ANSWER 1.05 (1.00)

d. HYDROGEN atoms in the water molecules REFERENCE G. E. Reactor Theory and Heat Transfer Section II Neutron Physics Quad Cities Rx Theory, Fuel Nuclei and Neutron Properties for Fission pg 12 ANSWER 1.06 (1.50)

a. Less

b. More

c. More REFERENCE HTFF, pages 53, 54, and 58 Quad Cities Reactor Theory Coefficients of Reactivity pg 50, 52, 54 t

ANSWER 1.07 (1.00)

a. FALSE

b. TRUE l

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

.1 . PRINCIPr.um OF NUCf.u4R POWER PLANT OPERATION. Paca 22 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE ,

HTFF, pp 31 Quad Cities Heat Transfer and Fluid Flow pp 9-24a.

ANSWER 1.08 (3.00) ,

a. Control rods in the center of the core are exposed to a higher 4 thermal flux than those at the core periphery and, therefore, have

- a greater worth. (Rod worth would be increased anywhere in the core where a radial thermal neutron flux peak existed even.if it were an edge rod.)

b. As moderator temperature increases, its density decreases resulting in longer thermal diffusion lengths. This allows thermal neutrons to travel further and be absorbed by the rod; thus, rod worth increases with an increase in moderator temperature.

c. As voids increase, again less moderations takes place. Again, thermal neutrons travel further; however, voids tend to depress thermal flux because of the very poor moderation. Thus, as voids increase, rod wort decreases.

REFERENCE QC Reactor Theory, pages 58 and 60 i

ANSWER 1.09 (3.00)

n. Head inceases [0.5] the pump is still putting the same amount of work into the fluid, therefore the same delta pressure across the l

pump, so as the auction pressure increases so will the j discharge head [0.5]

b. Head increases [0.5] as system resistance to flow increases, pump l

8 head increases [0.5]

c. Head decreases [0.5] as temperature increases, system resistance to flow decreases (lower viscosity) , therefore head decreases [0.5].

l

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

1. PRINCIPMR OF NUCMAR POWER PLANT OPERATION. Pcs 23 I THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE Monticello, Thermodynamics and fluid flow ch 7 pg 111 Quad Cities, HT & FF Fluid Statics, Pump Characteristics, pg 6 6-102 ANSWER 1.10 (3.00)

a. 0.64% ( Will accept 0.6 to 0.7 %) [0.5]

b. Decreases [0.5] due to the production of PU-239 which has a lower delayed neutron fraction than U-235. [1.0)

c. Delayed neutrons increase the average neutron generation time increasing the control time of the reactor by a very large factor. Acceptable answers may vary from this wording. [1.0]

REFERENCE Quad Cities , Reactor Theory, page 10,38,40, and 85 ANSWER 1.11 (2.00) 4 A. Neutron level would start and continue to increase until the

point of adding heat is reached. As the coolant heats up, negative reactivity is added and power turns. Power would stablize at the point of adding heat.

B. Period would take a step jump due to the production of prompt neutrons. Immediately after this step, the rate of power change decreases to a rate controlled by delayed neutrons until the reactivity is no longer being increased. Then a sharp drop would occur as the rate of reactivity addition drops

. to zero. A stable period would continue until negative reactivity is inserted. Stabilizes at infinity.

1

)

i i

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

w m- c---- - - --,s. - - - - - - - - - - - - - - - ---we ---w - --t--- --s -.ne,- -,-. - - -,--,.w - - - - - - - - - . - - , , - , -- - - - - - - - , - - - - - - - - - - - - - - -

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. Para 24 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE Dresden NUS Theory Quad Cities Theory Review pages 44-45 and 78.

ANSWER 1.12 (1.50)

1. They are physically located as far below the normal water line as possible to provide the greatest static head.

2. With feed flow less than 20% they are kept on minimum speed.

3. At high power operation adequate NPSH is obtained from feedwater subcooling.

4. Low reactor Vessel water level trip, cavitation interlock.

5. Suction valve closed trip, cavitation interlock.

(Any 3 0 0.5 ea)

REFERENCE 4

DRESDEN - Recire System Lesson Plan pg 16 & 18 GE Thermodynamics, Heat transfer & Fluid Flow, page 7-93 & 94 Quad Cities LIC 0202-1 pg 19 and 21 ANSWER 1.13 (1.50)

a. F3. L2

b. F1. L3

c. F2, L1  ;

REFERENCE EIH: GPNT, Vol VII Chapter 10.2-23 Quad Cities Heat transfer and Fluid Flow pg 9-16a, 9-19a, and 9-34a

(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)

1. PRINCIPTVR OF NUCTRAR POWER PLANT OPERATION. Paga 25 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWER 1.14 (1.50)

Fuel rod power level Fuel rod exposure Rate of power increase Fuel pellet design Previous power history Presence of embrittling agent T3-~r~eisirid at 075 each)- A" j '" , j / ,,, ';,, ., w REFERENCE Fuel System Description, pg 15, 16 GE HTFF ps 9-107, 108 Quad Cities Beat Transfer and Fluid Flow pg 9-45a andL9-46a ' W 1- 17 Gs.) C. i . . v. . . //r., jfa - 0 2:1~ 2 , pa;). . , , ,a j ,

g;,

I ANSWER 1.15 (1.00)

a. Decreases

b. Decreases REFERENCE MNS Thermodynamics, pp. 5-18 , 5-20, & 5-21 Quad Cities, HT & FF , Beat Transfer , pp 7-45

, t 1

l

(***** END OF CATEGORY 1 *****)

. c

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 26 SYSTEMS ANSWER 2.01 (3.00)

a. Breakdown bushing [1.0]

b. Two at ambient, one at rated., [0.5 each]

c. RBCCW Oil Coolers (Will also accept bearings instead of oil coolers)

Seal Coolers

[0.33 each] s.

REFERENCE Quad Cities, Lesson Plan, LIC-0202-1, PP 6 and 8 ANSWER 2.02 (2.00) I

1. Less time is required during refueling outages to remove and .

reinstall the reactor head since control rod drives are not a factor.

2. Internal moisture removal and steam separation can be more easily accomplished if there is no interference from top mounted control reds.

3. A large percentage of voids exist in the upper part of the core which significantly reduces ttie power in this area. If the control blade were to be used entering from the top of the core it would over-depress the flux in the upper part of the core.

4. Control rods are used for axial power shaping by leaving some of them partially inserted in the lower portion of the core. This helps to control flux peaking in local areas of the core and yields the optimum fuel burnup.

(Any 3 0 0.66 ea)

REFERENCE Quad Cities, Lesson Plan, LIC-0300-1, pp 2

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

~~'

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paga 27 SYSTEMS ANSWER 2.03 (2.00)

a. Rod Block

b. Half Scram )

c. "A B1m'i '- l-4 ( t'" b'S' '"~ (p m d l*;*'

d. No Reactor Protection System Action 1

I REFERENCE Quad Cities RPS and APRM Lesson Plans ANSWER 2.04 (2.00)

Higher flow is required on withdrawal to accommodate collet seal leakage. ,

l REFERENCE I Quad Cities, Lesson Plan, LIC-0300-2, PP. 10  !

ANSWER 2.05 (3.00)  !

a. High RPV pressure - 1250 psig ( + or - 5 pais)

Low-Low RPV W/L - -59" (+ or - 1") [0.5 for par. & 0.5 for value]

(2.0)

b. Recire pump HG field breakers trip. [1.0)

REFERENCE Quad Cities, Lesson Plan, LIC-0300-3, PP.4 l

l

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

l l

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Paca 28 SYSTEMS ANSWER 2.06 (2.00) 3 n -f,

1. High radiation in the Rx building vent system 3 1? u A.r.

G

2. High radiation on refuel floor,100 nr/hr.

n

~

3. High Drywell Pressure 3 2 PSIG.

4. Low Rx water levelg + 8" .

3. pry 4, .il n.L.eu- s sc:: her.

/,,5 . After,v4jEsec. time delay if primary system does not operate, the back up syst'ela initiates.

[ Any 4 0 0.5 each]

REFERENCE

- M Citi.. , L...vu T1 , Oteca , Gas iremi,-ou. 0.etivu '

Gs J C:6*s,In a r' , c : .iu ,-t ,g .,j .

a,, % f ;.~ c , . y , j 4 c,

(/. 50 )

ANSWER 2.07 (f.-edt)

a. Local fuel damage (by generating a rod withdrawal block). (0.5)

b. Units = volts [0.5], number of operable LPRM inputs can be l calculated (by using 1 volt per operable input) [-h91

o. 5 REFERENCE QC Lesson Plan Rod Block Monitor - LIC-9700-5, pp. 2 of 22.

QC Lesson Plan Rod Block Monitor - LIC-9700-5, pp. 18 of 22, j Sec. 2.b.(4).

4 QC LessonPlan Rod Block Monitors - LIC-9700-5, pp. 21 of 22, Sec III.C.b.

i

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

. . . _ . . _ . - _ . _ _ _ , - . . . , . . - - . , . _ _ , . ,.,_...,._,,_.,,._.._m_ _ _ _ _ - , _ . _ _ _ . , , - _ _ , - _ _ _ . . . _ _ _ . . . . . - . . , - _ , , ,

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Pcsa 29 SYSTEMS ANSWER 2.08 (3.00)

a. Spool piece, 43 valve, pump HX, 19 valve, spool piece. '(1.0)

b. 7 valve, pump, HX or 16 valve, 34 valve, and 37 valve.

(1.0)

c. 7 valve, pump, HK or 16 valve, 23 valve and 26 valve. (1.0)

(HX OR 16 valve acceptable)

REFERENCE QC Lesson Plan, RER LIC-1000-1 Fig. 8, Fig. 5, and Fig. 4.

ANSWER 2.09 (2.00)

An excessive injection rate could result in poor mixing [0.5] of the coolant and boron [0.5] and result in reactivity or power chugging

[0.5) which in turn could cause fuel damage [0.53 REFERENCE QC Lesson Plan - Standby Liquid Control LIC-1100, pp. 11 of 14, Sec.

IV.A.B.C.

ANSWER 2.10 (2.00)

a. Pump Low Flow (30 spa)

b. Eigh bearing cooling water temperature.(140 deg. F)

c. The TWO valve not full open (OB Inlet Isolation Valve)

d. The FIVE valve not full open (IB Inlet Isolation Valve) l e. EIGHTY valve full closed (Outlet to FW Bdr.)

