ML20137R682

Exam Rept 50-338/OL-85-02 on 851015-18.Exam results:12 Candidates Passed Oral exams,11 Candidates Passed Simulator exams,11 Candidates Passed Written Exam & One Failed.Exam Questions,Answer Key & Facility Comments Encl

ML20137R682
Person / Time
Site:	North Anna
Issue date:	11/19/1985
From:	Lawyer S, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20137R667	List: ML20137R662 ML20137R682
References
50-338-OL-85-02, 50-338-OL-85-2, NUDOCS 8512060044
Download: ML20137R682 (200)

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 .                                                                                                                          y ENCLOSURE 1 EXAMINATION REPORT 338/0L-85-02

, Facility Licensee: . Virginia Electric and Power Company Richmond, Virginia 23261 Facility Name: North Anna Facility Docket Nos.: 50-338 and 50-339 Written, simulator and oral examinations were administered at North Anna near Mineral, Virginia.

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Chief Examiner: Sandy Lawyerj f Date Signed Approved by: mu- - /A Bruce /A. Wilson, Section Chief I ICi II Date $1gned

                     /

Summary

Examinations on October 15-18, 1985 Oral examinations were administered to 12 candidates; all of whom. passed. Simulator examinations were administered to 11 candidates; all of whom passed. Written examinations were-administered to 11 candidates; ten of whom passed. d 8512060044 85112038 DR ADOCK 050 i

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e . REPORT DETAILS

1. Facility-Employees Contacted:

D. C. Hawkins, Senior Instructor-Nuclear /SRO Lead Instructor W. Shura, Senior Instructor-Nuclear L. Edmonds, Superintendent, Nuclear Training L. R.-Buck, Supervisor-Training (Power Station OPS) R. O. Enfinger, Superintendent, Operations B.~L. Shriver, Director, Nuclear Training E. R.' Smith, Jr. , Assistant Station Manager (NS&L) M. D. Crist, Senior Instructor-Nuclear D. C. Fellows, Lead Simulator Instructor J. Ballard, Nuc? ear Training Auditor L. Johnson, Nuclear Traning Auditor (All Attended Exit Meeting)

2. Examiners:

        " Sandy Lawyer, NRC Examiner Tom Rogers, NRC Examiner Bill-Dean, NRC Examiner Gerry Douglas, NRC Examiner

• Chief Examiner

3. Examination-Review Prior to the exit meeting, the exasniners met with W. Shura, D. Hawkins, L. Edmonds, and R. Buck to' review the written examination and facility comments. The following are the NRC responses to the facility comments and recommendations of Enclosure 4.

a. R0 Exam (1) Question - Section 2 General Comments The comment states that some of the examination questions, particularly on Category 2, are not operationally oriented and are not supported by the (North Anna)- job task analysis '(JTA).

The NRC's responsibility,: at the present time, is to ensure that

                     . written examinations comply with the requirements of 10 CFR 55, specifically paragraphs 55.21 and 55.22.       NUREG-1021, Operator Licensing Examiner Standarcs, -provides clarif.ication concerning the scope and content of examinations. NUREG-1122, the recently issued K/A Catalog, provide:2 additional guidance to the examiners on testable knowledges and t.bilities. Also included in NUREG-1122 are relative- importance     ratings    that   serve as guides

2 in determining which topics should or should not be included in examinations. Both of these NUREG's are intended to clarify and provide guidance, but not supercede the pertinent regulations. Since the North Anna JTA was not provided to the NRC with the other reference material, we are unable to datermine if it encom-passes the requirements and guicance of the above three mentioned documents. It is therefore, not inconceivable that some NRC questions may not be supported by the North Anna JTA, nor meet the North Anna definition of operationally oriented. (2) Question 2,. 02 - We agree with the comment. Section (c) of Question 2.02 was deleted. (3) Question 2.05 - Step 4.1.14 in 1-0P-14.1 refers to manipulation of CC from tne RHR heat exchanger to ensure that CC return temp is

          <700 F while step 4.1.25 refers to positioning of RHR heat exchange outlet to control the cooldown rate. The question asked for the valve that the operator manually adjusts to reduce RCS temp. Only answer (b) is correct. No change required.

(4) Question 2.06 - This question was intended to determine the candi-dates knowledge of the pressurizer pressure control system and each distractor (including b) was specifically worded to test oc misconceptions or concepts important to satisfactory job performance. Since the wording of distractor (b) was ambiguous, (i.e., improved wording would be " .. each bypass line is sized to permit 1 gpm flow rate . . .") but is clearly incorrect, this choice cannot be accepted for credit. Due to the apparent con-fusion created, this question was deleted from the examination. We are concerned that no candidate was able to choose the correct answer, choice (c), which may indicate a training deficiency in this area. (5) Question 2.09 - We disagree with the recommendation and in part with the comment. This question tests knowledge of the design of an operationally important feature of the loop isolation valves. No change required. (6) Question 2.17 - We disagree-with the comment in that the question does not "ask for obscure descriptions and words taken out of context." We agree that a variety of correct answers to fit the format of the blanks could result. All reasonable answers were accepted. No change required. (7) Question 2.18 - We disagree with the comment and agree with the recommendation. Fa cili ty 's recommended additional answers are essentially the same as the answers given in the answer key. Any reasonable answer similar to the answer key was accepted. No change required.

3 (8) Question 2.19 - We disagree with the comment but agree with the recommendation. Knowledge of rod internal. pressure buildup is important operator knowledge due to its impact on fission product poison leakage and structural strength of the individual fuel rods. However, conflicting curves given in lesson plan NCRODP 88.1 and in NAPS FSAR Fig. 4.2-5 could lead to some confusion as to which curve is the correct choice. Therefore, the question was deleted. (9) Question 2.20 - We disagree with the comment. Knowledge of design basis accidents, including assumed mitigating events, is required knowledge. No change required. (10) Question 2.23 - The question states the Decontamination Factor (DF); it does not require the candidat? to 1(ate it. The K/A Catalog, NUREG-1122, lists derineralizer performance and design attributes-including isotopic control in two different areas of required R0 knowledge. Although not asked for in this question, NUREG-1122 also lists DF's as an RO knowledge area; however, it is relatively low in importance. No change required. (11) Question 2.25 - We disagree with the comment and recommendation in part. Design of important components such as the MSRs is necessary knowledge for the reactor operator (Ref: NUREG-1122, System 039). However, we agree that the candidates may have been confused by the wording of the second part of the question. The portion of the question dealing with the effect on the plant's thermal power capabilities was deleted. (12) Question 2.26 - Answers listing correct physical or system location will be accepted. The recommended answers were added to the answer key. (13) Question 2.31(1.c) - Tours of the facility indicated that (2) is the only correct answer. No change required. (14) Question 3.03 - Lesson Plan 91.1 conflicts with TS on this setpoint and should be changed. RCS pressure T2000 psig was accepted. The answer key was changed accordingly. (15) Question 3.06 - We do not concur that information asked for is beyond the scope of an RO's knowledge and abilities. The R0 candidate must be familiar with the conditions that require the use of safety and emergency systems and why such protection is required with emphasis on areas where a malfunction will require immediate operator action. Malfunctions of instruments and controls taken credit for in accident analyses are prime examples. No change is required. 1 Lm.._..__._.__ -.-- . - - - -- - - - - - - - - . - - . _ - - - - - - - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - - - - - - - - - - - - - - ^ - - - - - - - - - - - - - - - ------- - - - - - - - -------

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     . (16) ' Question -3.09 - We consider this a required knowledge area for R0's and SR0's. It is in fact specifically listed in NUREG-1021, ES-202, Paragraph B3; "The candidates should have sufficient knowledge of the nuclear instruments       ... to answer questions concerning   ...  (the) ' diagrammatic representation of . instrumen-tation' systems."

It is also listed at a relatively high importance rating in NUREG-1122, System 015. No change to question or answer key. (17) Question 3.16 - We agree that the first part of the required answer is more important operator knowledge. However, the second part of the answer does not require " knowledge of maintenance tasks performed by instrument technicians". It merely requires the operator to know that some tasks performed by the instrument tech are made more convenient - not what tN tasks are or how they are performed. The answers were weighted (.75) for the first part and (.25) for the second part. (18) Question 3.18 - During staff review, a typographical error was - discovered in the answer to part (a). The correct answer is 7. The answer key was changed accordingly. (19) Question 3.22 & 3.23 - It was discovered during the examination that these questions were testing the same candidate knowledge. Since question 3.23 was in the preferred wording, it was retained and a question substituted for 3.22. The candidates were informed and the master question and answer key have been changed accordingly. (20) Question -3.27 - We agree. The FSAR reference supplied supports the comment and the recommendation is an appropriate resolution. The answer key was changed to accept either (b) or (c).- (21) Question 4.12 - We disagree. Recognition of a correct answer from among incorrect distractors does not require ' memorization of the correct answer. In addition, candidates are responsible for knowledge of the contents of normal, off-normal, and emergency procedures, regardless of whether the procedure contains immediate action steps. These steps which are not immediate actions need not be memorize.d, but the candidate must' be able 'to describe, conceptually, the objectives and methods used to achieve those objectives for all emergency and off-normal procedures. No change required. (22) Question 4.14 - We disagree. This question did not require verbatim recitation of the five items. This is an - example of asking the candidate to describe, conceptually, the five methods used 'to achieve the objective of ensuring ~ that certain Tech Spec requirements are met.

5 During staff post examination review, two additional answers were judged to be correc.t. These were added to the answer sheet. No other-changes were required. (23) Question 4.23 - It was discovered by the staff during onsite post examination review that a fifth answer was correct. This was a result of the utility supplied material being outdated. The answer key was changed accordingly. , (24) Question 4.38 - We agree with the comment and the utility recommendation. Any answer from the RWP was accepted,

b. SRO Exam (1) Question 6.02 - Utility comment and NRC response same as Question 2.06.

(2) Question 6.06 - Utility comment and NRC response same as Question 3.27. (3) Question 6.10 - Utility comment and NRC response same as Question 2.19. (4) Question 6.19(c) - Utility comment and NRC response same as Question 2.31(c). (5) Question 6.24 - Utility comment and NRC response same as Question 2.25. (6) Question 6.28 - Knowledge of Fire Protection Systems (FPS) including fire, smoke and heat detectors and ability to monitor automatic operation of the FPS, including . actuation of fire , detectors is listed in NUREG-1122. Both areas are rated suf-ficiently high in Importance to Safety particularly for SRO's. Actuation setpoints must be known in order to properly monitor automatic operation. In this instance, the staff judged that 10 F was an acceptable tolerance range for knowledge of setpoint actuation. The answer key was' changed accordingly.  !

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(7) Question 7.02 - Utility comment and NRC response same as Question 4.12. (8) Question 7.30 - The Purpose of AP-52, " Loss of Refueling Cavity Level During Refueling" is the following: This procedure provides the indication of, probable cause for, and the immediate.and long term action to be .taken in the event of a' Rapid Decrease in Refueling Cavity Level. , i

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6 The first response in this procedure is to "Immediately commence make-up to the refueling cavity." Question 7.30 asked for this immediate makeup lineup and the answer key stated simply:

                    "From RWST through LPSH into hot legs."

We believe the question and answer are entirely appropriate since:

1. -Rapid loss of refueling cavity level is a distinct possibility.

2. North Anna has an aporoved procedure for this condition in which an immediate response is indicated.

3. Long term action memorization, Tech Spec operability require-ments and detailed valve lineups were not required in the answer.

No change to question or answer key, 4 2 (9) Question 7.34 - Utility comment and NRC response same as Question 4.23. (10) Question 7.38 - Utility comment and NRC response same as Question 4.38. (11) Question 7.39 - Utility comment and NRC response same as Question 4.14. (12) Question 8.01 - We agree with the latter part of the comment and with the recommendation. Based upon disc'ussions with facility senior operations personnel, we have determined that a- conflict exists between NA Administrative Instruction 16.5, which lists several functions of the SS when performing an Emergency Work Request and the facility's current policy of. having others handle these administrative requirements. The facility Administrative Instruction 16.5 will be changed to reflect the current practice. The question was deleted.

(13) Question 8.05 - We agree with the comment and recommendation. The question was deleted.

(14) Questict 2.07 - We agree in part. Raview of the sdaject amend-ments reveal that only distractor.(c)~is affected. Both answers (b)-

and (c) were accepted, i

;       (15) Question 8.33(b) - We agree with the comment and recommendation.

Updating of training material should be accomplished as soon as possible. Both answers were accepted. (16) Question 8.35 - We agree with the comment and recommendation. Updating of training material should be accomplished as soon as possible. Both answers were accepted.

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4. Exit Meeting At the conclusion of the site visit, the examiners met with representatives of the plant staff to discuss the results of the examination. Those individuals who clearly passed the oral and simulator examinations were identified.

There were no generic weaknesses (greater than 75 percent of candidates giving incorrect answers to one examination topic) noted during the oral examination. The cooperation given to the examiners and the effort to ensure an atmos-phere in the control room conducive to oral examinations were also noted and appreciated. The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners. 6 m- - ,--r ,r-.- - -

ENCLOSURE 3 U. 5. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: NORTH ANNA 1&2 REACTOR TYPE: PWR-WEC3 DATE ADMINISTERED: 85/10/15 EXAMINER: TOM ROGERS i APPLICANT:~ __ _h)_[_kh_____ _ j/ INSTRUCTIONS TO APPLICANT: 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 s the examination starts.