I

[any FOUR S 9.5 each] (1.0)

REFERENCE QC Lesson Plan - Reactor Water Cleanup System - LIC-1200, pp. le of 13 Sec. B.4.

i l

(***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)

l

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 33 SYSTEMS ANSWER 2.11 (1. M)

Small Line Break (1.0)

REFERENCE QC Lesson Plan LIC-23N-1, pp. 2 ANSWER 2.12 (2. N)

1. Ci:rculating Pump Upper Bearings

2. Service Air Compressors and Aftercoolers

3. Sperge Air Compressors

4. Control Rod Drive Pumps

5. Feed Pump Oil and Seal Coolers

6. Condensate and Condensate Booster Pumps

7. EHC Coolers

8. Bus Duct Coolers

9. Alterex Coolers

10. Turbine Building Process Sampling system Sink

11. Condensate Domin. Air Surge Compressor

[any EIGHT e 0.25 each]

REFERENCE QC Lesson Plan TBCCW LIC-38 N , pp. 4 of 7, Sec. II.E.

I l

(***** END OF CATEGORY 2 *****)

~

3. INSTRUMENTS AND CONTROLS Pcco 31 l

ANSWER 3.01 (1.50)

SRM's - Region C IRM's - Region B LPRM's - Region B REFERENCE Quad Cities, Lesson Plan, LIC-9700-1, SRM, Pg. 4 ANSWER 3.02 (1.00)

To prevent reaching the design lifetime of the detectors prematurly.

REFERENCE Quad Cities, Lesson Plan, LIC-9700-2 IBM's, Pg. 5 ANSWER 3.03 (1.50)

a. After a le sec. time delay. [0.5]

b. Allows time to ensure all rods have completed their SCRAM stroke.

[1.0]

REFERENCE QC, Leson Plan,LIC-0500-1,RPS, pp 44.

l ANSWER 3.04 (2.50)

a. - Rod Block Monitor [0.5] j

- Process Computer [0.5] i

- APRMs [0.5] i

b. High - Red light on full core display [0.5]

c. Low - Blue light on full core display [0.5]

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

-- - - , - .._e_. - , . . , ..,.,,y....,.-m-__m.,,

3. INSTRUMENTS AND CONTROLS Paca 32 REFERENCE -

Vermont Yankee NPC Systems Manual, Nuclear Instrumentation Quad Cities, Lesson Plan, LIC-700-3, PP. 11, 3 ANSWER 3.05 (1.50)

1. SCRAM discharge volume high level ALARM [0.25] 10 gal. [0.25].

2. Control rod block [0.25] 25 gal. [0.25]

3. Reactor SCRAM [0.25].40 gal. [0.25]

REFERENCE Vermont Yankee NPC Systems Manual, Control Rod Drive Hydraulic System Quad Cities, Lesson Plan, LIC-0300-2, Pg. 28 ANSWER 3.06 (3.50)

a. - Drive water

- Charging water

- Cooling water [3 0 0.5 each]

b. The purpose is to maintain a relatively constant flow through the pressure control valve so stable drive water pressure is maintained during rod movement. [1.0)

c. The stabilizing valves close on a rod movement signal to offset the additional flow through the drive water header. [1.0]

REFERENCE Vermont Yankee Systems Manual, Control Rod Drivo System Quad Cities, Lesson Plan, LIC-0300-2, Figure 5, Ps. 26 l

l l

(***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)

3.' INSTRUMENTS AND CONTROLS Paca 33 ANSWER 3.07 (3.00)

a. 1. Increase [0,5]. Masterlimiter,(lowspeedlimit)[0.5]. .

2. Increase [0.5]. Scoop tube positioning unit or electrical 2*e

( '. "t*;"" 5'J L a u s N

b. The null voltmeter is used to match the speed from the tachometer with the speed demand from the speed controller [0.5], to prevent a flow transient during reset [0.5].

REFERENCE Recire Flow Control, LIC-0202-2, PP. 10, 12, 14, 36 ANSWER 3.08 (3.50)

a. 1. The local power may be significantly lower than the core avg. ,

[0.5] which could result in withdrawing a rod of abnormally high worth [0.5]. The gain is increased to restrict the rate of power rise. [1.0] l

2. Several of the highest reading LPRMs might be bypassed [0.5].

Gain increased to compensate for lower core average [0.5]. [1.0)  ;

b. 1. Joystick (on 90x-5)

2. Edge rod selected

3. Reference APRM < 30% [3 required 0 0.5 each]

REFERENCE Rod Block Monitor, LIC-0700-5, Ps. 6, 14, 8, 10

ANSWER 3.09 (2.00)

a. Drywell press >/= 2 psig [0.53 and RPV water level </= -59"

< [0.5].

b. Injection loop is selected by the comparison of recire system i riser DP's [0,5]. If loop A > B, A is selected and if loop B is '.

>/= loop A, loop B is selected [0.5].

I i

• • • • • )

(***** CATEGORY 3 CONTINUED ON NEXT PAGE

- - - - . , ...-..._,.-,..n,--,.-- . , - - - - , _ . . , , - , - ---.,,,,n__,.,n..n,_.,_.w...,..n--._--,,.n,nw,,-._,-.,----.----- , . . - . = - .

. 3.' INSTRUMENTS AND CONTROLS P=a 34 REFERENCE Quad Cities, Lesson Plan - RER LIC-1000-1 Sec. 4, Pg. 44 of 67 ANSWER 3.10 (3.00)

a. Actual vessel level is decreasing,

b. Level Channel "A" will indicate increasing water level.

c. FWLCS will close the FRVs to try to maintain level.

d. Reactor will scram 9 +8" due to low reac

p. -FC-", Cre;; 1 isel.;,leef_ HPCI initiation,','ttor -59". water level. , n

f. PCIS Group 2 and 3 isolations -aleet 9 +8" x1 t'; t 6 - y L n. .. " - -

REFERENCE If:~ > y < - ' , Eb & l l Z. s s'* . : g 0 5 ' ' ' " * " * ' ' E M**' IY A Quad Cities, Reactor Level and Pressure, LIC-0263, PP. 12

s. t  :.c.. /. . .- p4. , zz -isos,.'.. 7,.,, y _n . g c ,. . . , . . . i. .

t'/ .- 1: at: -$ ,,,,,,.y ANSWER 3.11 (2.00)

a. Ammonia Clorine Sulfur dioxide [3 0 0.33 each]

b. Control Room HVAC " Major Trouble" Alarm. [0.5]

c. Manual isolation by operator. [0.5]

REFERENCE Quad Cities, Lesson Plan, LIC 5750. Ventillation, PP. 27

(***** END OF CATEGORY 3 *****)

• 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pecs 35 AND RADIOLOGICAL CONTROL ANSWER 4.01 (2.00)

1. Suppression pool temperature above 95 F.

2. Drywell temperature above 180 F.

3. Drywell pressure above 2.0 psig.

4. Suppression pool water level above +2.0 in.

5. Suppression pool water level above -2.0 in. [any 4 0 0.5 each]

REFERENCE Quad Cities, Procedure QGA 200-1, Suppression Pool Temperature Control, PP. 1 ANSWER 4.02 (2.00)

1. The recirculation pump flows (0.34] differ by more than le percent

[0.33] from established speed-flow characteristics [0.33].

2. The indicated total core flow [0.34] is more than le percent greater than [0.33] the core flow value derived from established power-core flow relationships [0.33].

REFERENCE Quad Cities QOA-202-1 Jet Pump Failure Rev 2 pg 1 ANSWER 4.03 (1.00)

1. Insert statement at point of termination as to reason for termination.

! 2. Insert title of subsequent procedures at point of termination

! 3. Insert NA in all remaining blanks

4. Date, time, and sign the last page

5. Retain the terminated procedure in the startup package. ,

(Any 4 0 0.25 ea) I i

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

- . . _ - . , , - . - - . _ , . . , , - - - - - , . . - ~ _ - , - . - - - - - . - - - - - - - , - - - - - - - , , . - - . ------------w------ ---- ------*w ~

~

- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Peca 36 AND RADIOLOGICAL CONTROL 1

REFERENCE Quad City QGP 1-1 Sec C.8 pg 3 ANSWER 4.04 (3.50)

1. Trip EGC [0.25] return reciro flow control to Manual [0.25]

2. Reduce recirculation flow [0.25] by at least 204 speed [0.25),

3. Insert control rods [0.25] starting at the present location in sequence and work backwards [0.25]. Any rod inserted should be

- continuously inserted all the way to position 90 [0.25] /6ven if the sequence calls for stopping at an intermediate position./CO3C After each group of rods check the flow control line [0.25] and continue this process until the reactor is under the 100 % FCL

[0.25]

If a LSSS has not been exceedef,gverify that reactor power is g

4.

within operating limits of exsisting recirculation system flow

[0.25]. If reactor power continues to increas Scram the reactor ,

I.O . 25 ] . , __ _ ,_ ___ ,

.. . ~ . . , . . . . .. . . . . . . . . . < . . ~ -

$. .vvamp_ amu.5 b 1OTA EEU LWUT.LUKA bG K5e r u

_ h.aw 7_ _ _

TO S'*e ^ y mel mQ(

- . -- r ;

.v. .....v ,vuwwo -%u 14v7 6 . ., a .

REFERENCE Quad Cities QOA 400-1 Reactivity Addition Rev 3 pg 1 & 2 ANSWER 4.05 (0.50)

False REFERENCE Quad City QGA Caution #1 and Statement immediatly after each sec of entry conditions pg 1 plus l

l l

l

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) l l

• 4.' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 37 AND RADIOLOGICAL CONTROL ANSWER 4.06 (1.50) I M/' lo" '

- ' /= /'~* / T-

< g. n, lr. - .

1. Stop rod withdrawals. '

2. Notify the shift engineer. "

. Eta - t. . . ut, da<,e l L I* ' ' - -

3. Refer to the abnormal procedures x .- .

REFERENCE g d g :).5 e. .

Quad Cities, Procedure QGP 1-2, Unit Startup to Hot Standby, PP. 16 ANSWER 4.07 (2.00)

1. Operate RCIC -

2. RWCU system reject flow

3. Vary power with CR's

4. SRV manual operation to prevent reaching 1960 psig if the other three methods fail to maintain pressure.

REFERENCE Quad Cities, Procedure, QGP 1-2, Unit Startup to Hot Standby, PP. 6

,.r**'

,tv i ANSWER 4.08 (2.00) 7,. . .

e

1. (QTP 1600-S2) Control Rod Sequence Review Sheet 4 has been reviewed and approved.