                                                       % OF CATEGORY            % OF    APPLICANT'S      CATEGORY VALUE          TOTAL         SCORE         VALUE                         CATEGORY I____     _    I__   _.._________     ________

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 3pj

        $9799            '5.00                                      PLANT DESIGN INCLUDING SAFETY

________ __'____ ___________ ________ 2. AND EMERGENCY SYSTEMS 40.00 25.00 INSTRUMENTS AND CONTROLS

      ._______ ._____            ___________      ________ 3.

40.00 '5.00 PROCEDURES - NORMAL, ABNORMAL, ________ __~.___ ___________ ________ 4. EMERGENCY AND RADIOLOGICAL ' CONTROL jggg bda >b 100.00 TOTALS FINAL GRADE _________________% . All work done on this examination is my own. I have neither given nor received aid. IEELiCIUY~5~555UdiUR5~~~~~~~~~~~~~~

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1. PRINCIPLES OF NUCLEAR POWER PLANT GPERATION, FAGE 2 THERMODYNAMICS, -- RANSFER HEAT T AND FLUID FLOW t QUESTION 1.01 (1.00)

An ECP is es1culated for a reactor startup 4 hours'after a reactor trip

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from 100% equilibrium conditions. Which of the following condi. ions would cause the actual critical position to be lower than the ECP?

a. The startup is delayed until 8 hours after the trip.

b. Actual boron concentration is 10 ppm more tha,n the predicted boron concentration.

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c. A rod finger is. separated from its spider assembly.

d. The steam dump pressure setpoint is lowered by 100 j psi prior to reactor startup.

QUESTION 1.02 (1.00) Which of the following will cause the A::ial Flux Dif ference to become more positive (less negative)?

a. Power increase with power defect compensated for by dilution only.

b. Power increase with power defect compensated for by rod withdrawal only.

c. Buildup of :enon in top portion of core.

d. Burnup of xenon in bottom portion of core.

QUESTICN 1.03 (1.00) Which of the following will cause the Moderator Temperature Coefficient to become more negative? i a. Increasing boron concentration.

b. Insertion of control rods.

c. Decreasing moderator temperature.

d. Flux moving towards center of core.

(**xas CATEGORY 01 CONTINUED ON NEXT PAGE ****a) 4

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION'

    ~~~~iE5EE6655 E555,~E5 i~iR E555R IEU~ELU56~ELUE QUESTION      1.04               (1.00)

Which of the following witl cause the Fuel Temperature Coefficient (pcm per de3ree) to become less negative?

a. Fuel temperature increase.

b. Baron concentration decrease.

c. Control rod insertion (at const. ant power). ,

d. Core age increase.

o GUESTION 1.05 (2.00) If steam goes throu3h a throttling processe indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME.

a. Enthalpy (0.5)

Pressure (0.5) b. Entropy (0.5)

  . c.

Temperature (0.5) d. (***** CATEGORY 01 CONTINUED ON HEXT PAGE x**xx) J t 1 1

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PAGE 4

1. PRINCIPLES OF NUCLEAR POWER PLANT OPEPATION,

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 ~~~~iEEEE66iEIEiC5,~EE T Yk E5E5R~d 6~5LUi6"iL6E GUESTION       1.06         (2.00)

Answer TRUE or FALSE to the following. -

e. The use of a ramped Tave program allows the plant to operata with a higher thermodynamic efficiency than does a constant (0.5)

Tave program.

b. Increasing condensate depression (subcooling) will cause BOTH a decrease in plant efficiency'AND an increas'e in condensate (hotwell) pump available NPSH. (0.5)

c. During a RCS heatup, as temperature gets higher, it will take a smaller letdown flow rate to maintain a constant pressuri:er (0.5) level.

d. The diffeience between pump suction pressure and the saturation pressure of the fluid being pumped is referred to as net post- (0.5) tive suction head.

GUESTION 1.07 (1.00) State whether you AGREE or DISAGREE with the following statements.

a. Starting with a reactor that is critical low in the intermedi-ate range, an equal amount of positive or negative reactivity (50 pen) insertion will produce stable SURs of equal magni- (0.5) tude.

b. Whenever NR-45 (SRNI) is drawing a straight vertical line, the reactor is critical. (0.5)

) (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE *****) J i . J

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1. PRINCIPLES OF NUCLEAR POWER FLANT OPERATION, PAGE 5 THERMODYNAMIC -_S, HEAT TRANSFER AND FLUID FLOW QUESTION 1.08 (1 00)

The -1/3 DPM SUR following a reactor trip is caused by which of the following?

a. The decay constant of the longest-lived group of delayed ,

neutrons. 3 b. The ability of U-235 to fission with source n,eutrons.

c. The amount of negative reactivity added on a trip being greater than the Shutdown Margin.

d. The doppler effect adding positive reactivity due to the l

temperature decrease following a trip. 1 ] QUESTION 1.09 (1.00) ] The reactor is producing 100% rated thermal power at a core delta T of 60 degrees and a mass flow rate of 100% when a blackout occurs. Natu-ral circulation is established and core delta T goes to 40 degrees. If decay heat is 2%, what is the core mass flow rate (in %)? I

a. 1.3

b. 2.0

c. 3.0 i

d. 4.0 I

(***** CATEGORY 01 CONTINUED ON NEXT FAGE asass) e l

lo PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6

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 ~~~~iA5EE66I5355C5,~55IT iE3E5555 356 fEU16~fL65 QUESTION        1.10                 (1.00)

Which of the following statements concerning Xenon-135 production ind removal is correct?

a. At full power, equilibrium conditions, about half of' the xenon is produced by iodine decay and the other half is produced as a direct fission product.

b. Following a reactor trip from equilibrium con'ditionse

cenon peaks because delayed neutron precursors continue to decay to xenon while neutron absorption (burnout) has ceased.

c. Xenon production and removal increases linearly as power level increases; i.e., the value of 100% equilibrium xenon is twice that of 50% equilibrium xenon.

d. At low power levels, xenon decay is the major removal method. At high power levels, burnout is the major removal method.

QUESTION 1.11 (1.00) Which of the following is the units of heat flur?

a. Watts / cubic centimeter

b. BTU / (hr square ft)

c. C a l o r i e s / .3 r a m

d. kW / ft

(***** CATEGORY 01 CONTINUED ON NEXT PAGE unass) m e'

1

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 7

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  ~~~~iE5EEd6iEEE5C5~~E5Ei~iEEE555R EE6"ELU56~ FLUE OUESTION    1.12            (2.00)

A motor driven centrifugal pump is operating at rated flow. You then start closing down on the discharge valve. How (INCREASE, DECREASE, or REMAIN THE SAME) will each of the following be affected?

a. Flow (0.5)

b. Discharge Presure (0.5)

c. Available NpSH (0.5)

d. Motor Amps (0.5)

GUESTION 1.13 (1.00) The main condenser must remove more heat energy to condense which of the following?

a. One pound of steam at 0 psia.

b. One pound of steam at 300 psia.

c. Two pounds of steam at 400 psia.

d. Two pounds of steam at 1200 psia.

QUESTION 1.14 (1.00) Which of the following describes the changes to *he water that occur between the inlet and outlet of a steam generator?

a. Enthalpy DECREASES, Entropy DECREASES, Quality DECREASES

b. Enthalpy INCREASES, Entropy INCREASES, Quality INCREASES

c. Enthalpy CONSTANT, Entropy DECREASES, Ovality INCREASES

d. Enthalpy DECREASES, Entropy INCREASES. Quality DECREASES

(**xxx CATEGORY 01 CONTINUED ON NEXT PAGE as***) o

1. PRINCIPLE 5 0F NUCLEAR POWER PLANT OPERATI0rl, PAGE 8

 ~~~~iE55366IE3ElC5,~EE3i'_iE355E55~356'iL616 fL6U QUESTION     1 15          (1 00)

The reactor is critical at 10,000 cps when a S/G PORV fails open. Assuming BOL conditions, no rod motion, and no reactor trip. choose the answer below that best descri.bes the values of Tavs and nuclear power for the resviting new steady state. (POAH = point of adding heat).

a. Final Tavs greater than initial Tavg, Final power above POAH.

b. Final Tavs greater than initial Tavg, Final power at POAH.

c. Final Tavs less than initial Tavs, Final power at POAH.

d. Final Tavs less than initial Tavs, Final power above POAH.

QUESTION 1.16 (1.00) Which of the following statements concerning the use of water as the moderator is correct? .

a. Water has a HIGH scattering cross-section, a LOW absorption cross-section, and a LARGE energy decrement per collision.

b. Water has a LOW scattering cross-section, a HIGH absorption cross-section, and a LARGE energy decrement per collision.

c. Water has a HIGH scattering cross-section, a LOW absorption cross-section, and a SMALL energy decrement por collisicn.

d. Water has a LOW scattering cross-section, a HIGH absorption cross section, and a SMALL energy oecrement per collision.

(****m CATEGORY 01 CONTINUED ON NEXT PAGE ammmm),

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 9

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_______________ i E I E l C 5 5 '_ 5 5 I T i 5 A 5 5 5 5 5 ~ I E 6 ' E L U 1 6 '_ f 4 1 l QUESTION 1 17 (1.00) Which of the following statements describes the relationship between . , integral and differential rod worth?

!    a. Integral rod worth (at any location) is the slope of the differential rod worth curve at that iocation.

, b. Integral rod worth (at any location) is the total area under the differential rod worth curve from t'he end of the rod to that location. 5 c. Integral rod worth (at any location) is the square of the differential rod worth at that location.

d. There is no relationship between integral and differential i rod worth.

j QUESTION 1.18 (1.00) The reactor is operating at 30% power when one RCP trips. Assuming no reactor trip or turbine load change occur, which of the following parameters will DECREASE?

a. Flow in operating reactor coolant loops

b. Core delta T

c. Reactor vessel delta P

d. Steam generator pressure in affected loop QUESTION 1.19 (1.00)

Which of the following would cause an INCREASE in the conductive heat transfer rate across a slab?

3. Decrease heat transfer area of slab.

, b. Decrease delta T across slab. i i

c. Decrease thickness of slab.

! d, Decrease thermal conductivity of slab. (susus, CATEGORY 01 CONTINUED ON NEXT PAGE *****) i

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t 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 10

  ~~~~iA5E55675%51C5~~55%i~iR355555~356 5EUl6~5L65 2_

QUESTION 1 20 (1.50) Indicate whether the following statements concerning delayed neutrons are TRUE or FALSE.

a. If the reactor is supercritical, then the fraction of delayed neutrons shifts to the shorter lived precursors and the value of the effective decay constant (lambda) decreases. (0.5)

b. Due to the significant decrease in the percen'tage of fast fission occuring over core life, the value of the effective delayed neutron fraction decreases over core life. (0.5)

c. Delayed neutrons are produced at some time after fission as a result of the radioactive decay of fission products. (0.5)

QUESTION 1.21 (1.00) With the reactor initially at a keff of 0.99, a certain reactivity change causes the count rate to double. If this same amount of reactivity is again added to the reactor, which of the following will be the status of the reactor?

a. Suberitical

b. Critical

c. Supercritical

d. Prompt Critical GUESTION 1.22 (1.00)

A 1/h curve that predicts criticality early is referred to as which of the following?