2. The RWH has been loaded and verified (by checking for the sign-off on QTP 1600-S2, Control Rod Sequence Review Sheet).

3. (QTP 1600-S3) Control Rod Sequence Sheet has been signed and dated by both a Qualified Nuclear Engineer (QNE) and an on-shift SRO before moving control rods in the group.  ;

4. Either the Rod Worth Minimizer (RWM) is in NORMAL or, if in BYPASS, l that a caution card has been placed on the rod movement control l switch (per procedure QOP 207-2). 1 l 5. Permission is obtained from the Shift Engineer to commence rod 1 I notion, j

[any 4 0 0.5 each]

l l l

l l (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

l

• 4.' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pcca 38 AND RADIOLOGICAL CONTRCL REFERENCE Quad Cities, QGP 4-1, Control Rod Hovements and Control Rod Sequences, PP. 4 ANSWER 4.09 (2.50)

1. Sleeping.

2. Habitual or chronic lack of attentiveness.

3. . Card playing, games, or other distractions from prescribed duties.

4. ' Alcohol or drug use.

{ 5. Practical jokes which could reduce the ability of persons or equipment to perform as required.

6. Other acts which could adversely affect the ability of individuals or equipment to perform their intended safety functions.

[Any 5 0 0.5 each]

REFERENCE (wull 'l" *'"r* ''.'s',<' ".;i L ~- ,

^ b

'"'~' ~ " "

y . 6 . , , ,-r . * * ,. <le 3 *' '

~

Quad Cities, Procedure QAP 300-2, Conduct of Shift Operations, PP. 6 ANSWER 4.10 (2.50)

1. Attempt to close the affected relief valve by placing its key-lock switch to OFF.

2. Verify that all reactor parameters are within operating limits.

3. If the valve will not close manually, determine the cause of the malfunction and attempt to close the relief valve.

4. If the valve cannot be closed, SCRAM the reactor,(and refer to QOA 201-2.)

5. IF an entry condition for a QGA procedure occurs, THEN enter that procedure.

6. If pool temperature reaches 160 F, perform an external torus inspection. (TS)

[any 5 0 0.5 each]

1 1

l

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

4.' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Poca 39 AND RADIOLOGICAL CONTROL l

REFERENCE Quad Cities. Procedure QOA 201-3, Inadvertent Actuation of One Main Steam Relief Valve, PP. 1 ANSWER 4.11 (2.00)

1. Operators and other personnel on R-rounds

2. Jobs when radio or sound powered communications with the control center is possible.

3. Personnel entering the sample hood fenced area where the area is clearly visible to personnel outside the R area.

[2 required 0 1.0 each]

REFERENCE Quad Cities, QAP 1120-6, Rev. 8, Entering a locked high radiation area without a timekeeper; Pg. 2 1

ANSWER 4.12 (2.00)

a. 1. Reactor Scram

2. MSIV's close

3. Off-Gas isolation valves close

4. Air ejector auction valves close

, 5. Mechanical vacuum pump trips

6. Main Steam Line drains close -

7. Primary sample valves close 6 0 0.25 ea:

b. This will prevent re-opening of the Off-Gas iso..ation valves [0.25]

j when the high radiation condition is reset [0.25].

i l REFERENCE 5

l Quad Cities QOA-1700-5 MSIV High Rad Rev 4 pg 1 s

(***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)

'4.' THOCEDURES - NORMAL. ABNORMAL. EMERGENQY Pcga 49 AND RADIOLOGICAL CONTROL ANSWER 4.13 (1.50)

a. False

b. Falso

c. True (3 0 0.5 ea]

REFERENCE Quad Cities Rad Protection LIL B-3 l

(***** END OF CATEGORY 4 *****)

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

- - - - - . , . , - - . -- ,--- ,----- - ,,,e. , , , . - . . - - - ,,,-n,n... , , - - - - - - - -

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TEST CROSS REFERENCE Pese 1 QUESTIOM VALUE REFERENCE 1.01 1.00 ZZZ0000001 1.02 1.00 ZZZ0000002 1.03 1.00 ZZZ0000003 1.04 1.50 ZZZ0000004 1.05 1.00 ZZZ0000005 1.06 1.50 ZZZ8000006 1.97 1.00 ZZZ8000007 1.08 3.00 ZZZ0000008 1.99 3.00 7.ZZ0000009 1.10 3.00 ZZZ8000010 1.11 2.00 ZZZ0000011 1.12 1.50 ZZZO900012 1.13 1.50 ZZZ8000013 1.14 1.50 ZZZ0000014 1.15 1.00 ZZZ0000015 24.50 2.01 3.00 ZZZO900016 2.02 2.00 ZZZO900017 2.03 2.00 ZZZ0000018 2.04 2.00 ZZZ0000019 2.05 3.00 2220000020 2.06 2.00 ZZZ8000021 2.07 2.00 ZZZ8000022 2.08 3.00 ZZZO900023 2.09 2.00 ZZZO900024 2.10 2.00 ZZZ8000025 2.11 1.00 ZZZ0000026 2.12 2.00 ZZZ0000027 26.00 3.01 1.50 ZZZ0000028 3.02 1.00 ZZZ8000029 3.03 1.50 ZZZ0000030 3.04 2.50 ZZZ8000031 3.05 1.50 ZZZ2000032

3.06 3.50 ZZZO900033 1

3.07 3.00 ZZZO900034

, ". . so 3.50 ZZZ8000035 3.09 2.00 ZZZ0000036 3.is 3.90 ZZZ8000037 3.11 2.00 ZZZ8000038 25.00 4.01 2.00 ZZZ0000039 4.02 2.00 ZZZ8000040 4.03 1.00 ZZZ8000041 4.04 3.50 ZZZO900042 4.05 0.50 22Z0000043 4.06 1.50 2Z20000044 4.07 2.00 ZZZ8000045 4.08 2.00 ZZZ0000046 l

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O TEST CROSS REFERENCE Po 2 QUESTION .. VALUE REFERENCE 4.09 2.50 ZZZ8000047 4.10 2.50 ZZZ8000048 4.11 2.00 ZZZ0000049 4.12 2.00 ZZZ0000050 4.13 1.50 ZZZ8000051 25.00 100.5 4

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U. S. NUCLEAR REGULATORY COMMISSION

, d M / M REACTOR OPERATOR LICENSE EXAMINATION FACILITY: QUAD CITIES 1&2 REACTOR TYPE: BWR-GE3 DATE ADMINSTERED: 86/10/14 EXAMINER: MORGAN. T.

CANDIDATE INSTRUCTIONS TO CANDIDATE: ,

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

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

E OF CATEGORY % OF CANDIDATI'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY I

25.59 51 5. THEORY OF NUCLEAR POWER PLANT 2 5. c.5 OPERATION, FLUIDS,AND g ,. ; THERMODYNAMICS ge 2 5.00 24.Se S4-20 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION z 3.0O 4

25.25 2t-Ft 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 27*O 14. 13 4 hee- . 2fr.-6 B 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS

/D1 101.2 Totals All work done on this examination is my own. I have neither given '

nor received aid. i I

Candidate's Signature 1

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

, During the administration of this examination the following rules apply:

1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.  !

2. Restroom trips are to be limited and only one candidate at a time may I leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3. Use black ink or dark pencil only to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the l examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.

8. Consecutively number each answer sheet, write "End of Category _" as appropriato, start each category on a new page, write only on one side of the paper, and write."Last Page" on the last answer sheet.

9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lines between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.

12. Use abbreviations only if they are commonly used in facility literature.

13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.

14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16. If parts of the examination are not clear as to intent, ask questions of the examiner only.

17. You must sign the statement on the cover sheet that indicates that the ~

work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

i (2) Exam aids - figures, tables, etc.

(3) Answer pages including figures which are part of the answer.

b. Turn in your copy of the examination and all pages used to answer the examination questions.

i- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.

d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

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5. 'NBIORY OF NUCLEAR POWER PLANT OPERATION. Page 4 FLUIDS.AND THERMODYNAMICS QUESTION 5.01 (1.00)

The ratio of Pu-239 and Pu-240 atoms to U-235 atoms changes over core life. Which one of the pairs of parameters below are most affocted by this change? t

a. moderator temperature coefficient and doppler coefficient

b. Doppler coefficient and beta bar

c. beta bar and moderator temperature coefficient

d. moderator temperature coefficient and void coefficient QUESTION 5.02 (1.00)

Choose which of the following statements about Sm-149 is/are TRUE7

a. It is removed from an operating reactor by burnout and radioactive decay.

b. When a reactor, at MOL, is restarted after a temporary shutdown Sm-149 concentration increases for several days, i c. It has less effect on reactor operation than Xe-135 due to its smaller fission yield and smaller microscopic neutron cross I section, l

d. The equilbrium concentration of Sm-149 at 50% FP is about two thirds of the equilibrium concentration at 100% FP.

QUESTION 5.03 (1.50)

Indicate how fuel pin centerline temperature will change (INCREASE, DECREASE or REMAIN THE SAME) for each of the following conditions.

! a. A 0.001 inch thick layer of corrosion product deposits on the clad surface.

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b. The Pressure Set on EHC is lowered by le pais.

c. A fuel bundle reaches DNB.

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

5.- hRYOFNUCLEARPOWERPLANTOPERATION. Page 5 FLUIDS.AND YBERMODYNAMICS e

QUESTION 5.04 (1.00)

The fission process in a commercial reactor requires the neutrons

" born" by fission to be "thermalised". The interaction in the reactor core which is most efficient on thermalizing neutrons for fission occurs with the ... (CHOOSE ONE)

a. OXYGEN atoms in the water molecules

b. BORON atoms in the control rods

c. ZIRCONIUM atoms in the fuel cladding

d. HYDROGEN atoms in the water molecules QUESTION 5.05 (1.50)

For the following conditions, state if the DOPPLER COEFFICIENT becomes more or less negative.

a. Increase in fuel temperature

b. Increase in moderator temperature

c. Increase in void fraction QUESTION 5.06 (3.00)

Describe HOW and WHY a centrifugal pumps discharge head is affected for each of the following. (Consider each condition seperatly and assume NPSH is maintained in all cases.)

a. Suction pressure increases.

b. The discharge valve is throttled closed.

c. The temperature of the fluid being pumped increases.

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5. THEORY OF NUCLEAR POWER PLANT OPERATION. Pasa 6 l FLDIDS.AND THERMODYNAMICS l

QUESTION 5.07 (3.00)

Explain HOW and WHY Rod Worth changez for the following conditions.

a. Rod Worth of a center rod compared to a peripheral rod at 50 percent power.

b. Rod Worth when plant conditions change from cold to hot at 1 percent power.

c. Rod Worth when plant conditions change from hot at 1 percent power to hot at les percent power.

QUESTION 5.08 (2.00)

Prior to startup (all rods full in), the SRM count rate is le CPS and

'K' effective is 0.96 I

a. Assume the control rods are pulled to give a delta 'K' of positive

8.035.

WHAT count rate on the SRMs should be expected when the period becomes infinite?

b. Now assume additional control rods are pulled to give a delta

'K' of 0.003.

WOULD the time requ.eed to reach an infinite period be greater or less than the time in part "a"?

Give the reason for your answer.

QUESTION 5.99 (2.00)

Your reactor operator informs you MAPRAT is 1.02.

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! a. Is the MAPRAT, as stated, conservative? Explain your answer.

(1.0)

I b. In regards to MAPRAT which of the following statements are TRUE l and which are FALSE?

1. MAPRAT maintained within limits ensures transition boiling will not occur in 99 percent of the fuel bundles. (0.5)

2. Maintaining MAPRAT limits ensures the APLEGR limits are met l (0.5).

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5. THEORY OF NUCLEAR POWER PLANT OPERATION. Pese 7 FLUIDS.AND THERMODYNAMICS 0

QUESTION 5.10 (2.M)

The reactor is exactly critical LOW in the intermediate range. A control rod is withdrawn one notch.

a. Describe what happens to indicated neutron level AND why?

(Continue your discussion until a steady state condition is reached.

1 Assume no futher operator action other than ranging the IRN meters.

(No other parameters are changed.)

b. Describe how reactor period would respond AND why?

(Continue your discussion until a steady state condition is reached.

Assume no futher operator action other than ranging the IRH meters.

(No other parameters are changed.)

1 QUESTION 5.11 (1.50)

What are THREE of the design or operational factors that insure adequate Net Positive Suction Head (NPSB) for the recirculation pumps?

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5. 'k'HEORY OF NUCLEAR POWER PLANT OPERATION. Page 8 FLDIDS.AND THERMODYNAMICS .

1 QUESTION 5.12 (1.50)

MATCH the appropriate THERMAL LIMIT (a-c), to each FAILURE MECHANISM AND to each LIMITING CONDITION given below:

THERMAL LIMIT l

a. Linear Heat Generation Rate (LEGR)

b. Average Planar Linear Heat Generation Rate (APLHGR)

c. Minimum Critical Power Ratio (MCPR)

FAILURE MECHANISM LIMITING CONDITION i

F1. Clad melting caused by L1. Coolant transition decay heat & stored heat boiling following a LOCA ,

F2. Clad cracking from the surface L2. Clad plastic strain becoming vapor " blanketed" < 1%

F3. Clad cracking caused by L3. Maximum clad teap-high stress from pellet erature of 2200 des F expansion QUESTION 5.13 (1.50)

The Pellet-Cladding Interaction (PCI) failure mechanism is dependent upon many factors. List three (3) of these different factors.

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(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

'. 5 hERORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS.AND THERMODYNAMICS Pasa 9 t QUESTION 5.14 (3.99)

A reactor startup is in progress. You have been given the estimated critical rod position for the conditions at 9899.

You start to pull control rods at 9800 for the approach to critical. HOW WOULD IACH OF THE FOLLOWING conditions or events AFFECT the ACTUAL CRITICAL ROD POSITION (more rod withdrawal, less rod withdrawal, or no significant effect)?

a. One reactor recirculation pump is stopped (Hypothetical situation only)

b. Xenon is changing due to extended power operation, terminated 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> previouly.

c. Shutdown cooling is stopped (significant decay heat)

d. Reactor head vent is inadvertently closed.

e. Moderator temperature is gradually decreasing.

f. Reactor Water Cleanup System isolates (significant decay heat),

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6. hLANTSYSTEMSDESIGN. CONTROL.ANDINSTRUMENTATION Page le QUESTION 6.01 (2.00)

For each of the following, stats whether a ROD BLOCK, HALF-SCRAM, or NO REACTOR PROTECTION SYSTEM ACTION is generated for that condition.

NOTE: If two or more actions are generated, i.e., rod block and half-scram, state the most severe, i.e., half-scram,

a. APRM 'B' downscale, Mode Switch in Run

b. 12 LPRM inputs to APRM 'C', Mode Switch in Sturtup

c. Flow units 'A' and 'B' upscale (less than 108 percent flow) Mode Switch in Run

d. Reactor water level 55 inches, Reactor power 18 percent, Mode Switch in Run.

25 QUESTION 6.02 ( 3.49-)

Answer the following questions concerning the feed water level control system.

, n. There are two feed water regulating valves provided to control the 1

feed flow. For each of the valves provide the following information.

1. How each is operated (e.g. electrically, pneumatically, etc.)

[0.5]

i i 2. The type of valve for each (e.g. sate, globe, ball etc.) (0.5]

i! 3. What conditions will cause a " Lock up" on each valve. (l.d$)

b. How does the Steam Leak Detection Device perform its function? [0.75]

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8. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PcO3 11 I QUESTION 6.03 (1.00) l For each of the areas listed choose which type of fire protection system i protects it.

AREA FIRE PROTECTION SYS

a. Computer Room 1. Halon 1211

2. Halon 1301 l b. Recire MG Set Room 3. Wet Pipe

4. Deluge (Dry pipe)

c. Diesel Generator Roon 5. Preaction

6. Foam Suppression

d. Main Transformer 7. Cardox

.Soh QUESTION 6.04 ((*)3r00)

With regard to an Anticipated Transient Without a Scram (ATWS) protective system, it is stated "All of the anticipated transients, which require mitigation in the unlikely event of an ATWS, quickly reach at least one of the two conditions which are readily sensed and from which mitigated actions may be initiated."

a. WHAT are four of the seven anticipated transients which require mitigation? / .

b. WHAT are the two conditions that will automatically initiate mitigating actions and WHAT are their setpoints?

c. WHEN will this system be MANUALLY Initiated, How is it manually initiated, and WHAT is expected to occur when manually initiated? s -

QUESTION 6.05 (2.00)

Answer the following questions in regard to LPCI loop select logic:

a. How does the logic determine how many recire pumps are running

[0.5]?

b. How does the logie determine which is the undamaged recire loop (1.0]?

c. If the logic determines neither loop is damaged, which loop will be selected for LPCI injection [0.5]?

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

6. PLJdfT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pc;;a 12 QUESTION 6.06 (3.00)

HULTIPULE CHOICE For each of the failures listed below choose the most correct answer.

Assume the failure exists at the time an automatic initiation signal is received for the High Pressure Coolant Injection System. Consider each failure separately. Assume normal 1985 plant temperature and pressure. ,

FAILURES

a. Auxiliary Oil Pump Shaft is seized. (Pump will not rotate.)

b. The Motor Gear Unit is failed at its' Low Speed Stop.

c. The bypass valve is open on the flow detector that feeds the Flow Controller.

CHOICES

1. HPCI will initiate and so to its runout condition (Maximum flow).

2. HPCI will initiate and inject as it is designed. (Normal flow).

3. HPCI will initiate but will not inject any water. (No flow)

4. HPCI will not initiate automatically or manually. (Does not start) s s

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l 8.- PLANT SYSTEMS DESIGN CONTROL. AND INSTRUMENTATION Pasa 13 l l

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QUESTION 6.07 (3.00)

The plant is operating at 50% power when a load reduction occurs which is equal to 10% power. There is an EHC diagram at the end this '

examination for reference.

a. Indicate which of the following EHC setting (s) are NOT at their normal setting for this plant condition. (0.5)

1. Load Selector 8 100%

2. Load Limit 9 1004

3. Max Combine e 105%

b. What would be the final condition /value for each of the initial parameters listed below. (Assume the EHC settings in part a.)

(5 9 0.5 ea)

1. Reactor Power 50%

2. Header pressure 935 pais

3. Turbine Speed 1800 rpm

4. Bypass Valves e valves open

5. Control Valves 9 their 50% flow position.

QUESTION 6.08 (2.50)

With regard to the Emergency Diesel Generators answer the following.

a. The emergency start relays prevent engine shutdown trips from all trips except 1 , 2 and 3 . [0.75]

, b. What position (s) are the 1/2 Deisel Generator Keylock switch (es) normally in? (Assume 100% power and no system faults.) [0.25]

c. What are six conditions that will initiate an automatic emergency

< start sequence? (Assume the DG control switch on the 90X-8 Panel is in " Auto") (Setpoints are not required for full credit.) [1.5]

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6.- PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Peca 14 1

QUESTION 6.09 (3.00)

Refer to attached Figure 5, Recirculation Speed Control Network, for the following:

a. The plant is operating at 30% power with both recirc pump M/A transfer stations in MANUAL. FOR EACH of the following instances, INDICATE HOW the speed of Recire Pump 'A' would change (Increase, i Decrease, or Remain the Same) AND WHICH components (s) of the control system is(are) limiting. (2.0)

1. Recirc Pump 'A' M/A transfer station is placed in ' AUTO'.

2. The generator speed tachometer output feedback signal fails low due to a loss of continuity through the field breaker contact.

b. WHAT action must be taken by the control room operator prior to resetting a ' LOCKED OUT' scoop tube? WHY? (1.0) l QUESTION 6.10 (2.00)

The control room emergency zone is protected against some toxic gases by a control room HVAC System isolation.

a. WHAT gases are monitored by this isolation system (1.0)

b. WHAT indication is available to alert the operators the isolation system has actuated? (0.50)

c. HOW is the control room emergency zone protected against toxic gases, that are not monitored by the Automatic Isolation System.