a. Useless

b. ' Conservative

c. Non-conservative

d. Ideal (smuss. CATEGORY 01 CONTINUED ON NEXT FAGE asssa) i 9

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4 f PAGE 11

1. PRINCIPLE 5 0F NUCLEAR, POWER PLANT OPERATION,

  ~~~~iEEEE665E 55C5I~EENY~ikEE5h k~ N6~ELU 6~5L6U QUESTION                             1.23                                       (1.00)

Which of the following can be defined as 'the number of neutrons causing fission that were originally born delayed divided by the total number of neutrons causing fission'?

a. Lambda effective

b. Rho ,

c. Beta effective

d. Tau

QUESTION 1.24 (1.00)

I Which of the following is NOT a reason for pressuri:ing the fuel rods with helium?

a. Minimi:e clad creeping inwards toward fuel pellets.

b. Increase gap (pellet to clad) thermal conductivity.

c. Allow detection of clad failure by helium analysis of the coolant.

d. Maintain lower fuel centerline temperature.

1 QUESTION 1.25 ( .50) TRUE or FALSE? For similar heat exchangers operating under the same inlet temperatures and flow rates, a counterflow heat exchanger will transfer more heat'than a parallet flow heat exchanger. (** mas CATEGORY 01 CONTINUED ON NEXT PAGE summa) e 1 h

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i PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 12 1.

    ~~~~iE55566ish51C5I~55Ai~iE355EEE~356"FEUl6~EL65 QUESTION     1.26            (1.00)

Which of the following statements concerning samarium reactivity offects is correct? I

a. The equilibrium (at power,) value of samarium is dependent upon power level. The peak value of samarium following a shutdown is dependent upon power level prior to shutdown.

b. The equilibrium (at power) valve of samarium *is dependent upon power level. The peak value of samarium following a shutdown is independent of power level prior to shutdown.

c. The equilibrium (at power) value of samarium is independent of power level. The peak value of samarium following a shutdown is dep'endent upon power level prior to shutdown.

d. The equilibrium (at power) value of samarium is independent of power level. The peak value of samarium following a shutdown is independent of power level prior to shutdown.

GUESTION 1.27 (1.00) Reactivity is defined as which of the following?

a. The ratio of the number of neutrons at some point in this generation to the number of neutrons at the same point in the previous generation.

2 b. The fractional change in neutron population rer generation. , c. The factor by which neutron population changes per genera-tion.

d. The rate of change of reactor power in neutrons per second.

(***** CATEGORf 01 CONTINUED ON NEXT PAGE sassa)'

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 13

_________________S, THERMODYNAMIC HEAT TRANSFER AND FLUID FLOW e QUESTION 1.28 (1.00) The following are the boiling phases associated with nucleste boiling and departure from nucleate boiling. , a 1) Transition Boiling

2) Bulk Boiling

3) Film Boiling ,

4) Sub-cooled Nucleate Boiling .

Which of the following is the order in which they would occur in a channel with normal flow and high heat flux?

a. 2, 4, 3, 1 4

b. 2, 4, 1, 3

c. 4, 2, 3, 1 ,

l

d. 4, 2, 1, 3

. QUESTION 1.29 (1.00)

! Which of the following statements best characterizes Natural Circulation?

a. It needs a pump to get started.

l

b. The elevation of the heat source must be abose that of the heat sink.

c. The driving force is a difference in density.

d. Heat transfer is more efficient i f steam is mixed with water, i

d j i (s**** CATEGORY 01 CONTINUE 0 ON NEXT PAGE asssa) l . t 1 4 4

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 14

         ~~~~iE55556iE555C5~~55Ei~iEE55555~Z56~ FLU 56 fLdU QUESTION       1 30           (1.00)

The reactor startup procedure requires that the critical rod position be taken at 10E-8 amps on the intermediate range. If, during a xenon free reactor startup at h0L, the operator ' overshoot' 10E-0 amps and instead leveled off at 10E-7 amps, which of the following statements is correct?

a. At 10E-7 amps, there are little or no effects,from nuclear heati but, since the reactor is a decade higher in power, the critical rod position would be higher than at 10E-6 amps.

b. At 10E-7 amps, there are little or no effects from nuclear heati therefore, the critical rod position should be the same as at 10E-8 amps.

c. At 10E-7 amps, there are substanial effects from nuclear heati therefore, the critical rod position will be higher than at 10E-0 amps.

I d. At 10E-7 amps, nuclear heati xenon, and the decade higher in power level will result in a higher critical rod posi-tion than at 10E-8 amps. QUESTION 1.31 (1.00) The reactor trips from full power, equilibrium xenon conditions. Six hours later the reactor is brought critical at 10E-9 amps on the inter-mediate range. If power level is maintained at 10E-6 amps which of the following statements concerning rod motion requirements for the next-two hours is correct?

a. Fods will have to be withdrawn since xenon will closely follow its normal build-in rate following a trip.

b. Rods will have to be inserted since xenon will closely follow its normal decay rate following a trip.

c. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of burnout.

d. Pods will have to be rapidly withdrawn since the critical, reactor will cause a higher than normal rate of build-in.

(sus ** CATEGORY 01 CONTINUED ON NEXT PAGE asass) _ , _ . __ -m _ . , _ . . _ . _ _ , . _ , _ _ _ ~ . _ _ , . _ , . _ , _ _ . , . . , _%______--,__,__...,,-_w , . , . , - 1w,-3 ,yw-

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 15

   ~~~~iE5EEd6IEIElC5,~55II~TEIE5F55"EE6~iLUi6~EL U QUESTION    1.32          (1 00)

The Technical Specifications allow operations for a 2-hour time period with a Quadrant Power Tilt Ratio (GPTR) of greater than 1 02. Which of the following is the reason for allowing these operations?

a. To allow time for corrective action in the event of xenon redistribution following power changes.

b. To allow time for correction of a dropped or m'isaligned control rod.

c. To allow time for boron concentration changes to restore the GPTR to less than 1.02.

d. To allow time for correction of RCS flow imbalances.

QUESTION 1.33 (2.00) i Indicate how (INCREASE, DECREASE, or REMAIN THE SAME) an increase in moderator temperature will affect the following parameters.

a. Resonance Escape Frobability (0.5)

b. Thermal Utilization Factor (0.5)

c. Fast Non-Leakage Probability (0.5)

d. Fast Fission Factor (0.5)

GUESTION 1.34 (1.00) Which of the following causes differential boron worth (pcm/ ppm) to. decrease (become less negative)?

a. Boron concentration decreases

b. hoderator temperature decreases

c. Fission product poison buildup

d. RCS pressure increases (xxxxx CATEGORY 01 CONTINUED ON NEXT PAGE **xxx)

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 16

         ~~~~

iAEEE66iEAEIC5I~ESIi _________________________________ iE355__________ FEE ~AA6"_EL0i6 FL6s QUESTION 1.35 (1.00) Which of the following describes the flux shifts that occur to the critical low power flux profile over core life? Assume a cycle 1 core loading.

a. Flux shifts upward and towards the edges of the core.

b. Flux shifts upward and towards the center of,the core.

c. Flux shifts downward and towards the edges of the core.

d. Flux shifts downward and towards the center of the core.

QUESTION 1.36 (1.00) Which of the following is NOT one of the conditions necessary for brittle fracture to occur?

a. Plastic deformation at or below the yield point.

b. Temperature at or below the NDTT.

c. Nominal stress level.

d. Flaw such as a crack present.

(***** END OF CATEGORY 01 auxas) e e -g- - - . -m ...--,----n -cu,-, - - , - . ,-a-, -- - -~ .-o , ,-r -,--.v -,g, , --

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 17 GUESTION 2.01 (1 00)

Which of the following describes the Service Water System automatic , actions on a single unit SI?

a. All SW pumps start, that unit's spray header isolation NOVs receive an 'open' signal

b. Only the SW pumps supplied from that unit's emergency buses start, that unit's spray header isolation MOVs receive an 'open' 2;3nal

c. All SW pumps start, all spray header isolatio'n MOVs rGisi.sg en

       'open' signal.

d. The SW pumps supplied from that unit's emergency buses starte all spray header isolation MOVs receive an 'open' signal.

QUESTION 2.02 (2.50) Match the RCS penetrations in Column A with the appropriate RCS loop segment listed in Column B. (answers may be used more than once) Column A Column B

a. Normal Letdown 1) Loop A cold les

b. P:r Surge Line 2) Loop A hot les C:17 1: 5 :10: 3) Loop A intermediate leg

    ](  c.

d. PZR Spray Line 4) Loop B intermediate les

e. RHR Suction 5) Loop B hot les

6) Loop C cold les

7) Loop C hot leg (xxumm CATEGORY 02 CONTINUED ON NEXT PAGE maamm)

I

2. PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEMS PAGE is GUESTION 2.03 (1.00)

Which of the following statements concerning the operation of the letdown isolation valves (LCV-1460A and B) is correct?

a. Neither of the letdown isolation valves can be opened if the containment letdown isolation valve (TV 1204) is shut.

b. Shutting all orifice isolation valves will automatically shut the letdown isolation valves

c. All orifice isolation valves most be shut in ' order to open the letdown isolation valves.

d. High temperature on the letdown outlet of the regenerative heat exchanger will automatically shut the letdown isolation valves.

QUESTION 2.04 . (1.00) According to 10CFR50.46, which of the following is NOT a design criteria of the Emergency Core Cooling System subsystems.

a. The calculated peak centerline temperature shall not exceed 2000 degrees F.

b. The maximum cladding oxidation shall not exceed 17% of the total cladding thickness.

c. The calculated total amount of hydrogen generated from the cladding reaction with water shall not exceed 1% of the amount that would be generated if all cladding surrounding the fuel reacted.

d. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

4 (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE samma)

2. PLANT DESIGN INCLUDING SAFETY AND ENERGENCY SYSTEMS PAGE 19 GUESTION 2 05 (1.00)

Which of the following does the operator MANUALLY adjust to reduce the RCS temperature when the RHR system is in service for a normal plant cooldowts?

a. Throttle open CCW from RHR Heat E:: changer outlet isolation valve.

b. Throttle open RHR Heat Exchanger outlet isolation valve, c.

Throttle closed RHR Heat Exchanger bypass val,ve.

d. Throttle closed RHR mintflow recirculation valve.

QUESTION 2.06 (1.00) Which statement below regarding pressurizer spray is correct?

a. The position of the auxiliary spray valve can be adjusted from fully shut through the fully open position by the operator at the main control board.

)

b. Each of the normal spray valves has a i spm bypass line to fj prevent cooldown of the spray lines.

c. The maximum spray flow is designed to prevent the PORVs from g; opening on a 10% step decrease in power.

d. Av,<iliary spray utilizes a separate spray no::le due to the large temperature differential that exists when auxiliary spray is used.

GUESTION 2.07 (1.00) Which location below is the discharge point for the pressuri:er head' vent?

s. Containment fuel canal ,

b. Upper region of containment below quench spray rings

c. Pressuricer Relief Tank

d. Svetion side of containment Hydrogen Recombiners.

(xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx). i I 1

20 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY S(STEMS PAGE 20 QUESTION 2.08 (1.50) Indicate whether the following statements regarding RCP seals are TRUE or FALSE. a) The floating ring seal, located between the pump radial bearing and the 41 seal, will limit leakage to 50 spm if the el seal fails. (.5) b) #3 seal is designed to withstand full RCS pressure. (.5) c) Seal water injection from CVCS enters the RCP between the seal package and the pump radial bearing. (.5) GUESTION 2.09 (1.50) The RCS Loop Isolation Valves have two penetrations used for internal pressuri:ation to minimize leakage during maintenance operations. List the two ways in which the valve internals can be pressurized and which one is the preferred method. QUESTION 2.10 (1.50) Indicate whether the following statements regarding the ECCS are TRUE or FALSE. a) The discharge of DNLY UNIT l's LHSI pumps can be lined up to supply or receive water from the discharge of the Outside RS pumps. (.5) b) All three HHSI Pumps will re'ceive a start signal on an SI. (.5) c) The throttle valves located in each of the HHSI Cold Les Injection lines will ensure a rupture in a loop will not cause all the charging flow to spill out that Dreak. (.5) (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE xxxxx)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21 QUESTION 2.11 (1.00)

Which of the following describes the method of NaOH solution addition to the Quench Spray System?

a. An eductor utilizing OS pump discharge draws NaOH solution from the Chemical Addition Tank (CAT) into the QS pump output.

b. Gravity feed from the CAT to the RHST near where the QS pumps take a suction.

c. Gravity feed from the CAT to the area between*the 03 pump inlet isolation valve and the suction side of the pump.

d. The CAT pump discharges the contents of the tank into the QS pump suction with a pre-determined flow rate set by a manual throttle valve. .

QUESTION 2.12 (1.00) Listed below are valves associated with the Recirculation Spray (RS) System. Indicate whether the valves listed are NORMALLY OPEN or CLOSED. a) MOV-SW-102A and B (Service Water supply header X-connects) b) MOV-SW-105A and B (Service Water B return header isolation valves) c) NOV-RS-101A (Casing Cooling Pump A FIRST discharge valve) d) MOV-RS-1558 (Outside RS Pump B suction valve) QUESTION 2.13 (1.00) Which of the following signals, acting independently, will automatically CLOSE the Main Feedwater Regulating Valves (MOV-FW-1154A, B snd C)?

s. Hi-Hi S/G 1 eve 1~ in any S/G (2 out of 3 detectors)

6. Low Tavs (2 out of 3)

c. Reactor Trip

d. Phase B Isolation

(***** CATEGORf 02 CONTINUED ON NEXT P AGE * ****)* I i

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 QUESTION 2.14 (2.00)

Attached is a drawing of the Auxiliary Feedwater System. List the valves downstream of the AFW p'mps u which are closed in a normal Mode 1 lineup. QUESTION 2.15 (1.50) The Fire Protection Water Main is an alternate source of water for the AFW System. List the 3 independent sources of water to this Fire Main System. QUESTION 2.16 (1.00) The steam driven AFW pump is a 6-stage centrifugal pumpe nearly identical to the 6-stage centrifugal motor driven AFW pumps. Yet the steam driven AFW pump is rated at a much higher capacity. Why does this difference exist? QUESTION 2.17 (1.50) Fill in the blanks in the following statements about the Reactor Vessel Head Assembly. a) There are ____ ____ between the upper head flange and the vessel flange that form an _______ and ________ _____. (1.0) b) The center partial length control rod penetration is used for the ______ ______ System. (0.5) QUESTION 2.18 (2.00) List the 4 flow paths within the reactor vessel which BYPASS the fuel rods. (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE **mmm)

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 QUESTION 2.19 (1.00) dbE Which graph on the attached page represents the change in the internal pressure of a fuel rod versus fuel burnup?

a. A

b. B

c. C

d. O

o. E ,

GUESTION 2.20 (1.00) The pressurizer safety valves are designed to limit RCS pressure to within the safety limit of 2735 psis following a complete loss of turbine generator load while at rated thermal power. Which of the following citigating events is assumed to have occurred in the analysis for this design basis?

a. Reactor trip due to turbine loss of load

b. PORV operation

c. Hi PZR pressure trip.

d. Both spray valves actuate.

e. Relief valves on all 3 S/Gs actuate.

QUESTION 2.21 (1.00) Which of the following correctly describes the distribution of J-tubes around the Feedwater Distribution Ring in the S/Gs?

a. They are evenly distributed.

b. There is no pre-determined distribution pattern.

c. More J-tubes are located at the outlet chamber side of the downcomer than the inlet chamber side.

d. More J-tubes are located at the inlet chamber side of the downcomer than the outlet chamber side.