(0.50)

(***** END OF CATEGORY 6 *****) I

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'f. NROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pasa 15 AND RADIOLOGICAL CONTROL l

QUESTION 7.01 (2.00)

Answer the following questions concerning the seven caution notes associated with the QGA (Symptom oriented procedures),

a. Complete the sentence according to general caution #1; "If an entry condition for a QGA procedure occurs, enter that procedure irrespective'of whether that procedure ..." (0.5)

b. According to caution #3, HOW is the Drywell temperature determined?

(Include order of preference.) (1.0)

c. As discribed by caution #5, EXPLAIN when Yarway indicated levels are not reliable and WHICH level instrument should be used in its place?

(0.5) i QUESTION 7.02 (2.50)

What are the entry conditions for QGA-200 Primary Containment Procedures.

(Five required for full credit with parameter values.) .

QUESTION 7.03 (2.00)

According to QOA 202-1 " Jet Pump Failure", there are two indications or symptoms where simultaneously occurring would indicate a failure of a jet pump.

WHAT are these two indications?

(Be specific, include setpoints if needed.)

QUESTION 7.04 (3.50)

If while operating at 955 power, with EGC in operation, an 'A' heater high level alarm and a step increase in APRM readings are received.

WHAT Immediate Action Steps would be taken to control reactivity addition? (Limit your answer to the immediate action steps required by QOA 400-1 " Reactivity Addition") (An action step may contain more than one action item.)

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

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7.- PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pasa 16 AND RADIOLOGICAL CONTROL QUESTION 7.05 (2.00)

Answer the following questions according to QOA 1700-5, Main Steam Line High Radiation, immediate operator action steps for High Radiation on both channels.

a. What are six of the automatic actions verified by step 1.b.

" Verify Automatic Actions ...". [1.5]

b. Step 1.c. has the operator place the Off-Gas isolation valve AO-5406 control switch to close. EXPLAIN the BASIS for taking this action? [0.53 QUESTION 7.06 (2.00)

During the performance of QGP 1-2, Unit Startup to Hot Standby, when Reactor Pressure reaches approximatly 875 psig rods are notched in to stop the pressure rise and pressure is maintained between 850 psig and 920 psig.

- WHAT are the four methods allowed to maintain RPV pressure within the specified band?

4 QUESTION 7.07 (2.00)

QGP 4-1, Control Rod Movements and Control Rod Sequences, specifies the NSO's responsibility in regard to use of the Control Rod Sequence Package prior to normal in-sequence control rod movements.

tLu LIST ase four items the NSO must verify prior to commencing normal in-sequence control rod movements? (An item may consist of more than one action.)

QUESTION 7.08 (2.50)

Procedure QOA 201-3, Inadvertent Actuation of One Main Steam Relief Valve, requires six Immediate Operator Actions in the event a SRV inadvertently j opens.

WHAT are five of these actions?

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

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7.- PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pzca 17 AND RADIOLOGICAL CONTROL QUESTION 7.09 (3.25)

If operating in shutdown cooling and both recirc pumps are lost, QOP 1000-5 (SBUTDOWN COOLING START-UP AND OPERATION) states to monitor for temperature stratification.

Stratification is indicated by increasing (A.1.) without a corresponding (A.2.) or (A.3) change.

a. List parameters A.1., A.2., and A.3. (0.75)

b. List five of the suggested actions to minimize stratification if both recirc pumps are off. (2.5)

QUESTION 7.10 (2.00)

QGP 2-3, Reactor Scram procedure prerequisites states "If su.*ficient time is availabe, before scramming the reactor, perform one or more of the following:"

One of the Prerequisites is "If the unit is operating in EGC, trip EGC and return recirculation flow control to Manual."

WHAT are the four remaining action steps that should be performed prior to scramming the reactor 7 (An action step may contain more than one action item.)

QUESTION 7.11 (1.50)

Concerning the use of radiation dose meters, answer the following TRUE or FALSE.

a. Earphones are only required in areas of low visability conditions, be A GM detector is preferred for setting dose rates.

c. Before entering a suspected radiation area, the meter selector switch should be turned to the highest range.

(***** END OF CATEGORY 7 *****)

8.- ADMINISTRATIVE PROCEDURES. CONDITIONS. Pr.ca 18 AND LIMITATIONS QUESTION 8.01 (1.00)

According to QGP 1-1, Normal Unit Startup, if a procedure is terminated at any time during its execution, several administrative requirements must be met.

List four of these requirements that must be performed.

QUESTION 8.02 (1.50)

Per the Technical Specifications Limiting Safety System Settings; SETPOINTS NOT REQUIRED.

a. WHAT are two protective actions designed to prevent exceeding the Reactor Coolant System pressure safety limits? (1.0)

b. WHAT is the protective action that is provided as a backup for the two protective actions in "a." above. (0.5)

QUESTION 8.03 (2.50)

a. List TWO requirements to make a temporary change to an operating procedure. (Assume the intent is not changed.) (1.5)

b. What is the effective time a temporary change can be used without renewal? (0.5)

c. TRUE or FALSE A temporary change is required to alter a valve lineup check list. (0.5)

QUESTION 8.04 (2.00)

Under WHAT four conditions will the Shift Engineer accept direct responsibility for the activities and operations in the Control Room?

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8.- ADMINISTRATIVE PROCEDURES. CONDITIONS. Pe.cs 19 AND LIMITATIONS QUESTION 8.05 (3.00)

According to Technical Specifications, what is the basis for the following:

a. Condenser Low Vacuum Scram

b. Turbine Control Valve Fast Closure Scram QUESTION 8.06 (2.00)

According to QAP 300-3, Shift Manning procedure, each operating shift for Units 1 and 2 normally consists of WHAT man power for normal power conditions?

(Include the number of people for each position and type of license, if one is required.)

QUESTION 8.07 (3.50) ,

a. What action is required by any individual who exceeds his approved exposure limit at Quad Cities Station? (1.5)

b. What are the le CFR limits for;

1. whole body exposure (assume NRC-4 completed)?

2. skin of whole body?

3. hands and feet? (3 0 0.5 ea).

c. TRUE or FALSE Pocket type direct reading dosimeters are used to furnish the i

exposure data for the 10 CFR limits. (0.5)

(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)

8 .- ADMINISTRATIVE PROCEDURES. CONDITIONS. Pzcs 23 AND LIMITATIONS QUESTION 8.08 (1.00)

In,accordance with 10 CFR 55, " if a licensee has not been actively performing the functions of an operator or senior operator for a period'of -----?----- months, or longer, he shall, prior to resuming activities' licensed pursuant to this part, demonstrate to the Commission that his knowl'6dse and understanding of facility operation and administration are' tisfactory. "

Select from the below list 4 he correct answer to complete the above statement.

a. 2 A '

b. 4  ; .' , -

c. 8 \

d. 12 QUESTION 8.09 (1.50)

According to the General Limiting Conditions for Operation Technical Specifications 3.0/4.0 it states: "When a system, subsystem, train, component, or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

a. WHAT additional conditions must be met inorder to satisify this limitation? (1.0)

b. WHAT plant modes of operation does this Limiting Condition for Operation NOT apply? (0.5)

)

)

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(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)  ;

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8 .- DMINISTRATIVE PROCEDURES. CONDITIONS. Paga 21 AND LIMITATIONS QUESTION 8.10 (2.50)

Regarding Unit 2's Technical Specification on Recirculation Pump Flow Mismatch:

- a. What is the allowed mismatch between the recirculation pumps?

(Include parameters and limits for the full spectrum of operations.)

[1.5]

b. What is the required action if the above Limiting Condition for Operation can not be met? [0.5]

c. What is the basis for this Limiting Condition for Operation? [0.5]

QUESTION 8.11 (2.50)

a. How does an operator distinguish between an Personnel protection card and an Out-of-Service card? (1.5)

b. The miniature Out-of-Service card is authorized to be used for WHAT? (0.5)

c. Whenever a piece of equipment is taken out-of-service, the Master Out-of-Service card will be hung on the pegboard in the Control Room.

WHAT is/are the exception (s) to this and WHERE is/are they hung?

(0.~ 5 )

QUESTION 8.12 (1.00)

Unit 2 Technical Specifications require that the coupling integrity of a Control Rod be demonstrated by withdrawing the rod to the Fully Withdrawn position and verifying that the rod does not go to the overtravel position.

LIST two circumstances under which the Technical Specifications REQUIRE that this coupling check be performed.

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(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****)  !

l

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8 .* ADMINISTRATIVE PROCEDURES. CONDITIONS. Pcca 22 AND LIMITATIONS I

QUESTION 8.13 (2.00)

According to a NOTE in the Conduct of Shift Operation QAP-300-2 it states, "If an emergency situation should arise prudence may require allowing a Unit Operator to leave the 'at the controls' of his/her

' Stable and under control reactor' to help."

a. A Quad Cities operator is considered 'At the Controls

• when he/she is in WHAT area? (0.5)

b. These emergency situations are defined as rare and serious circumstances which if uncorrected could result in, WHAT three items? (0.75)

c. The control room supervisor may authorize a Unit Operator to leave his 'At the Controls' if his/her ' Stable and under control reactor' to help on the other Unit, only if WHAT three conditions are met? (0.75) l

(***** END OF CATEGORY 8 *****)

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

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, EQUATION SHEET

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t = ma v - s/t 2

• v = as a = v,t + lsac Cycia efficiency = 'I '

E = ac 2 .

, , Cyg , V,)jt gg . Iggy 2 y +g A = AN "

.y

, A = A,e PE = mah . = e/t A = h 2/t,s = 0.693/t,i ,

W = vaP' g(aff) = (t,,)(q) t AK = 931Am .

( , )

6-5C,aT ,

I - I,a -Ix

, Q = UAAT I . I

1. X Pwr = Wg ' a"

, I-I gto -xh l

y.y to M (t). TVI. = 1.3/u y.y .t/T -

RVI.

• 0.693/u a .

'SUR = 26.06/T _

T = 1.44 DT SCR = S/(1 - E,gg)

' sh f gA {f, SUR = 26 CR , = S/(1 - E,gg ,)

I aff)1 Il Inff)2 T = '(1*/o ) + [(i 'o)/A,ggo ] I 2 T = g*/ (p . py M " 1/(1 - K,gg) = CR g/CR0 l I

• II ~ 8)I aff' M = (1 - K,gg)0 0 ~ aff)1 8 " IEsff'I)I aff " #eff/Kaff SDM = (1 - K,fg)/K,gg

~

p= [1*/TK,'gg .] + [I/(1 + 1,ggT )] 1* = 1 x 10 seconds y = I4V/(3 x 10 0) A,fg A? 0.1 seconds

~

E = Na -

Idgg=Id22 WATER PARAMETERS Id g =I02 1 gal. = 8.345 lba R/hr = (0.5 CE)/d 2 (,,,,,,)

I gal. = 3.78 liters R/hr = 6 CE/d (feet) -

1 fg = 7.48 gal. MTSCELI.ANEOUS CONVERSIONS .