(***** CATEGORY 02 CONTINUED ON NEXT PAGE <*xxx) l x-I

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 QUESTION 2.22 (1.50)

What are the three groups of RCS piping penetrations made BELOW the hori=ontal centerline of the piping? (penetrations with a similar purpose are to be treated as one group) QUESTION 2.23 (2.00) Fill in the blanks in the paragraph belcw regarding CVCS demineralizers. The CVCS contains 5 domineralizers consisting'of _____ mixed bed, _____ cation bed and _____ deborating deminerli ers. The mixed bed's resin will provide a minimum decontamination factor of 10 for all fission products except ________, ________ and _______. The cation demineralizer's purpose is to control the concentration of ________ and the buildup of ________. QUESTION 2 24 (1.50) Indicate whether the follLwing CVCS valves will FAIL OPENr CLOSED or AS IS , on a Loss of Instrument Air. a) Low Pressure Letdown Valve (PCV-1145) b) Charging Flow Control Valve (FCV-1122) c) Letdown Orifice Isolation Valves (HCV-1200A, B and C) GUESTION 2.25 (1.00) Unit 2's Moisture Separator Reheaters (MSRs) have a hexagonal tube bundle which nas almost twice as many tubes as Unit l's MSRs, but the Unit 2 hSRs are wider and shorter than Unit l's. What effect, if any, does this difference in MSR construction have on the Differential Pressure across Unit 2's MSRs =ra +he plantis-thermal power-es p s b i l i t i c ? (Indicate any changes by direction and approximate magnitude.) QUESTION 2.24 (1.00) List the location of all the Radiation Monitors associated with the Service Water System. (If there are monitors on redundant components list as one location) (sxxxx-CATEGORY 02 CONTINpED ON NEXT PAGE **xxx)

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEh5 PAGE 25 QUESTION 2 27 (1.00)

Which of the following is a load using Service Water as a E:ACKUP source of heat removal?

a. Service Air Compressors

b. Instrument Air Compressors

c. Control Room Air Conditioning Units

d. Containment Air Recire Coolers .

QUESTION 2.20 (1.00) Which of the following describes the NORMAL operating range of containment pressure?

c. 1-4 psia

b. 5-9 psia

c. 9-12 psia

d. 13 psia-2 psis

o. 3 psis-6 psis GUESTION 2.29 (1.00)

Which of the following functions.is NOT performed by the hydraulic power system associated with the Containment Personnel Air Lock?

a. Operation of the Equalizing Valve

b. Door Locking Ring Rotation

c. Interior Door Operation

d. Exterior Door Operation (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE ***x*)

8

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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 OUESTION 2.30 (1.50)

Indicate whether the following containment isolation valve combinations are ALLOWED or NOT. c) One LOCKED CLOSED isolation valve INSIDE and a simple check valve used as an AUTOMATIC isolation valve OUTSIDE b) An AUTOMATIC isolation valve INSIDE (not a simple check valve) and a LOCKED CLOSED isolation valve OUTSIDE containment. c) AUTOMATIC isolation valves BOTH INSIDE and OUTSIDE containment with the INSIDE valve a simple check valve. QUESTION 2.31 (1.50) Match the Containment Atmospheric Cleanup System component in Column A with the correct location in Column B. COLUMN A COLUMN B a) H2A-HC-200 H2 Analy:er 1) Aux Bldg near Drumming Area b) H2A-HC-201 Remote Analy:er Panel 2) Below Rod Drive Room c) 1-HC-HC-1 H2 Recombiner 3) Aux Bids near Demin Alley

4) Unit 2 Instrument Rack Room

5) By High Rad Sampling Panels (xxxxx END OF CATEGORY 02 ***ma) l I

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30 INSTRUMENTS AND CONTROLS PAGE 27 GUESTION 3 01 (1.00) Which statement below regarding the pressuri:er pressure control and protective system is NOT correct?

a. The master Pressure controller (NPC) provides the control signal for only one of the PORVs

b. There is a lead /las compensation circuit for pressure inputs to the low pressure reactor trip that varies the trip setpoint with the rate of pressure decrease .

c. The two pressuri er spray valves are controlled by separate transmitters

d. To block SI actuation on a normal plant depressurization, the operator must operate TWO block switches to prevent inadvertant ECCS actuation QUESTION 3.02 (1.00)

PT-402 and 403 measure RCS pressure upstream of RHR suction isolation valves HOV 1700 and 1701. Which of the following statements concerning the effect(s) of PT-402 failing upscale (high) is correct?

a. If both NOV 1700 and MOV 1701 are open they will shut.

b. If open, MOV 1700 will shut and if closed, MOV 1701 will not be able to be opened.

c. If both MOV 1700 and HOU 1701 are shut, they will not be able to be cpened.

d. If open, MOV 1700 will shut, but MOV 1701 can be positioned as desired by the operator.

e. MOV 1700 can be positioned as desired by the operator, but MOV 1701 will not be able to be opened if it is shut.

GUESTION 3.03 (1.00) List the TWO conditions that will provide signals to automatically open the ECCS accumulator discharge valves (1865A, B, and C). (*xxxx CATEGORY 03 CONTINUED ON NEXT PAGE *****) 9 l l

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3. INSTRUMENTS AND CONTROLS PAGE 28 QUESTION 3.04 (1.00) ,

Which of the following is NOT an input into the OT Delta T trip point i caleviator?

a. RCS Flow  !

I

b. RCS Pressure

c. Tavs

d. AFD ,

QUESTION 3.05 (1.00) L With the pressuri er level control selector switch in position I/II, a failure causes the following plant events. (Assume no operator , actions taken.)

1. Charging flow reduced to minimum

2. Pressurizer level decreases

3. Letdown secured and heaters off

4. Level increases until high level trip Which of the following failures occurred?

a. Level channel I failed high

b. Level channel I failed low

c. Level channel II failed high

d. Level channel II failed low ,

OUESTION 3.06 (1.00) Which of the Reactor Trips listed below is taken credit for in accident analysis?

a. Over Power Delta T

b. Pcwer Range High Neutron Flu >:-Low Set Point

c. S. team Flow > Feed Flow, coincident with Low S/G Level  ;

d. Pressurizer High Water Level (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx)*

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INSTRUMENTS AND CONTROLS PAGE 29 30 QUESTION 3.07 (1.00) List the two conditions which necessitate the use of 2/4 reactor trip protection logic vice 2/3 logic. QUESTION 3.08 (2.00) For the following protection / control circuitse indicate whether the listed permissive function occurs as a result of MANUAL or AUTOMATIC Action. a) Bypassing the Power Range ' Neutron Flux-Lowsetpoint above P-10. b) Reenergizing the Source Range instrument below the P-6 setpoint. c) Bypassing S/G 'Lo-Lo Water Level Trip' when RCS Loop Stop Valves are closed. d) Blocking High Steam Flow SI when Tavs < 543 degrees F (P-12). QUESTION 3.09 (2.00) Attached is a block diagram of the Source Range instrument. Indicate on your answer sheet the missing components / outputs labelled A-H on the diagram. QUESTION 3.10 (1.50) Answer the following questions regarding the Excore NIS TRUE or FALSE. c) An Intermediate Range detector that is overcompensated could result in the P-6 setpoint never being reached on a shutdown. b) Placing the Source Range instrument in ' Bypass' at the instrument drawer will prevent a reactor trip signal if either the Instrument or the Control Power Fuses are pulled. c) Neither the Source Range or the Intermediate Range 'Hi Flux' trips are taken credit for in accident analysis. (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) 8

i INSTRUMENTS AND CONTROLS PAGE 30 3. l QUESTION 3.11 (1.00) Which of the following statements describes the two Delta Ts measured on the Core Cooling Monitor when the Loop 1 button is depressed?

a. (Loop A Th-Loop A Tc) and (Highest core thermocouple-Loop A Tc)

b. (Loop A Th-Loop A Tc) and (Highest core thermocouple-loop A Th)

c. (Average core thermocouple-Loop A Th) and (Average core thermocouple-Loop A Tc)

d. (Loop A Th-Highest core thrmocouple) and (Highest core thermocouple-Loop A Tc)

e. (Lonp A Th-Average core thermocouple) and (Loop A Th-Loop A Tc)

GUESTION 3.12 (2.00) Indicate whether the following situations will cause the Steam Dump System to ARM ONLY, ARM AND ACTUATE or have NO EFFECT. (Assume all switches are in normal positions unless otherwise indicated)

   ,   a)   60% power, 3 %/ min load reject, Tavg> Tref by 8 degrees F,    Mode g          Selector Switch in Tavs.

b) Turbine trip from 25% power, Tavg=545 degrees F, Mode Selector Switch in Tavg. c) Hot Zero Power, Tavg=550 degrees F, Mode Selector Switch in STM PRESS, 1020 psig setpoint selected. d) 40% power, 5%/ min load increase, Pimp CH-III fails low, Tavg=560 degrees F, hode Selector Switch in Tavs (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE **xxx) O I i l

3. INSTRUhENTS AND CONTROLS PAGE 31 QUESTION 3.13 (1.00)

Which of the following describes the 3 inputs used in the S/G Bypass Watar Level Control System?

a. S/G Level, Steam Flow, Feed Flow

b. S/G Level, Steam Flow, 1st Stage Turbine Pressure

c. Reactor Power, Feed Flow, 1st Stage Turbine Pressure

d. Reactor Power, Steam Flow, Feed Flow ,

e. Reactor Power, S/G Level, 1st Stage Turbine Pressure GUESTION 3.14 (1.50)

With the reactor at 50% Power, S/G A's steam pressure input to the Steam Generator Water Level Conteol System fails high. Explain how this will affect the Level Control System for S/G A, and why this action occurs. QUESTION 3 .1' 5 (1.00) Which of the following describes the 3 Reactor Vessel Water Level ranges ceasured by RVLIS?

a. Lower, Full, Dynamic

b. Lower, Full, Upper

c. Dynamic, Full, Upper

d. Dynamic, Lower, Upper GUESTION 3.16 (1.00)

Describe the reasons why the D/P cells used for RVLIS are located outside of containment. (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) O s

                                                 ~         - -        -        ,  , , _ -

3. INSTRUMENTS AND CONTROLS PAGE 32 QUESTION 3.17 (1.00)

Which of the following rod withdrawal blocking permissives is ONLY 1 offective when Rod Control is in AUTO? l

a. Intermediate Range Overpower (C-1)

b. Power Range Overpower (C-2) l

c. Overtemperature Delta T (C-3)

d. Turbine Load Interlock (C-5) .

QUESTION 3.18 (2.00) Match the function in Column A with the appropriate component in the Rod Control System listed in Column B. Column A Column B a) Minimizes rod motion at higher power 1) Bank Overlap levels to limit transients 2) Pulse Amplifier

3) Non-Linear Gain Unit b) Indicates to power cabinets how to 4) Pulsar move the rods 5) Haster Cycler

6) Slave Cycler c) Automatically selects proper control 7) Variable Gain Unit bank d) Selects which group within a shutdown bank is to be moved

(***** CATEGORY 03 CONTINUED ON NEXT PAGE **xxx) 0

                         -  _     ,                   ~-        _2      * , , .. _ . . . , . . _ . , _ , .     . . .

3. INSTRUMENTS AND CONTROLS PAGE 33 QUESTION 3.19 (1.00)

Which of the following describes the operation of the search coil associated with Individual Rod Position Indication (IRPI)?

a. Located about i f;ot from the Detector Bottom position, it allows the operator to check rods on the bottom.

b. Located about 6 inches from the Detector Bottom position, it provides input to the Rod Bottom Light circuitry.

c. It runs the length of the rod housing and has'a variable tap that allows the operator to appro:<imate rod position when IRPI is in doubt.

4 d. No such coil is as,sociated with IRPI. QUESTION 3.20 (1.00) Which statement below is correct regarding implementation of the Reduced Pressure Operations Protective System as the plant cools down?

a. PORV reduced setpoints are automatically inserted when Te<320 degrees F and a second pair of PORV setpoints must be manually

          ' keyed' in when Te<140 degrees F.

b. PORV reduced setpoints are automatically inserted when Tc<320 i

degrees F and also when Tc<140 degrees F.

c. PORV reduced setpoints must be manually ' keyed' in when Te<320 .

degrees F and again when Te<140 degrees F. i d. PORV reduced setpoints are manually ' keyed' in when Tes320 degrees F and are automatically inserted when Tc<140 degrees F. GUESTION 3.21 (1.00) 3 tate.the reason for having backup heaters automatically energi:e when actual pressuri er level - reference level by 5% or more. QUESTION 3.22 (g,$oy Indicate whether the following controls for 1H EDG are transferred to/from 4 the 1H diesel room using the

• Control Room Emergency' (CRE) switch or the

   ' Diesel Control Isolation * (OCI) switch.

a) Control and synchroni:ation for the 4.16 KV Emergency Bus emergency supply breaker. ' b) Alternate governor controls by means of.the Droop /Isochronous Selector i switch. c) Diesel Start /Stop control e

                           ,             ,        , . - - - - - -, n .,..:,,9 -.-
                                                                                          ..-.---,,,r.       - - . - ,

3. INSTRUMENTS AND CONTROLS PAGE 34 QUESTION 3.23 (1.50)

Describe the improvements to the flexibility and reliability of the electrical distrubution system after a Unit 1 trip due to the installation of the Unit 1 main generator breaker G12. QUESTION 3.24 (2.00) Answer the following questions regarding a DEGRADED voltage condition on an Emergency 4160 VAC Bus. . a) A degraded voltage condition occurs when ____ out of ____ relays detect <_____% nominal voltage and initiates a ____ timer. (Fill in the blanks) (.75) b)* Describe the sequence of events from the timing out of the timer until all required loads are reenergized on the Emergency Bus. (Be sure to reference any loads NOT shed) (1.25) QUESTION 3.25 (1.00) Which of the following conditions is NOT required for automatic swapover

 . of the LHSI pumps to the Recirculation hude following an SI?

a. RWST Lo-Lo Level

b. A LHSI pump recire isolation n0V closed for each pump L. SI signal present

d. SI Recire Mode signal present GUESTION 3.26 (1.00)

Following a Guench Spray system actuation, what 2 conditions must be met in order to shut the suction valve for a Guench Spray pump? J (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE **xxx) r 0 we

INSTRUMENTS AND CONTROLS PAGE 35 30 QUESTION 3.27 (1.00) Which of the following describes the purpose of the time delay (195 see) in starting the Inside Recirculation Spray (IRS) pumps on a CDA signal?

a. Prevents overloading of the emergency buses as loads are sequentially energized during accident conditions.

b. Allows time for the containment sumps to collect sufficient fluid to provide NPSH for the pumps.

c. Enhances core cooling by increasing reflood c' ate after a LOCA as the pressure drop between core exit and the break is reduced with a higher containment pressure.

d. Allows time for the fluid collecting in the containment sumps to cool, to avoid flashing in the RS heat exchangers as the fluid is cooled by Service Water.

QUESTION 3.28 (1.00) Which of the following describes the normal path of power to Unit l's Charging Pump 1C?

a. RSS A--4160 VAC Xfer Bus D--4160 VAC Emerg Bus J

b. RSS C--4160 VAC Xfer Be- F--4160 VAC Emers Bus J

c. RSS A--4160 VAC Xfer Bus D--4160 VAC Emerg Bus H

d. RSS C--4160 VAC Xfer Bus F--4160 VAC Emers Bus H

e. RSS A--4160 VAC Xfer Bus F--4160 VAC Emerg Bus H (xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) 9

                                                -w-

3. INSTRUMENTS AND CONTROLS PAGE 36 QUESTION 3.29 (2.00)

Indicate whether the following statements regarding Containment Isolation are TRUE of FALSE. a) The Hi-Hi containment pressure input to the CDA logic matrix must ENERGIZE relays to initiate a CDA signal. b) Phase B isolation can be manually initiated by turning one of two switches on the back of the console in the control room. c) If Phase A is reset following s'ignal initiation, without having the initiating signal disappear, than the phase A isolation signal will reoccur when the reset buttons are released. d) S/G Blowdown is isolated by both Phase A and Phase B isolation signals. QUESTION 3.30 (1.00) List all the inputs to the electrical starting circuit that must be natisfied in order to start a Reactor Coolant Pump. QUESTION 3.31 (1.00) Which of the following conditions will NOT result in 'C' Charging Pump receiving a ' Start-Permissive Trip' signal when it is powered from the H bus? (all breakers listed below are open unless stated as closed)

a. Breaker 15J7 racked in

b. Breaker 15H6 racked in

c. Ereaker 15J6 racked out

d. Breaker 15J7 closed

(***** CATEGORY 03 CONTINUED ON NEXT PAGE ammax) e e e

            ~ - - - *                -    -

A - ,, . ~ , , , , _

3. INSTRUMENTS AND CONTROLS PAGE 37 QUESTION 3 32 (1 00)

Which type of sensor belcw is used to indicate when pressurizer spray flow is too low?

a. Temperature

b. Pressure

c. Flow

                'd.         Differential Pressure                                                                       .

3 (mmmmx END OF CATEGORY 03 **xam) e 0 a e i . . t _ , , ', , . , . t

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 38

                       ~        ~~~~~~~~~~~~~~~~~~~~

~~~~Ed656[655EIL EU TE6L QUESTION 4.01 (1.00) Given the following procedural steps, select the choice that provides the proper sequence for paralleling a diesel generator to a live bus.

1. Turn the synchroni:er switch on.

2. Start the emergency diesel generator.

3. Turn off the synchronizer switch. -

4. Operate the governor control switch so the synchroscope moves slowly in the fast direction.

5. Close the generator output breaker.

6. Wait for the synchroscope needle to pass the 11 o' clock position.

7. Operate the exciter control switch to match incoming voltage to running voltage.

a. 1,2,4,6,5,7,3.

b. 2,1,7,4,6,5,3. -

c. 2,1,4,6,5,3,7. _,

d. 1,2,7,6,5,4,3.

QUESTION 4.02 (1.00) When placing the RCS in a solid water condition per OP-3.4, the criterion used to verify the RCS is solid is

a. a high pressure spike occuring on the pressurizer pressure instruments.

b. when spray-i'ng the pressuri er no longer will decrease RCS pressure.

c. when indication of flow to the PRT through the pressurizer PORVs is verified. _-

                                                              ~~~

d. when an increase in letdown flow has been verified.

                                            ~~

d a a . (x**** CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) e

I

4. PROCEDURES - NORMAL, AENORMAL, EMERGENCY AND PAGE 39

   ~~~~                     ~        '~~~~~~~~~~~~~~~~~~~~~~~

rad [dLEGfdAL ddNTRdL QUESTION 4.03 (1.00) Which of the following is the preferred method of providing a means of decay heat removal during a refueling outage with the reactor vessel head removed if all RHR pumps are lost?

o. Using natural circulation by feeding the SGs with AFW and bleeding them using SG blowdown.

b. By lining up the LHSI pumps to r'ecirculate flo*w through the RWST and the reactor vessel.

c. By lining up the refueling purification system to recirculate water from the reactor cavity to the spent fuel pit using

          .the spent fuel coolers for cooldown.

d. By feeding the RCS using a LHSI pump and bleeding it from a SG primary manwa9 QUESTION 4.04 (1.00)

Prior to transferring rod control from MANUAL to AUTO during a reactor startup, Tave and Tref should be within

a. 1 degree F.

b. 3.5 degrees F.

c. 5 degrees F.

d. 10 cestees F.

QUESTION 4.05 (1.00) List four of the critical conditions required to be recorded.during a startup when 1 X E-8 amps is attained. QUESTION 4.06 ( .50) For a reactor startup, it is required by the procedure to go critical while the power indication is still in the source range. TRUE or FALSE) (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE *summ) e i

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 40

~~~~E3656[5555 E~50ETRUL'~~~~~~~~~~~~~~~~~~~~~~~ GUESTION 4.07 (1.00) If criticality is not acheived before the upper limit of the ECP is reached during a reactor startup, the operator should

c. insert Bank D control rods to the bottom.

b. stop control bank withdrawal and approach criticality via boron dilution. '

c. proceed with the startup utilirins an Inverse Count Rate Plot from the time the ECP upper limit is reached until the time criticality is reached.

d. proceed with the startup utili:ing a pull-and-wait approach to criticality.

QUESTION 4.08 (1 00) Control rod cooling air is required to be operating

a. only if the control rod mechanism is energi=ed when the plant temperature is above 350 F.

b. whenever the plant temperature is above 350 F, regardless of the rod mechanism status.

c. during any plant cooldown evolution below 350 F.

d. whenever the individual rod position indication system or any rod drive mechanism is energized.

(xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) 9 i .

1 1

4. PROCEDURES - NORhAL, ABNORhAL, EMERGENCY AND PAGE 41

 --- RA5i5E55153E        55EiR5E------------------------                          ;
       --------------------                                                        l l

GUESTION 4.09 (1.00) The boron concentration is equali:ed between the pressurizer and the rest of the reactor coolant system by

a. char 3i ng pure water or borated water to the PZR through the aux.

spray line as required.

b. charging a blended flow equal to the Reactor Coolant System baron concentration into the Reactor Coolant System'to force more of the RCS inventory up the surge line and into the PZR.

c. taking MANUAL control of the Pressuri er Level Control System and .

then increasing and decreasing the pressuri:er level as required in order to get an exchange between the RCS and PZR water inventory.

d. operation to the pressurizer heaters and spray valves as requ. ired.

QUESTION 4.10 (1.00) The administrative temperature limit between the pressuri:er and spray delta T is

a. 150 F.

b. 200 F.

c. 260 F.

d. 320 F.

QUESTION 4.11 (1.00) Which of the following is NOT required operator action upon loss of power range enannel N-41 during Mode 1 operations?

4. Place the test switch for the OT delta T to the TEST position for the failed channel,

b. Place or verify the control rods are in nANUAL contro17

c. Place the test switch for the OP delta T to the TEST position for the failed channel.

d. Remove the ' Control Power' fuses for N-41 at the power range drawers.

(***** CATEGORY 04 CONTItfuED ON NEXT PAGE *****)

4. PROCEDURES - NORhAL, ABNORMAL, EMERGENCY AND PAGE 42

~~~~EA655L5555AL C5NTRUL'~~~~~~~~~~~~~~~~~~~~~~~ OUESTION 4.12 (1.00) Which of the following is an INAPPROPRIATE operator action if, during refueling, the containment gas and particulate radiation monitors alarm?

a. Stop the containment purge supply and exhaust fans.

b. Close the purse and exhaust fan NOVs. ,

c. Stop the containment air recirculation fans.

d. Start the containment iodine filtration fans.

QUESTION 4.13 (1.00) The trip bistables of a failed power range detector are placed in the trip condition by

a. placing the applicable bistable test switch in the ' Test' position in the Reactor Protection Cabinet.

b. removing the applicable control and instrument power fuses on the power range drawers.

c. placing the applicable Power Mismatch Bypass switch to the failed position at the hiscellaneous Control and Indication Panel.

d. placing the applicable Comparator Channel Defeat switch to tne failed channel position at the Detector Current Comparator Panel.

QUESTION 4.14 (1.25) List five surveillance items that must be done by operators to ensure Tech Spec requirements are met in the event of a loss of both units PRCCAC-250 Computers. QUESTION 4.15 ( .50) A rod control logic cabinet internal failure can be verified in the control room. TRUE or FALSE? , (***** CATEGORY 04 CONTIAUED ON NEXT PAGE *****)

40 PROCEDURES - NORMAL, ABNORhAL, EMERGENCY AND PAGE 43

                       -              ~~~~~~~~~~~~~~~~~~~~~~~~

~~~~EE65dL655 DEL EUNTR6L QUESTION 4.16 (1.00) If instrument air is lost outside the containment and the plant is solid, what is your required immediate actions? QUESTION 4.17 (1.00) What is the required operator actions if both source range instruments fail when >P-6 and <P-10 during a power increase 'to 100% power? QUESTION 4.18 (1.00) List the immediate operator actions for one dropped rod at 75% power. QUESTION 4.19 (1.00) List the immediate operator actions for a rod insertion limit low-low olarm. QUESTION 4.20 (1.00) List the immediate operator actions for a failure of the controllin3 first stage impulse pressure instrument when at 50% power. QUESTION 4.21 (1.00) List the immediate operator actions for an uncontrollable increase in main condenser pressure with the reactor at 50% power. QUESTION 4.22 (1.00) What operator actions are required upon evacuating the control roon if the reactor could not be tripped before exiting the control room? (***xx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) 3

40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 44

   ~~~~                             ~~~~~~~~~~~~~~~~~~~~~~~~

R 655L5555EE'55NTR5L 00ESTION 4.23 (1.00) List the immediate operator actions for an Anticipated Transient Without Trip. QUESTION 4.24 (3.25) List the immediate operator actions for the Reactor Trip or Safety Injection procedure. OUESTION 4.25 (1.00) Which of the following indications require SI reinitiation following a spurious SI that has been secured?

a. RCS pressure at 1950 psis.

b. RCS subcoolin3 at 35 degrees F.

c. Pressurizer level at 17%.

d. All steam generator levels at 15%.

QUESTION 4.26 (1.00) During a natural circulation cooldown, the preferred method of reactor coolant system depressurination is by

a. opening the normal spray valve.

b. opening the auxiliary spray valve.

c. opening the pressurizer PORVs.

d. opening the reactor vessel vent valves.

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) e - w em+- -

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45 4.

                     ~          ~~~~~~~~~~~~~~~~~~~~~~~~
 ~~~~E3656[6535 L 55ETRUL QUESTION     4.27           (1.00)

A hydrogen bubble formed in the reactor vessel is eliminated by

e. increasing pressuri:er temperature above core thermocouple readings.

injecting oxygen into the reactor coolant system via the I b. chemical and volume control system. ,

c. maximizing coolant flow by running all reactor coolant pumps, increasing letdown flow to 120 spa, and placin3 the cation bed demineralizer in service in parrallel with the mixed bed demineralizer.

d. venting the reactor vessel head.

QUESTION 4.28 (1.00) A fire found in the 4160 switchgear is best extinguished with a fire extinguisher marked with the symbol

a. A

b. B

c. C

d. D GUE5 TION 4.29 (1.00)

The Gai-Tronics channel reserved for emergencies is

3. channel 1.

b. channel 3.

c. channel 5.

d. channel 7.

(*xx** CATEGORY 04 CONTINUED ON NEXT PAGE *****)' s l

                           /

4. PROCEDURES - NORMAL., ABNORMAL, EMERGENCY AND PAGE 46

                        ~
  ~~~~5I616E3GlEIL EENTREE~~~~~~~~~~~~~~~~~~~~~~~~

QUESTION 4.30 (1.00) What should the operator do about an SI signal if it exists during a loss of all AC power?

 -GUESTION     4.31          (1.00)

What two conditions constitutes adverse containme,nt conditions? GUESTION 4.32 ( .50) Will adverse containment conditions cause affected control room indications to indicate higher or lower than actual conditions? GUESTION 4.33 (1.00) Li.st the whole body administrative limits per calender quarter that can be achieved (a) without any additional approval and (b) with the highest level of approval. QUESTION 4.34 (1.00) Which of the following is a 10 CFR 20 exposure limit?

a. 5 ram / year-wnole body.

b. I rem / quarter-whole body.

c. 18.75 rem / quarter-hands.

d. 7 rem / quarter-skin of whole body.