3 Density = 62.4 lba/fc 1 Curia = 3.7 x 10 dps 10 Density = 1 gm/cm 1 kg = 2.21 1ha Heat of vapori:stion = 970 Ecu/lba 1 hp = 2.54 x 103 BTU /hr Heat of fusica = 144 Beu/lba 6 1 Hw = 3.41 x 10 Btu /hr 1 Atm = 14.7 psi = 29.9 in. I's. 1 Stu = 778 f t-lbf 1 ft. H 2O = 0.4333 lbf/in 1' inch = 2.54 cm F = 9/5*C + 32

C = 5/9 (*F - 32)

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5, THEORY OF NUCLEAR POWER PLANT OPERATION. Pcca 23 FLUIDS.AND THERMODYNAMICS s 4 3

po

$g v/4l w. u) 4"6

w. '{ .;

gaf ANSWER 5.01 (1.00)

b. doppler coefficient and beta bar ,

REFERENCE G. E. Reactor Theory, pg. 70 WNP-II Version f Quad Cities Rx Theory Coefficients of Reactivity pg 54 of 85 and Reactor Kinetics pg 40 ANSWER 5.02 (1.00)

c. It has less effect on reactor operation than Xe-135 due to its smaller fission yield and smaller microscopic neutron cross section.

REFERENCE

( G. E. Reactor Theory, pg. 87 WNP-II Version Quad Cities Rx Theory - Fission Product Poisons pg 72 ANSWER 5.03 (1.50)

a. increase

b. decrease

c. increase [3 0 0.5 ea]

REFERENCE l G. E. BWR Training Center Thermodynamic Heat Transfer and Fluid Flew Quad Cities Thermodynamic Heat Transfer and Fluid Flow pg 7-55 thry 7-58 l ANSWER 5.04 (1.00)

d. HYDROGEN atoms in the water molecules t

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

5 TurORY OF NUCr.rAR POWER PLANT OPERATION. Pe.ca 24 FLUIDS.AND THERMODYNAMICS REFERENCE G. E. Reactor Theory and Heat Transfer Section II Neutron Physics Quad Cities Rx Theory, Fuel Nuclei and Neutron Properties for Fission Pg 12 ANSWER 5.05 (1.50)

a. Less

b. More

c. More [3 0 0.5 ea]

REFERENCE HTFF, pages 53, 54, and 58 Quad Cities Reactor Theory Coefficients of Reactivity pg 50, 52, 54 1

ANSWER 5.06 (3.00)

a. Head increases [0.5] the pump is still putting the same amount of work into the fluid, therfore the same delta pressure across the pump, so as the auction pressure increases so will the discharge head [0.5]

b. Head increases [0.5] as system resistance to flow increases, pump head increases [0.5]

c. Head decreases [0.5] as temperature increases, system resistance to flow decreases (lower viscosity), therfore head decreases [0.5]

REFERENCE HTFF, page 25 4

Quad Cities, Heat Transfer and Fluid Flow Pump Characteristics, pg 6 6-105

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

. *5 THEORY OF NUC MAR POWER PLANT OPERATION. Par;a 25 EQIDS.AND THERMODYNAMICS ANSWER 5.07 .(3.00)

a. Control rods in the center of the core are exposed to a higher thermal flux than those at the core periphery and, therefore, have a greater worth [9r 3. (Rod worth would be increased anywhere in the core where a fa"+ dial thermal neutron flux peak existed even if it were an edge rod)[".5E

b. As moderator temperature increases, its density decreases resulting in longer thermal diffusion lengths. This allows thermal neutrons to travel further and be absorbed by the rod; [0.5] thus, rod worth increases with an increase in moderator temperature [0.5].

c. As voids increase, again less moderations takes place. Again, thermal neutrons travel further; however, voids tend to depress thermal flux because of the very poor moderation [0.5]. Thus, as voids increase, rod worth decreases [0.5].

REFERENCE QC Reactor Theory, pages 58 and 60 ANSWER 5.08 (2.00) -

a. CR1/CR0 = 1-Keff0/1-Keff1 CR1 = X, CR0 = 10, Keffe = .96, Keffi = .995 X/10 = 1 .96/1 .995 X/10 = .04/.005 X/10 = 8 X = 80

[ formula = 0.75, math = 0.25]

b. The time to reach an infinite period would be greater, [0.5] due to the fact that there are more generations, each representing a period of time, required to reach equilibrium [0.5].

REFERENCE QC Theory, pages 32,34 and 36

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

l

, . _ ~ . . _ _ __ __. _ _ ,,_ . __,,.,_ ,,___. - ... ,_ . . _ _ . _ , _ _ _ _. _ _ _ _ _ _ , . . . _ _ _ . _ , , . , _ _ . . , , , _ . , _ . - , . . . _ . _ _

5.. W nRY OF NUC uAR POWER PLANT OPERATION. Pasa 26 50 IDS.AND THERMODYNAMICS ANSWER 5.09 (2.00) r

a. The MAPRAT of 1.02 is not conservative [0.5]. With a MAPRAT greater than one it means that the MAPLBGR has been exceeded because: fM.v vi. o,. '

MAPRAT =eMAPLHGR (actual)/MAPLEGR (LCO)

, (Actual formula is not required but the relationship explained is.) [0.5]

b. 1. False

2. True [2 0 0.5 ea]

REFERENCE HTFF, page 31 Quad Cities Heat Transfer and Fluid Flow pg 9-24a i

ANSWER 5.10 (2.00)

a. Neutron level would start and continue to increase [0.25] until the point of adding heat is reached [0.25]. As the coolant heats up, negative reactivity is added and power turns [0.25]. Power

would stablize at the point of adding heat [0.25].

b. Period would take a step jump due to the production of prompt neutrons [0.25]. Immediately after this step, the rate of power change decreases to a rate controlled by delayed neutrons until the reactivity is no longer being increased [0.25]. Then a sharp drop would occur as the rate of reactivity addition drops to zero [0.25]. A stable period would continue until negative reactivity is inserted. Stabilizes at infinity [0.25].

REFERENCE Dresden NUS Theory ,

Quad Cities Theory Review pages 44-45 and 78. '

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(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

4 e

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' 5m ' THEORY OF NUCLEAR POWER PLANT OPERATION. Peca 27 FLUIDS.AND THERMODYNAMICS 1

ANSWER 5.11 (1.50)

~ 1. They are physically located as far below the normal water line as possible to provide the greatest static head.

2. With feed flow less than 20% they are kept on minimum speed.

3. At high power operation adequate NPSH is obtained from feedwater subcooling.

4. Low reactor Vessel water level trip, cavitation interlock.

5. Suction valve closed trip, cavitation interlock.

[3 0 0.5 ea]

REFERENCE DRESDEN - Recire System Lesson Plan pg 16 & 18  !

GE Thermodynamics, Heat transfer & Fluid Flow, page 7-93 & 94 Quad Cities LIC 0202-1 pg 19 and 21 ,

ANSWER 5.12 (1.50)

a. F3. [0.25] L2. [0.25]

b. F1. [0.25] L3. [0.25]

c. F2. [0.25] L1. [0.25]

REFERENCE EIH: GPNT, Vol VII, Chapter 10.2-23 Quad Cities Heat transfer and Fluid Flow pg 9-16a, 9-19a, and 9-34a ANSWER 5.13 (1.50)

1. Fuel rod power level

2. Fuel rod exposure

! 3. Rate of power increase l 4. Fuel pellet design

5. Previous power history

6. Presence of embrittling agent

'l * (~3~0 0 5 es5hT - - Ar A * # r~' ,b._ s -

l

(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)

' 5. . THEORY OF NUCMAR POWER PLANT OPERATION. Pasa 28 FLUIDS.AND THERMODYNAMICS 1

REFERENCE l Fuel System Description, pg 15, 16 GE HTFF pg 9-107, 108 l Quad Cities Heat Transfer and

• u Fluid Flow o*

pg 9-45a,en&  ?

9-46a,,,, /. g f, )

4: *- *

.,... t 4: , ,;.,>., _

r ANSWER- 5.14 (3.00)

a. No significant effect

b. Less rod withdrawal

c. More rod withdrawal

d. No significant effect

e. Less rod withdrawal

f. More rod withdrawal [6 4 0.5 ea)

REFERENCE Reactor Physics Review, Reactivity Coefficients I Quad Cities Rx Theory pg 59, 60, and 81 l

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

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, 8.. ')LANT SYSTEME DESIGN. CONTROL. AND INsinUrENTATION Paca 29 ANSWER 6.01 (2.00)

a. Rod Block

b. Half-Scram '

c. -Red-8&eele i- - < 8"

d. No Reactor Protection System Action [4 0 0.5 ea]

REFERENCE Quad Cities RPS and APRM Lesson Plans ANSWER 6.02 (3.99)

a. Valve 1 Valve 2

1. Hydraulic [0.25] Pneumatic [0.25]

2. Drag Disc [0.25] Globe (double) [0.25] ~

3. Loss of power to the Loss of the Controller pump (hydraulic) [0.25] output; signal [0.25]

Low oil pressure [0.25] Low air pressure (0.25] -

Loss of Control Signal [0.25] M " c'" * * <w -

- / ^ C# 2 L'

b. Compares Total Steam Flow with turbine first stage pressure [0.25]

with a mismatch of greater than 105 [0.25] longer than 30 seconds

[0.25] an alarm is actuated.

REFERENCE Quad Cities Yessel Level Control System Discription LIC-0600 ps 6, 8, 28 &

30.