(***** CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx) O

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 47

                           -          '~~~~~~~~~~~~~~~~~~~~~~~
 ~~~~R K5i5t55isiL C5kTR5L GUESTION        4.35           (1.00)

If you are in a 100 mrad / hour samma field for 45 minutes, what is your dose in MREM after 45 minutes?

a. 45

b. 75

c. 450 ,

d. 750 I

QUESTION 4.36 (1.00) Which of the following radiation exposures would inflict the greatest biological damage to man?

a. 1 Ren of GAMMA.

b. 1 Ren of ALPHA.

c. 1 Rea of NEUTRON.

d. HONE of the abovel they are all equivalent.

QUESTION 4.37 (1.00) It is required to exit a radiation area and report to HP whenever your 0-200 mr self-reading dosimeter r.eaches

s. 100 mr.

b. 125 mr.

c. 150 mr.

d. 175 mr.

QUESTION 4.38 (1.00) List five pieces of information an RWP will tell you that you will need to know in order to work in the applicable restricted area. (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE xxxxx)* e

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 48

                                 ~~~~~~~~~~~~~~~~~~~~~~~~
 ~~~~53555L5555IE'555555L GUESTION      4.39           (1.00)                                     .

An RWP number is necessary to obtain

a. a dose card.

b. a TLD.

c. a self-reading dosimeter.

d. your current quarterly dose.

(xxxxx END OF CATEGORY 04 xxxxx) (xxxxxxxxxxxxx END OF EXAMINATION xx***xxxxxxxxxx) we 8 e 8 4 l _ a _-

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 49 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- NORTH ANNA 182 -05/10/15-TOM ROGERS ANSWER 1.01 (1.00) d REFERENCE NUS, Nuclear Energy Training - Reactor Operation VEGP. Training Text, Vol. 9, Ch. 21 DPC, Fundamentals of Nuclear Reactor Engineering 001/010-K5.13 ANSWER 1.02 (1.00) b -

REFERENCE NUS, Nuclear Energy Training - Reactor Operation

 . ANSWER        1.03              (1.00) b REFERENCE NUS, Nuclear Energy Training - Reactor Operation, p. 9.2-5 Westinghouse Reactor Physics, pp. I-5.5, 10, a 11
   ' ANSWER        1.04              (1.00) a REFERENCE NUS, Nuclear Energy Training - Reactor Operation, p. 9.2-5 t                            %
                         .s
                                                                                      ! 1

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 50

      --- inERB55iREBIEi- Ri%i iR%siFER Es5 FEDiB FE5s ANSWERS -- NORTH ANNA 1&2                                                             -85/10/15-TOM ROGERS ANSWER                     1.05              (2.00)

a. REMAIN THE SAME (0.5)

b. DECREASE (0.5)

c. INCREASE (0.5)

d. DECREASE (0.5)

REFERENCE Steam Tables 010/000-K5.02 (2.6/3.0) ANSWER 1.06 .(2.00)

a. TRUE (0.5)

b. TRUE (0.5)

c. FALSE (0.5)

d. TRUE (0.5)

REFERENCE 4 NUS, Nuclear Energy Training - Plant Performance, p. 5-3.2 General Physics, HT & FF, pp. 155, 319, and 320 VEGP, Ovestion Bank, #752-s and 4 336 ANSWER 1.07 (1.00)

a. DISAGREE (0.5)

b. DISAGREE (0.5)

REFERENCE VEGP, Training Text, Vol. 9, pp. 21-25 & 49 and UEGP Ovestion Bank 460 l f e l l J

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 51

~~~~iEEEE33YEEEIC5I~H5di~i5EE5Fli 3EB FLUi6~5L55 ANSWERS -- NORTH ANNA 182 -85/10/15-TOM ROGERS ANSWER 1.08 (1.00) a REFERENCE VEGP, Training Text, Vol. 9, p. 21-47 Westinghouse Reactor Physics, pp. I-3.17 & 19 DPC, Fundamentals of Nuclear Reactor Engineering, p. 106 001/000-K5.49 (2.9/3.4) ANSWER 1.09 (1.00) c REFERENCE General Physics, HT & FF, Section 3.2 WBN, HT & FF, p. 13

-002/000-K5.01     (3.1/3.4)

ANSWER 1.10 (1.00) d REFERENCE Westinghouse Reactor Physics, PP. I-5.63 - 76 HBR, Reactor Theory, Sessions 38 and 39 DPC, Fundamentals of Nuclear Reactor Engineering, Section VI 001/000-K5.39 (3.5/4.1) ANSWER 1.11 (1.00) b REFERENCE General Physics, HT & FF, p. 229 002/000-K5.01 (3.1/3.4)

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 52

                                                            ~
  ~~~~iOERdEE7 5 fC5,~EEdi~ FREE 5EER dEE~ELEf6~ELEE ANSWERS -- NORTH ANNA 1&2                                  -85/10/15-TOM ROGERS 1

ANSWER 1.12 (2.00)

a. DECREASE (0.5)

b. INCREASE . (0.5)

c. INCREASE (0.5)

d. DECREASE (0.5)

REFERENCE General Physics, HT & FF, pp. 319 - 334 ANSWER 1.13 (1.00) c REFERENCE Steam Tables ANSWER 1.14 (1.00) b REFERENCE General Physics, Heat Transfer Thermodynamics and Fluid Flow, pp. 68, 82, a 156 ANSWER 1.15 (1.00) d REFERENCE Westinghouse Reactor Physics, Section I-5, MTC and Power Defect OPC, Fundamentals of Nuclear Reactor Engineering 002/000-K5.02 (3.3/3.6) . J

10 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,

 ------------------------------------------------                                                         PAGE          53    ,

THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS ANSWER 1.16 (1.00) a REFERENCE Westinghouse Reactor Physics, pp. I-2.19 - 21 HBR, Reactor Theory, Session 14, p. 3 DPC, Fundamentals of Nuclear Reactor Engineering, p. 53 001/000-K5.57 (3.0/3.2) ANSWER 1 17 (1.00) b REFERENCE Westinghouse Reactor Physics, p. I-5.40 HBR, Reactor Theory, Session 36, p. 2 ANSWER 1.18 (1.00) c REFERENCE General Physics, HTFF - Fluid Flow Applications for Systems and Components ANSWER 1.19 (1.00) c REFERENCE WBN, HT & FF. p. 19 General Physics, HT&FF, p. 105 1 3

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

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 54 THERMODYNAMICS -----------------------

HEAT TRANSFER AND FLUID FLOW ANSWERS -- NORTH ANNA 182 -85/10/15-TOM ROGERS ANSWER 1.20 (1.50)

a. FALSE (0.5)

b. FALSE (0.5)

c. TRUE (0.5)

REFERENCE Westinghouse Reactor Physics, pp. I-3.9 and I-3.4 HBR, Reactor Theory, Sessions 22 and 23 DPC, Fundamentals of Nuclear Reactor Engineerins ANSWER 1.21 , (1.00) c

\

REFERENCE

;    HBR, Reactor Theory, Session 42, pp. 3&4 DPC, Fundamentals of Nuclear Reactor Engineering 004/000-K5.08            (2.6/3.2)

ANSWER 1.22 (1.00) b REFERENCE NUS, Reactor Theory i

ANSWER 1.23 (1.00) c REFERENCE NUS, Reactor Theory ANSWER 1.24 (1.00)

C 1 1 , +

t 10 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 55 t THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS r i t I REFERENCE

General Physics, HT & FF, pp. 239 and 240 I 4 i d

ANSWER 1.25 ( .5e) f i TRUE [ REFERENCE General Physics, HT & FF, p. 176 i r ! 002/000-K5.01  ! I l ANSWER 1.26 (1.00) f C L REFERENCE [ DPC, Fundamentals of Nuclear Reactor Engineering, p. 170 j 001/000-K5.13 (3.7/4.0) , ANSWER 1.27 (1.00) b REFERENCE DPC, Fundamentals of Nuclear Reactor Engineering, p. 96

001/000-K5.56 (2.8/3.1) l l l ANSWER 1.28 (1.00) -

4 d REFERENCE NUS, Vol 4, pp 3.3-2  ; h f I

• L i

l  ! i F f 1 I

                                                                                                                  ' f
      - - _ - , - -            - - - .          . _ - - , ,   _ . . - . . .- - _ - - ,      . , , . - .   ..          --,_. c_ . -.- - .. - , -

i

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 56

, ~~~~IHEEE66f dEfCE,"_EST fEEEEEEE~dEE~FLEfE~ FLEE ANSWERS -- NORTH ANNA 182 -33/10/15-TOM ROGERS ANSWER 1.29 (1.00) < l c REFERENCE General Physics, HT&FF, pp. 355 - 358 ANSWER 1.30 (1.00) b REFERENCE Westinghouse Reactor Physics, Sects 3 and 5 NUS, Nuclear Energy Training, Units 6 and 8 ANSWER 1.31 (1.00) a ANSWER 1.32 (1.00) b ANSWER 1.33 (2.00)

a. DECREASE (0.5)

b. INCREASE (0.5)

c. DECREASE (0.5)

d. INCREASE (0.5) i REFERENCE Westinghouse Nuclear Training Operations, PP. I-2.31 - 36 f

ANSWER 1.34 (1.00) c 4

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 57

  ~~~~fHEREEEE EEIEE,~EEdf~fEdk5EEE~dEE~ELEEE~ELEE ANSWERS -- NORTH ANNA 1&2                             -85/10/15-TOM ROGERS REFERENCE Westinghouse Nuclear Training Operations,                   p. I-5.31 ANSWER        1.35           (1.00) i   a REFERENCE Westinghouse Nuclear Training Operations,                     pp. I-5.53 and 54 ANSWER        1.36           (1.00) a REFERENCE DPC, FNRE, pp. 200 and 221 I

i i

                                                                                                                    )

i

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 58 ANSWERS -- NORTH ANNA 182 -85/10/15-TOM ROGERS

, ANSWER 2.01 '(1.00) REFERENCE NA NCRODP 92 2, ' Service Water System' ANSWER 2.02 (2.50) M '. 'I a) 1

    -.           Te.k4.

d) 1,6 e) 2 REFERENCE Farley SD, 'RCS', Fi3 7 NA NCRODP, 'RCS'i 'ESF-ECCS'; 'CVCS'; 'RHR' ANSWER 2.03 (1.00) c 4 REFERENCE 4 Farley 50, 'CVCS*, pp 13 NA NCRODP 88.3, ' CVCS' ANSWER 2.04 (1.00:' a REFERENCE 10CFR50.46(b)

;        FNP, SD, 'ECCS', PP 5 NA NCRGDP 91.1,         '

ESF-ECCS' 006/050-PWG435 (4.2/4.3) a 9

5. . M

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 59 ANSWERS -- NORTH ANNA 142 -85/10/15-TGH ROGERS ANSWER 2.05 (1.00) b REFERENCE FNP, SD, 'RHR System', PP 17 NA NCRODP 88.2, 'RHR System'i NA OP 14.1 i PWG K/A 29 (3.6/3.9)

ANSWER 2.06 (1.00)

  -"                                 )c kth -M r,cr arne:

NA NCR00r CO.1, '"CC "ZP :nd * :::. oclief' ANSWER 2.07 (1.00) a REFERENCE NA NCRODP 88.1, 'RCS-PZR and Press Relief

• ANSWER 2.08 (1.50) a) True (+.5 ea) b) False c) False REFERENCE NA NCRODP 88.1, 'RCS-RCP' ANSWER 2 09 (1.50)

1) Portable system connecting to blank flanges.(+.5)

2) Temporary piping connection from accumulators via blank flanges (+.5)

   ----t2 is the preferred method                        (+.5) i e

t t _ - _ , _,e - - , _ . . - --- -,

e _s, - ' "' ( AUXILIARY FEEDWATER SYSTEM 1 ]

                                     . L.                        -
!                                                                                                                       EL2F l                                                                                                                           M.
                                                                                                    #D                  Gi T                   m' Lu:

1 I" PC$N , , a m i pleec - - - to [& l 6 I p pw A ~ ~ gT - w wi --- i g

o. a c

6 y m,o, ' , I n 1 x e 44 / MPm/ " E dx"! e-l p jeg y s s-x f a) o ' m

                                                        . ,                                 -           iv r

03- put hiteuf 3) [ m- - I l g m p m. m_e m. f 6 m m_ .oL Il ""u;; n-ngW . a nu i l 4 I i e i l[ ( t___________.______ _ _ . _ - _ _ _ _ _ _ _ _ _ .

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE' 60 ANSWERS -- NORTH ANNA 182 -

85/10/15-TOM ROGERS

,    REFERENCE i     NA NCRODP 88.1, 'RCS Piping and Instrumentation' ANSWER                     2.10                     (1.50) a) True                  (+.5 ea) b) True c) True ANSWER                     2.11                     (1.00) i     b REFERENCE                                       .                                                                                                                     ,

NA NCRODP 91.1, 'ESF-OSS' I ANSWER 2.12 (1.00) a) Open (+.25 ea) b) Closed

,     c)    Open d)    Open REFERENCE NA NCRODP 91.1, 'ESF-RSS' ANSWER                     2.13                     (1.00)

REFERENCE l NA NCRODP 89.4, 'Feedwater Systems-Main Feed' l I l i i B t

        -     ,. n - , - - - ,       - ~ , , - - ,--.       - . . . - . - - .   ...,.,-...--,,,,,,....--,,---,,c,.,n             --- , ,       ~ ,,,w-,,---,,,-m.--+--

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 61 ANSWERS -- NORTH ANNA 182 -85/10/15-TOM ROGERS ,

l ANSWER 2.14 (2.00) t M0'U-100C (+.25 ea) MOV-100A ' dPCV-100B i fPCV-100A 155 149 166 _ 190 See Attached Drawing l REFERENCE l NA NCRODP 89.4, 'Feedwater Systems-AFW' ANSWER 2.15 (1.50) [ t

    -Service Water Resevoir                      (+.5 ea)                                                                                          i
    -Lake Anna
    -Discharge Canal           kJW-5                                                                                                               f r

REFERENCE

!   NA NCRODP 89.4, 'Feedwater Systems-AFW'                                                                                                        j i

i ANSWER 2.16 (1.00) , l The Steam driven pump is operated at a higher speed. (+1.0)  ; t REFERENCE I NA NCRODP 89.4, 'Feedwater Systems-AFW' ANSWER 2.17 (1.50) r i l c) Hollow 0-rings (+.5)  !

Inner and outer seal (+.5) b) Reactor Vessel Head Vent System (RVHVS) (+.5)

REFERENCE , NA HCRODP 88.1, 'RCS-Reactor Vessi and Core Construction' l I I i l t I __. .. - .-_ _ _ . _ _ _ , , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , a. _ . _ _ _ _ _ . _ . _ _ _ _ _ _

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 62 ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS ANSWER 2.18 (2.00)

1) To upper head plenum via no::les in core barrel flange (.5 ea)

2) Between hot les discharge no::les and upper core barrel outlets

3) Between baffle plates and core barrel

4) Around inserts in guide thimble tubes in the fuel assemblies REFERENCE NA NCRODP 88.1, 'RCS-Reactor Vessel / Core Construction' ANSWER 2.19 (1.00) be$&b 9f REFERENCE NA NCRODP 88.1, 'RCS-Reactor Vessel / Core Construction

• ANSWER 2.20 (1.00) c REFERENCE NA NCRODP 88.1, 'PCS-PZR and Press Relief' ANSWER 2.21 (1.00) d ' improves cireviation ratio and' velocity across tube sheet)

REFERENCE NA NCRODP 88.1, 'RCS-S/G' ANSWER 2.22 (1.50)

 -RHR System Inlet         (+.5 ca)
 -Loop Drain Lines
 -Differential Pressure taps for RCS flow indication REFERENCE NA NCRODP 88.1, 'RCS-Piping and Instrumentation'

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 63 ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS 4

ANSWER 2.23 (2.00) 2; 1; 2; cesium; yttriumi molybenduni Li-7; Cs-137 (+.25 ea) REFERENCE NA NCRODP 88.3, 'CVCS' 4 ANSWER 2.24 (1.50) a) Open (+.5 ea) b) Open c) Closed REFERENCE r NA NCRODP 88.3, 'CVCS' ANSWER 2.25 (1.00) Medec

   -Lower Delta P (8' vs 12-14' on Urit 1)                                  (+.5 ea)
    - w ax. 3 MW sein f:               U r. ~. t :

REFERENCE NA NCRODP 89.1, ' Main Stam System' ANSWER 2.26 (1.00) i CC Heat Xer Dschs (+.25 ea) Dschs to Unit 2 Discharge Tunnel Oschs to Resevoir Recire Spray Ht Xer . . REFERENCE NA NCRODP 92.2, ' Service Water System' i

                                                                                \

ANSWER 2.27 (1.00) d i I REFERENCE NA NCRODP 92.2, ' Service Water System' j

                                             , ~ . -                                              .

CC Heat Exchanger /SW side. Basement of Aux Bldg. . East end of CC Heat Exchangers on floor. Discharge to tunnel - Turbine Blds. SW pit. Discharge to reservoir - Southeast corner of SW pump house. Racire Spray Hx's - Basement of Quench Spray pump house.

20 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 64 ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS ! ANSWER 2.28 (1.00) c REFERENCE I NA NCRODP 91.2, ' Containment and Containment Systems' ANSWER 2.29 (1.00) d l REFERENCE ' j NA NCRODP 91.2, ' Containment and Containment Systems'

\

I I ANSWER 2.30 '(1.50) l a) Not' allowed (+.5 ea)

!.                           b)                   Allowed c)                  Allowed REFERENCE NA NCRODP 91.2, ' Containment Isolation' ANSWER                          2.31      (1.50)

J a) 3 (+.5 ea) b) 4 c) 2 ,

!                             REFERENCE i                             NA NCR00P 91.2r ' Containment Atmospheric Cleanup System' 1

J l i 4 4 i i i 4 j 1 1 I i

3. INSTRUMENTS AND CONTROLS PAGE 65

. ANSWERS -- NORTH ANNA 182 -85/10/15-TON ROGERS i ANSWER 3.01 (1.00) c , REFERENCE Farley SD, 'PZR Press / Level Control', pp 6, 14, 15, Fig 3 NAPS Lesson plan, RCS Press. Instr. ANSWER 3.02 (1.00) d a REFERENCE FNP, RHR Lesson Plan, pp. 8&9 NA NCRODP 88.2, 'RHR System' 005/000-K4.07 (3.2/3.5)

  ,- ANSWER         3.03              (1.00)

1. RCS pressure > 2010 psig (+.5 ea) i 2. SI 7 rocc FS,g REFERENCE FNP, ECCS Lesson Plan, Figure 9 NA NCRODP 91.1, 'ESF-ECCS' 006/000-A3.01 (4.0/3.9)

ANSWER 3.04 (1.00) a REFERENCE Farley SD, ' Reactor Protection System', pp 33/34 NA Tech Spec pg 2-0 4 0 4 t j

3. INSTRUMENTS AND CONTROLS PAGE 66 ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS ANSWER 3.05 (1.00) a REFERENCE CAT,-PSM, Figures CN-IC-ILE-11 & 22 FNP, SD, 'PZR Pressure and Level Control', Fig 7 NA Lesson Plan, 'PZR Level Control and Protection' 011/000-A2.10 (3.4/3.6) -

j ANSWER 3.06 (1.00) -h!,d- - b REFERENCE FNP, TS pp B 2-3\B 2-6 NA TS pp B2-2/6 ANSWER 3.07 (1.00)

1) Trips not backed up by another protection circuit (+.5 ea)

2) The channel is also being used for control purposes REFERENCE i NA NCRODP 93.10, 'RPS' ANSWER 3.08 (2.00) a) Manual (+.5 ea) ~

b) Auto c) Auto l d) Manual REFERENCE tlA NCRODP 93.10, 'RPS' e r 9 9 {

POWER SUPPLY

                                                                                   -* SUR CK (E)
                                                                             ~

Loss of [S)

                                                                                  ~

A . b

                                             " DETECTOR VOLTAGE                   + HC8 ' RECORDER.

( F-) g) + Hl4H FL11Y @ S/Co (c)

                                  %pe -          0, iw.      De       U"I -

A r +. Ag. + COMT. EVAC.t 3 > Iso

                                  ,grl,      = AUDIO. COUNT RATE
                                                                            -k         > (Hl6H LE FO -

((D tocAt. ::NDicATloM

                                                                        ~

l. SOURCE IPAN6E I'~ y31 n N 32 .

f .* e'

                               .t         .

3. INSTRUMENTS AND CONTROLS PAGE 67 ANSWERS -- NORTH ANNA 182 -

85/10/15-TOM ROGERS ANSWER 3.09 (2.00) See attached dws REFERENCE NA NCRODP 93.2, 'Excore NIS' ANSWER 3.10 (1.50) a) False b) False c) True REFERENCE NA NCRODP 93.2, 'Excore NIS' pp 14-22 ANSWER 3.11 (1.00) c REFERENCE NA NCRODP 93.4, ' Core Cooling Monitor' ANSWER 3.12 (2.00) a) No effect (+.5 ea) b) Arm only c) Arm and Actuate d) No effect REFERENCE NA NCRODP 93.11, ' Steam Dumps' ANSWER 3.13 (1.00) e REFERENCE NA NCRODP 93.12, 'S/G Water Leve19BDwtsp1 and Protection' 1

                              --       - ,------,-----,-----..a-m,.       ,,.----,-e , a y.n-, }, - p. nm - n  v-- -g  -4, ..y.

30 INSTRUMENTS AND CONTROLS PAGE 68 ANSWERS -- NORTH ANNA 1&2 -85/10/15-TON ROGERS ANSWER 3.14 (1.50)

      -S/G Press used for density compensation of steam flow, so if pressure fails high, steam flow input to the SGWLCS will increase (+1.0)
      -Ws>Wf will cause FWRV to open, level will increase until level error counteracts the steam flow / feed flow mismatch              (+.5)

REFERENCE NA NCRODP 93.12, 'SCWLC and Protection' ANSWER 3.15 (1.00) C REFERENCE NA NCRODP 93.13, 'RVLIS' ANSWER 3.16 (1.00) 1) Eliminate accuracy reductions which would be caused by adverse l environment in the containment during an accident. (+.25)

2) Allow easier accomplishment of system operations (calibrations, cell replacements, etc.) (t25)

REFERENCE NA NCRODP 93.13, 'RVLIS' ANSWER 3.17 (1.00) d j REFERENCE NA Les son Plan, ' Rod Control System' 1 1 6 l

l l PAGE 69

3. INSTRUMENTS AND CONTROLS ANSWERS -- NORTH ANNA 182 -85/10/15-TOM ROGERS ANSWER 3.18 (2.00) a) (b)3I b) 6 c) 1 d) 5 REFERENCE NA Lesson Plan, " Rod Control System' ANSWER 3.19 (1.00) a REFERENCE NA Lesson Plan, ' Rod Position Indication' ANSWER 3.20 (1.00)

 . d REFERENCE NA Lesson Plan, 'RCS Pressure Instrumentation' ANSWER     3.21           (1.00)
     -Assume a cold insurge (+.25)
     -Which will cause a pressure drop (+.5)
     -So heaters energized to anticipate this (+.25)

REFERENCE NA Lesson Plan, 'PZR Level Control and Protection' E 4

3o INSTRUMENTS AND CONTROLS PAGE 70 ANSWERS -- NORTH ANNA 1&2 -85/10/15-TON ROGERS 3.13 ANSWER -iPret- (1.50)

 -Allows Unit 1 to have its normal station service busses supplied from its normal station service transformers backfed from 500 KV switchyard for most Unit 1 trips.             (+1.0)
 -This reduces probability of combined loading from both Units normal and emergency buses on the reserve station transformer (+.5)

REFERENCE NA NCRODP 90.1, ' Basic Electrical Distribution

• C4,) $G$ l

                                                                               +'

ANSWER eree .50)

                                                                      ~

1) Low Lube O' *ressure (+.3 ea) b) $) / I

2) High Cr . case Pressure } g7 jggj c, jgg}.

3) High be Oil Temperature v

4) Jacket Coolant Temperature g6 S.90 f 94:).'/, "Epil" 5 Start Failure

                                                                        }lAA$ wit fk"E*Y ANSWER          3.24            (2.00) a) 2 out of 3; 90%i 56 seconds                 (+.25 ea) b) -diesel starts            (+.25 ea)
      -emergency bus isolated from RS xfrmr
      -all loads shedded except Chg Pumps, LHSI pumps and 480 VAC loads
      -output bkr shuts when diesel up to speed and voltage
      -loads sequenced onto bus REFERENCE NA NCRODP 90.4,          'Emers Diesel Generator' ANSWER          3.25            (1.00)

C REFERENCE NA NCRODP 91.1, 'ESF'

• l

30 INSTRUMENTS AND CONTROLS PAGE 71 ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS ANSWER 3.26 (1.00)

1) 0S pump breaker open (+.5 ea)

2) CDA signal reset REFERENCE NA NCRODP 91 1, 'ESF-OSS' ANSWER 3.27 (1.00) c ey&

REFERENCE NA NCRODP 91.1, 'ESF-Recire Spray System' ANSWER 3.28 (1.00) d REFERENCE NA NCRODP 91.1, 'ESF-ECCS'; 90.1, ' Basic Elect Distribution' ANSWER 3.29 (2.00) a) True (+.5 ea) b) False c) False d) True REFERENCE NA NCRODP 91.2, ' Containment Isolation

• ANSWER 3.30 (1.00)

 -Loop stop/ Bypass isolation valves positioned to allow flow                          (+.25 ea)
  -Bearing lift pump running with >700 psis discharge no electrical faults on RCP motor
  -locked rotor protection speed. switch satisfied

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

3. INSTRUMENTS AND CONTROLS PAGE 72 )

ANSWERS -- NORTH ANNA 182 -85/10/15-TOM ROGERS REFERENCE NA NCRODP 88.1, 'RCS-RCPs' ANSWER 3.31 (1.00) b REFERENCE NA NCR00P 88.3, 'CVCS' ANSWER 3.32 (1.00) a REFERENCE NA NCRODP 88.1, 'RCS-PZR and Press Relief

• r W

F 0 r

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

i

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 73

 --- EA515E55153E 55EiiBE--------------~~~-------

ANSWERS -- NORTH ANNA 182 -85/10/15-TOM ROGERS ANSWER 4.01 (1.00) b. REFERENCE NAPS, 1-OP-6.2, p. 7. ANSWER 4.02 (1.00) d. RE'FERENCE NAPS, 1-OP-3.4, p. 13. ANSWER 4.03 (1.00) c. REFERENCE NAPS, 1-AP-11, p.9. ANSWER 4.04 (1.00) a REFERENCE MNS OP/1/A/6100/01 p. 19. CNS OP/1/A/6100/01, Encl. 4.1, 2.95.10. NAPS 1-0P-2.1, p.8. e

                     - . - ~ - .                    - , - ,  , _ _   _y   ,       ., , e- _ _ . - - -      , . - , , _ , ,

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 74 4o

  ~~~~                        ~    ~~~~~~~~~~~~~~~~~~~~~~~~

j RAD 50L66 6AL 66NTR6L ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS ANSWER 4.05 (1.00) Any 4 0 0.25 points each:

1. Bank C position.

2. Bank D position.

3. Auct. High Tavs.

4. IR N35.

5. IR N36.

6. RCS bacon concentration.

REFERENCE NAPS 1.0P-1.5, p.12. ANSWER 4.06 ( .50) FALSE. REFERENCE NAPS 1-OP-1.5, p.10. ANSWER 4.07 (1.00) a REFERENCE ^ NAPS, 1-0P-1.5, p.10. ANSWER 4.08 (1.00) b REFEPENCE NAPS 1-OP-1.1. ANSWER 4.09 (1.00) d e y- ,- , , s .- , - - , --.m.-

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 75

    ~~~~RdEfEEddfC5E~dd TREE ~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS REFERENCE NAPS 1-OP-8.3, p.5. . ANSWER 4.10 (1.00) c REFERENCE NAPS 1-0P-1.4, Attachment 1. ANSWER 4.11 (1.00) b. REFERENCE NAPS 1-AP-4, pp. 388. ANSWER 4.12 (1.00) c. REFERENCE NAPS 1-AP-5.1, pp.8-9. i ANSWER 4.13 (1.00) b REFERENCE

• MNS AP/2/A/5500/16, Case IV, p.9.