I

ANSWER 6.03 (1.00) 4 a. - 2.

b. - 6.

c. - 7.

d. - 4. [4 0 0.25 ea) ,

! REFERENCE Quad Cities Fire Protection Sys LIC-4100-1 pg 4, 6, 12 & 14

(***** *****)

CATEGORY 6 CONTINUED ON NEXT PAGE y ---.-w,. ,.-wg-- - - - - - - - - - - , - - -y---e, - w,. m.,n+eee.--ww.e-r

i

[- 'E. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pcco 33 I i

\

_ l ANSWER 6.04 (-3-067--

a. 1. Recirculation Flow Control failure with/ increasing flow l 2. Turbine Generator Trip l

3. Closure of all Main Steam Isolation Valves

4. Pressure Regulator Failure in Open Direction

5. Feed Water Controller Failure to Maximum Demand

6. Loss of Feedwater Flow <

7. Loss of Auxiliary Power [4 4 0.25 ea]

b. 1. High Reactor Pressure [0.25] 1250 psig [0.25]

2. Low-Low water Level [0.25] -59 inches (0.25]

c. In the event a scram occurs or should have occurred and reactor power [S.25] is still above 315 or can not be determined [0.25].

It is initiated by depressing BOTH ATWS Division pushbuttons simultaneously [0.25] which energizes the ARI valves 'for 30 seconds) [0. 25] . 3 . .. .,..._,

,. ., . .s 3 -. . , . , -

.w & M . L.: % , . , t_,.

% ....o:)

m: -EU. ~~?

.....s

. .,.o . ..

c / 17. !, : La C .. ..su... ya .* .-- '

Quad Cities Anticipated Transient Without Scram LIC-0300-3 pg 3, 4 & 18.

ANSWER 6.05 (2.00)

a. By monitoring the differential pressure across each recire pump

[0.5]

b. By comparing the pressure in the riser pipes en one recirc loop with the pressure in the riser pipes of the other loop [0.5]. The undamaged loop will have a higher pressure than the damaged loop

[0.5].

c. Loop B [0.5]

REFERENCE Quad Cities RER System Lesson Plan LIC-1000-1 pg 44 ANSWER 6.06 (3.00)

a. 4

b. 3

c. 1 [3 0 1.0 ea]

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

'8. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pcca 31

. REFERENCE Quad Cities High Pressure Coolant Injection LIC-2300 pg 11, 12 & 31 ANSWER 6.07 (3.00)

a. 1. Load Selector ( should be 10 % above Load) [0.5]

b. 1. Rx Pwr 9 504 EWJ

2. Bdr Press'937-986 psig # ins..;;; h;;;;; ef r" el e BiestCf 3 9)

3. 1854 RPM (45 rpm to overcome load selector 9 rpm to over come 10% load error) '. G

4. BPV2 valves 100%3rd25%openor21/4valvesopenId2l

5. CV's 40% flow positionFJ:V [5 0 0.5 ea)

]

REFERENCE l

Quad Cities EHC Pressure Control and Logic LIC-5650-2 pg 12 - 18, 21, 22 and figure 6 QGP 1-1 Normal Unit Startup pg 17 ANSWER 6.08 (2.50)

a. 1. Overspeed

2. Differential Overcurrent

3. Overcranking [3 0 0.25 ea]

b. Off [0.25]

c. 1. Drywell pressure (2 psig)

2. Low-Low Rx Water Level (-59 in)

3. Undervoltage on 14 (13)

4. Undervoltage on the respective Emergency bus (13-1 or 14-1)

5. Breaker from 13 to 13-1 or 14 to 14-1 open

6. Normal Feed breakers to bus 14 (13) open

7. Backup undervoltage or bus 14-1 (13-1) (with a 5 minute time delay.) [6 0 0.25 ea]

REFERENCE Quad Cities Emergency Diesel Generators LIC-6600 pg 14, 15 and 16

(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)

'6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pasa 32 ANSWER 6.09 (3.00) I

a. 1. Increase [0.5] Masterlimiter,(lowspeedlimi [0.5]

2. Increase [0.5] Scoop tube positioning unit or electrical limit switches [0. 5](q.f. wa.a. W W hL M)

b. The null voltmeter is used to match the speed from the tachometer with the speed demand from the speed controller [0.5], to prevent a flow transient during reset [0.5].

REFERENCE Quad Cities Recirc Flow Control, LIC-0202-2, pg 10, 12, 14, & 36 QOP 202-6 Resetting Recirc M/G Set Scoop Tube Rev 3 pg 1 ANSWER 6.10 (2.00)

a. 1. Ammonia

2. Clorine

3. Sulfur Dioxide [3 0 0.33 ea]

b. Control Room HVAC " Major Trouble" alarm. [0.50]

c. Manual Isolation by the operator. [0.50]

REFERENCE Quad Cities HVAC LIC-5750 pg 27

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(***** END OF CATEGORY 6 *****)

l a

7. PROCEDURES - NORMAL. ABHORMAL. EMERGENCY Pasa 33 AND RADIOLOGICAL CONTROL ANSWER 7.01 (2.00)

a. "has already been executed or is presently being executed. (Exact wording is not required for full credit) [0.5]

b. Drywell temperature is determined by reading the following sensors.

(The sensors are listed in order of prefered use.)

1. Process computer (point C166 (C266)) cn, Or> -

Process computer (point C165 (C265)) av 9 04

2. Valve leak recorder (260-20 pt 18)

Valve leak recorder (260-20 pt 19)

3. Drywell Environs Indicator (5741-170 pt 36)

4. Drywell Environs Recorder (5741-130 pt 32)

[4 sensors 0 0.2 ea = 0.8] [ Order of preference 0.23

c. Yarway level indicated levels are not reliable during rapid RPY e

dep'essurisatin below 500 psig [0.25]'. For these conditions, utilize the GEMAC instruments [0.25] to monitor RPV water level.

REFERENCE Quad Cities QGA-00 Rev 3 Caution #1 pg 1, Caution #3 pg 2, and Caution #5 Ps 3 ANSWER 7.02 (2.50) ,

1. Suppression Pool Temperature above 95 F -

2. Drywell temp above 180 F ',

3. Drywell pressure above 2.0 psig

4. Suppression pool water level above +2.0 in

5. Suppression pool water level below -2.0 in

[5 parameters 0 0.25 ea and 5 values e 0.25 ea]

REFERENCE Quad Cities QGA-200-1 Rev 1 pg 1 ANSWER 7.03 (2.00) ,

1. The recirculation pump flows (0.34] differ by more than le percent

[0.33] from established speed-flow characteristics [0.33].

2. The indicated total core flow [0.34] is more than 10 percent greater than [0.33] the core flow value derived from established power-core flow relationships [0.33].

a

(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)

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. 7 'PROCEDUBES - NORMAL. ABNORMAL. THERGENCY Paca 34 AND RADIOLOGICAL CONTROL REFERENCE Quad Cities QOA-202-1 Jet Pump Failure Rev 2 pg 1  !

ANSWER 7.04 (3.50)

1. Trip EGC [0.25] return recire flow control to Manual [0.25]

2. Reduce recirculation flow [0.25] by at least 204 speed [0.25],

3. Insert control rods [0.25] starting at the present location in sequence and work backwards (0.25]. Any rod inserted should be continuously inserted all the way to position 90 [0.25] Aeven if the sequence calls for stopping at an intermediate position.) D :3 After each group of rods check the flow control line [0.25] and continue this process until the reactor is under the 100 % FCL

[0.25] p ;.0

4. If a LSSS has not been exceeded,everify that reactor power is within operating limits of exsisting recirculation system flow

[0.25]. If reactor power continues to increase,vScram the reactor

[0.25]. , D.CJ 0.

etif7 tr.e C .1 L L.iueer-ted5)*-

.  ::eti , LJi Live Leteettorrto-hawthem sample-reactor coolant# -

--Eer26t-for-fissitnr predtzet-activity -[8. 25] . .

RENTRENCE Quad Cities QOA 400-1 Reactivity Addition Rev 3 pg 1 & 2

. ANSWER 7.05 (2.00)

a. 1. Reactor Scram

2. hSIV's close

3. Off-Gas isolation valves close i 4. Air ejector suction valves close

5. Mechanical vacuum pump trips

6. Main Steam Line drains close

7. Primary sample valves close [6 4 0.25 ea]

b. .This will prevent re-opening of the Off-Gas isolation valves (0.25]

l when the high radiation condition is reset [0.25].

l REFERENCE

, Quad Cities QOA-1700-5 MSIV High Rad Rev 4 pg 1 l

l

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(***** CATEGORY 7 CONTINUED ON NEET PAGE *****) l

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l _ . _ _ . _ _ _ , , . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _

7. 'PEOCEDURES - NORMAL. ABNORMAL. EMERGENCY Pasa 35 AND RADIOLOGICAL CONTROL ANSWER 7.06 (2.00) ,

1. Operate RCIC

2. RWCU system reject flow

3. Vary power with CR's '

4. SRV manual operation (to prevent reaching 1960 psig if the other i three methods fail to maintain pressure.) )

[4 0 0.5 ea] l REFERENCE Quad Cities QGP 1-2 Unit Startup to Hot Standby, pg 6 ,

ANSWER 7.07 (2.00) J.. d

1. (QTP 1600-S2,) Control Rod Sequence Review Sheet,has been reviewed and approved.

2. The RWM has been loaded and verified (by checking for the sign-off on QTP 1600-S2, Control Rod Sequence Review sheet.)

3. (QTP 1600-S3,) Control Rod Sequence Sheet has been signed and dated by both a Qualified Nuclear Engineer (QNE) and an on-shift SRO before moving control rods in.the group.

4. Either the Rod Worth Minimizer (RWM) is in NORMAL or, if in BYPASS, that a caution card has been placed on the rod movement control switch (per procedure QOP 207-2.)

5. Permission is obtained from the Shift Engineer to commence rod motion.

[4 0 0.5 ea]

REFERENCE Quad Cities QGP 4-1 Control Rod Movements end Control Rod Sequences, pg 4 4

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• 7. ' PROCEDURES - NORMAL. ABNORMAL. EtERGENCY Paga 36 AND RADIOLOGICAL CONTROL ANSWER 7.08 (2.50)

1. Attempt to close the affected relief valve by placing its key-lock I switch to OFF.

l 2. Verify all reactor parameters are within operating limits.

l

3. If the valve will not close manually, determine the cause of the salfunction and attempt to close the relief valve.

4. If the valve cannot be closed SCRAM the reactor, and refer to QOA 201-2. ,

l l 5. IF an entry condition for a QGA procedure occurs, THEN enter that l procedure.

6. If pool Temperature reaches ISO F, perform an external' torus inspection.

[5 0 0.5 ea]

REFERENCE Quad Cities QOA 201-3, Inadvertent Actuation of One Main Steam Relief l i

Valve, pg 1 ,

I ANSWER 7.09 (3.25) t

a. 1. Reactor metal temperature

2. Vessel level

3. Water temperature [3 0 0.25 ea)

(Items a.2. and a.3. maybe reversed.)

b. 1. Increase shutdown cooling flow

2. Raise vessel water level

3. Use head spray

4. Startup the RWCU system

5. Startup both CRD pumps

6. Bleed through cleanup system and feed through feedwater i system.

l

7. Flood the main steam lines, and drain through the main steam line drains if the cleanup sytem is not available.

[5 9 0.5 ea)

REFERENCE QC QOP 1000-5 pg 2

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. '7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Pass 37 AND RADIOLOGICAL CONTROL l

1

( ANSWER 7.10 (2.00)

1. Reduce power with recirculation (flow to reduce the transient.)

2. Transfer auxiliary power to T12 (T22).

3. Start the turbine main shaft suction pump and the EBOP

4. Raise Reactor Water level to high level alara point (~+44")

[4 0 0.5 ea]

REF1RENCE Quad Cities QGP 2-3 Reactor Scram, Rev 23, pg 1 ANSWER 7.11 (l.50)

a. False

b. False

c. True [3 0 0.5 eal REFERENCE Quad Cities Rad Protection LIL H-3 l

l l

l l

1

(***** END OF CATEGORY 7 *****)

l ..

8. 1DMINISTRATIVE PROCEDURES, CONDITIONS. Pro 38 AND LIMITATIONS ANSWER 8.01 (1.00)

1. Insert statement at point of termination as to reason for termination.

2. Insert title of subsequent procedures at point of termination

3. Insert NA in all remaining blanks

4. Date, time, and sign the last page

5. Retain the terminated procedure in the startup package.

[4 0 0.25 ea]

REFERENCE Quad City QGP 1-1 Sec C.8 pg 3 ANSWER 8.02 (1.50)

" ' ~

a. 1. High Neutron Flux Scram v~ uwLr-

2. Safety valve actuation [2 0 0.5 ea)

b. High Pressure scram [0.5] ~ p.'

n M,

% v.)M ." ' ' " 'w e' ' -

/ e. b..

,[ a : 4 ._

REFERENCE Quad Cities Technical Specification 1.2/2.2 Reactor Coolant System pg 1.2/2.2-1 and 2.2 Limiting Safety System Setting Bases pg 1.2/2.2-3 ANSWER 8.03 (2.50)

a. 1. The change is approved by two members of the plant management staff at least one of whom holds an SRO

License on the unit affected. O. Si 2.

The change is documented, reyiewed by the onsite review and investigative function & approved by the station - - .

i superintendent within 14 days of implementation. (4-MP) J Q,C 33 D

b. 30 days (0.5) l

c. False (0.5)

REFERENCE QC Tech Specs. 6.2b and QAP 1100-5 PG 2 l

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c- . . - _ _ . . _ - , _ , , . -,.._ ,_.,,. -. -,.. ,-,, , ,

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'8. ' ADMINISTRATIVE PROCEDURER. CONPITIONS. Poco 39 AND LIMITATIONS

)

1 ANSWER 8.04 (2.00) I l

1. During any transient which may lead to failure of fuel rods. l

2. A radioactive release which has the potential.of exceeding the l Technical Specification limits.

3. A failure of the reactor primary coolant system. '

4. A failure of the containment barriers.

5. The Loss of the function of a safety system. [4 0 0.5 ea] i REFE b.4 Any GSEP condition: reference QAP 300-2 page 11 '

2. i Flooding: reference Q0A 010-4 page 2

3. ? Fire: reference QEP 340-5 page 3 Lon, Rev 13 C.3.c '

.* Scram: reference QAP 300-1 page 5 '

J When relieving the SCRE during normal plant operation:

reference QAP 300-1 page 6.

ANS h I# 3 'O ' ' ' 'd'@ '" ~ ~ ~ ~ ' ' ' '

l

a. (Loss of condenser vacuum occurs when the condenser can no longer  !

handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser.) Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux [0.53 To prevent the cladding safety limit from being exceeded if this occurs,

[0.5] (a reactor scram occurs on turbine stop valve closure in the Run mode. The turbine stop valve closure scram function alone is adequate to prevent the cladding safety limit from being exceeded in the event of a turbine trip transient with bypass closure.)

The condenser low vacuum scram is anticipatory to the stop valve closure scram and causes a scram before the stop valves are closed and thus the resulting transient is less severe [0.5]. (Scram occurs in the Run mode at 21 inch Hg vacuum stop valve closure occurs at 20 in Hg vacuum, and bypass closure at 7 in Hg vacuum.)

b. The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux [0.5]

resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass. [0.5] i.e.,

it prevents MCPR from becoming less than the MCPR fuel cladding integrity safety limit for this transient [0.5]. (For the load rejection without bypass transient from 100% power, the peak heat flux (and therefore LHGR) increases on the order of 15% which provides wide margin to the value corresponding to 1% plastic strain of the cladding.)

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l 3- . -- y .,y,----------.-------,w,- , ,o, .. . . , ,_. , -. ,--,,

_ ,-,--.-,,..-#m_,, - , , , - .- - - , - - .----,---,---...-,--,-.,,-.-,----r.-,,- em,..-~ , , . . w. -

p. ADMINISTRATIVE PROCEDURES. CONDITIONS. Page 4D AND LIMITATIONS REFERENCE Quad Cities Tech Spec Sec 2.1 Limiting Safey System Settings Basis. part f, pg 1.1/2.1-9 and part j pg 1.1/2.1-10 ANSWER 8.06 (2.00)

1. One, Shift Engineer, SRO

2. Two Shift Foreman, SRO

3. Three, Nuclear Station Operators, RO

4. One Equipment Operator, None

5. Five, Equipment Attendents, None

6. One, Station Control Room Engineer (SCRE), SRO

[18 0 0.11 ea]

REFERENCE Quad Cities QAP 300-3, Shift Manning C.3 Rev 8 pg 2 and QAP 300-1 Op Dept Org Rev 13 pg 2, 7, 9, 13, 16, & 17 ANSWER 8.07 (3.50)

a. Any individual who exceeds his approved exposure level is required to promptly report his exposure to radiation protection and his own supervisor. (1.5)

b. 1. 3 rem /qtr not to exceed 5 (N-18)

2. 7 1/2 rem /qtr

3. 18 3/4 rea/qtr (3 0 0.5 ea)

c. False (0.5)

REFERENCE LIL: RAD PRO Lesson Plan Quad Cities QAP 1120-2 pg 54 ANSWER 8.08 (1.00) ,

- "] # '

_.lt - G ,2 m;2 ' r '

d

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'B- *DMINISTEATIVE_EBOCEDURESmCONDITIONS. Pega 41 AND LIMITATIONS REFERENCE FERMI - 10 CFR 55.31.e Quad Cities 10 CFR 55.31.e ANSWER 8.09 (1.50)

a. 1. Its corresponding normal or emergency power source is Operable.

[0.5]

2. All of its. redundant systems, subsystems, trains, components, and devices in the other division are operable, [0.5] or likewise satisfy the requirements of this specification.

b. Specification 3.0.B is not applicable in refueling [0.25] or cold shutdown. [0.25]

REFERENCE Quad Cities Technical Specifications 3.0/4.0 General LCO B. pg 3.0/4.0-2 ANSWER 8.10 (2.50)

a. Pump speed [0.25] shall be maintained within 10% [0.25] of each other when power level [0.25] is greater than 80% [0.25] and within 15% [0.25] of each other when power level is less than 80%

[0.25].

b. One recirculation pump shall be tripped. [0.5]

c. To prevent the possibility of LPCI Loop Selection logic selecting the wrong loop for injection [0.5].

REFERENCE N Quad Cities Technical Specifications 3.6/4.6 Primary System Boundary, 3.H.1 & 3.H.2 Pg 3.6/4.6-5 & Basis 3.6.H. pg 3.6/4.6-13 f

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~ . _ . - _ _ _ __ _ _ _ _ . _ - - _ . _ _ _ _ _ _ _ . . _ _ . . _ . _ _ _ _ _ _ _ _ _ . . _ . _ _ . - . _ . _ _ . . _ _ _ . . . __.

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s, '8 s

• ADMINISTRATIVE PROCEDURES. CONDITIONS. Page 42 l AND LIMITATIONS '

i I

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ANSWER 8.11 (2.50) l

a. Personnel Protection cards are white [0.25] with blue letters [0.25]  !

and a blue border [0.25] while the Out-ofService cards are white )

[0.25] with red letters [0.25] and a red cross-hatched border [0.25] )

b. Will be used for holding equipment out-of-service on the control panels only [0.5]

c. The exception is when radwaste equipment is taken out-of-service in

[0.25] which case the Master Out-of-Service card will be hung in the radwaste control room. [0.25]

REFERENCE Quad Cities QAP 300-13 Tagging Equipment Rev 7 pg 1. QAP 300-14 Equipment Out-of-Service Rev 9 pg 1 & 2 ANSWER 8.12 (1.00)

1. When the rod is fully withdrawn the first time subsequent to each refueling. [0.5]

2. After maintenance. [0.5]

REFERENCE BSEP: TS 3.1.3.6 Quad Cities Technical Specifications 3.3/4,3 Reactivity Control pg 3.3/4.3-2 i

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'8 *' ADMINISTRATIVE PROCEDURES. CONDITIO@. Pet;s 43 AND LIMITATIONS ANSWER 8.13 (2.00)

a. At the controls is when the operator is physically within the operating area in front of the Unit panels. (Defined in QAP-300-I'8 ) [0.5]

b. 1. Injury to the public or Company personnel

2. Release off-site above T.S. limits

3. Damage to equipment, if such damage is ties to a possible adverse effect on public health and safety. [3 0 0.25 ea]

c. 1. A licensed operator has specifically been assigned the responsibility of monitoring the controls of the unit and responding to all unit alaras.

2. This same licensed operator remains within line of sight of the unit's front panel.

3. The licensed Operator, on a periodic basis, ~ 5 to le minutes reviews the status of that unit from within the area designed as being close proximity of the main control panels of the unit. (3 90.25 ea]

REFERENCE Quad Cities Conduct of Shift Operations QAP 300-2 Rev 16 pg 13 & 14, QAP 300-T8 -

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