> CNS AP/1/A/5500/16, Case IV, p.14. MAPS 1-OP-4, p.S. J e i

40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 76

                           ~
 ~~~~EddfdLddfddL dd TddL'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS ANSWER 4.14 (1.25) 7- N FUE-Any 5 9 0.25 points each:

1. IRPI within demand position limit. 8. A&MT- g,gue(

2. Each IRPI operable.

3. AFD within limits.

4. Containment temperature.

, 5. Control room temperature.

6. Cire. water temperature.

REFERENCE NAPS 1-AP-42. ANSWER 4.15 ( .50) FALSE. (urgent failure alarm could be from logic or power cabinet failure) REFERENCE NAPS 1-AR-ir 1A-4. ANSWER 4.16 (1.00) 0 0.25 points each:

1. Start all available service air compressors.

2. Start all available instrument' air compressors.

3. Secure reactor coolant pumps.

4. Secure charging Pump.

REFERENCE NAPS 1-AP-28.1 p.3. ANSWER 4.17 (1.00) Place both level trip switches in BYPASS. (Ops may continue.) REFERENCE NAPS 1-AP-4, p.4.

4. PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 77

                                  ~
 ~~~~Ed656L66f6dL 66NT66L'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS ANSWER 4.18 (1.00) 0 0.25 points each

1. Shift rod control to manual.

2. Control turbine to match Tavs and Tref.

3. Ensure Tavs is > 541 or get >541 within 15 sin or trip Rx.

4. Notify Shift Supervisor.

REFERENCE NAPS 1-AP-1.4, p.3. ANSWER 4.19 (1.00) 0 0.25 points each:

1. Go to BORATE on Blender Mode Switch.

2. Adjust FC-113A to allow 10 or more gpm flow rate.

3. Turn Blender Control Switch to START.

4. Verify flow.

REFERENCE NAPS 1-AP-2.0, p.3. ANSWER 4.20 (1.00) 0 0.5 points each

1. Place or verify rods in MANUAL.

2. Place or verify steam dumps in steam pressure mode.

REFERENCE NAPS 1-AP-3, p.4. )

  -     ,, ~ - - , .         .m..   ,      ,   . . ,      , . - . . - - . - -

c - ~ -c. g --, ., - . , ,,.,,.w, ,

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 78

    ~~~~Rd656[6656dE~66HTR6E-~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTH ANNA 1&2 -85/10/15-TON ROGERS ANSWER 4.21 (1.00) 0 0.25 points each

1. Reduce generator load in attempt to stabilize vacuum.

2. Check vacuum breaker closed and water seal present.

3. Manually trip turbine at 9.5' HG abs if it has not automatically done so.

4. Trip reactor.

REFERENCE NAPS 1-AP-14, p.3. ANSWER 4.22 (1.00) 0 0.5 points each:

1. Trip turbine locally.

  -  2. Manually open reactor trip breakers or the rod drive MG output breakers.

REFERENCE NAPS 1-AP-20, p.3. ANSWER 4.23 (1.00) For 0.2F points each.

1. Attempt a reactor trip using both manual trip switches.

2. Attempt a turbine trip using both manual trip buttons.

3. Verify all AFW pumps are running.

4. Verify AFW flow to all SGs.

F$F rh NAPS 1-ECA-1, pp. 2&3. 9

C i

_ - , . , , . . - - , _ . _ . . + . - - ,. . - . . , , . - - -

i PAGE 79 l

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND

                          ~        ~~~~~~~~~~~~~~~~~~~~~~~~
 ~~~~R 5656L66fddL 66 TREL ANSWERS -- NORTH ANNA 1&2                              -85/10/15-TOM ROGERS ANSWER         4.24            (3.25)

For 0.25 points each.

1. Verify reactor trip.

2. Verify turbine trip.

3. Verify AC emergency busses energized.

4. Check if SI actuated.

5. Check charging /SI pump valve lineup.

6. Verify FW isolation.

7. Verify AFW flow.

8. Verify charging /SI flow.

9. Verify LHSI pumps running.

10. Verify containment phase A isolation.

11. Verify service water pumps running.

12. Verify RCS heat removal.

13. Check containment pressure.

REFERENCE NAPS 1-EP-0,pp. 1-5. ANSWER 4.25 (1.00) 6. REFERENCE NAPS 1-ES-0.2, Foldout page. 4 ANSWER 4.26 (1.00) b REFERENCE MNS EP/2/A/5000/1.1, p.5. CNS EP/1/A/5000/1A1, p.4. NAPS 1-ES-0.3, p.5. ANSWER 4.27 (1.00) d 1 l

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 80

 ~~~~R EEf6[E5fCd[~dd TEE [~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS REFERENCE MNS EP/2/A/5000/16.0 CNS EP/1//A/5000/2F3, p.7. NAPS 1-FRP-I.3A, p.3. ANSWER 4.28 (1.00) C REFERENCE Vapco GET, p.49. ANSWER 4.29 (1.00) C REFERENCE Vepco GET, p.49. ANSWER 4.30 (1.00) Reset it. REFERENCE NAPS 1-8CA-0.1, p.2. ANSWEP 4.31 (1.00) 9 0.5 points each'

1. Containment pressure at 19.7 psia.

2. Containment rad. levels at 40,000 R/hr.

REFERENCE NAPS 1-EP-0, p.2. ANSWER 4.32 ( .50) Higher. 9

       - _.      y

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 81 40

  ~~~~Ed6f6[66f6d[~66kTE6[~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTH ANNA 1&2 -85/10/15-TOM ROGERS REFERENCE .. NAPS 1-EP-0 Foldout page. ANSWER 4.33 (1.00) For 0.5 points each. (a) 750 mr/qtr (b) 2750 mr/qtr REFERENCE NAPS Radiation Protection Manual, p.2 3-7. ANSWER 4.34 (1.00) c REFERENCE 10 CFR 20.101 , ANSWEN 4.35 (1.00) i i t GF=1 for gamma 100(45/60)(1)=75 REFERENCE 10 CFR 20. ANSWER 4.36 (1.00) d REFERENCE 10 CFR 20. 4 l l l i I l 1 < 1 1

                     ~    - - ~ -
                                        ---r     - - - . - .--..y+u,_-,,-,_y..__      ,_. ,_ _ _ , _ . . _ [ ._,k,  . _    _

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 82

 ~~~~R dEf6[6dfCdE~dd TR6E~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTH ANNA 1&2 -85/10/15-TCM ROGERS ANSWER 4.37 (1.00) c REFERENCE Vepco GET, p.21. ANSWER .4.38 (1.00) Any 5 0 0.2 points each:

1. Job location.

2. Brief job description.

3. Radiological conditions at job site.

< 4. Anti C requirements.

5. Respiratory requirements.

6. Dosimetry requirements.

7. Necessary HP coverage.

8. Any special instructions.

REFERENCE Vepco GET, p.41 ANSWER 4.39 (1.00) C REFERENCE Vepco GET, p.41. e

                                        , . - .    ,c ._. , . , . . . . , ,   , , - 4   ,  .e ,   w,_

                                                                                           --                                                                       1 l   .

r s s' U. S. NUCLEAR REGULATORY COMMISSION l SENIOR REACTOR OPERATOR LICENSE EXAMINATION  ! l FACILITY: NORTH ANNA 1&2 REACTOR TYPE: PWR WEC3 DATE ADMINISTERED: 35/10/15 EXAMINER: DEAN, WH APPLICANT: _[_ _@;9_(_N_ _ [_ #f_E_ _ _ /[f INSTRUCTIONS TO APPLICANT: 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 Srade requires at least 70% in each category and a final grade of at least 80%. Enamination papers will be picked up sin (6) hours after the examination starts.

                                                         % OF CATEGORY           % OF        APPLICANT'S      CATEGORY VALUE         TOTAL               SCORE       VALUE                                 CATEGORY 40.00                                                              THEORY OF NUCLEAR POWER PLANT

________ __'5.00____ ___________ ________ 5. OPERATION, FLUIDS, AND THERMODYNAMICS 3 g,f

      **,M

__'5.OO ___________ ________

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 40 0 ~5 0 PROCEDURES - NORMAL, ABNORMAL,

___I__0__ __~_I_0 _ ___________ ________ 7. EMERGENCY AND RADIOLOGICAL CONTROL 3 g, p

          - ':         ,5

_ _";;;;_ _ _ _'_I_0 0_ ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS g y-160.00 100.00 TOTALS FINAL GRADE _________________% All work done on this e:: amination is my own. I have neither SivGn not received aid. 4 J f

50 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE : g-------------------------------------- GUESTION 5.01 (1.00) The reactor is Producing 100% rated thermal power at a core delta T of 60 degrees and a mass flow rate of 100% when a blackout occurs. Natu-tal circulation is established and core delta T goes to 40 degrees. If decay heat is 2%, what is the core mass flow rate (in %)?

a. 1.3

b. 2.0

c. 3.0

d. 4.0 QUESTION 5.02 (1.00)

Which of the following does NOT contribute to assuring that the enthalpy rise hot channel limits are maintained?

a. Axial power distribution is maintained within limits.

b. Control rod banks are sequenced with proper overlap.

c. Control rod insertion limits are maintained.

d. The MTC is within its analyzed temperature range.

QUESTION 5.03 (1.00) Which of the following is NOT a reason for pressurizing the fuel rods with helium?

a. Minimize clad creeping inwards toward fuel pellets.

b. Increase gap (pellet to clad) thermal conductivity.

c. Allow detection of clad failure by helium analysis of the Coolant.

d. Maintain lower fuel centerline temperature.

(xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxx*). r

U. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5.04 (1.00) Which of the following is NOT one of the conditions necessary for brittle fracture to occur?

a. Plastic deformation at or below the yield point.

b. Temperature at or below the NDTT.

c. Nominal stress level. .

d. Flaw such as a crack present.

QUESTION 5.05 (1.00) Which statement below describes centrifugal pump runout conditions?

a. High pressure, low flow, high power demand

b. High pressure, low flow, low power demand

c. Low pressure, high flow, high power demand

d. Low pressure, high flow, low power demand

e. Low pressure, low flow, high power demand QUESTION 5.06 (1.00)

An ECF is calculated for a reactor startup 4 hours after a reactor trip from 100% power, equilibrium Xe conditions. Which of the following would cause the actual critical, position to be lower than the ECP?

a. The startup is delayed until 8 hours after the trip

b. Actual boron concentration is 10 ppm more than the predicted boron concentration.

c. A rod finger is separated from its spider assembly.

d. The steam dump pressure setpoint is lowered by 100 psi prior to reactor startup.

(xxxxx CATEC0h) 05 CONTINUED ON NEXT PAGE xxxxx)* De

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 g

GUESTION 5.07 (1.00) Which of the following will cause the Axial Flux Difference to become core positive (less negative)?

a. Power increase with power defect compensated for by dilution only.

b. Power increase with power defect compensated for by '

rod withdrawal only.

c. Buildup of xenon in top portion of core.

d. Burnup of xenon in bottom portion of core.

QUESTION 5.08 (1.00) Which of the follow'ing will cause the Fuel Temperature Coefficient (pcm per degree) to become LESS negative?

a. Fuel temperature increase.

b. Boron concentration decrease.

c. Control rod withdrawal (at constant power).

d. Core age increase.

QUESTION 5.09 (1.00) Which of the following describes the changes that steam in a REAL turbine undergoes comparing conditions at the inlet to those at the outlet?

a. Enthalpy DECREASES, Entropy DECREASES, Quality DECREASES

b. Enthalpy INCREASES, Entropy INCREASES, Quality INCREASES

c. Enthalpy CONSTANT, Entropy DECREASES, Quality DECREASES

d. Enthalpy DECREASES, Entropy INCREASES, Quality DECREASES s (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE *****) ,

9

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

• QUESTION 5.10 (1.00)

Which of the following tensile stresses is higher on the outer wall than on the inner wall of the reactor pressure vessel?

a. Pressure stress.

b. Composite (Total) stress during cooldown.

c. Cooldown stress (due to delta T only). ,

d. Heatup stress (due to delta T only).

QUESTION 5.11 (1.00) Which of the following statements concerning Xenon-135 production and removal is correct?

a. At full power, equilibrium conditions, about half of the xenon is produced by iodine decay and the other half is produced as a direct fission product.

b. Following a reactor trip from equilibrium conditions, xenon peaks because delayed neutron precursors continue to decay to ,:enon while neutron absorption (burnout) has ceased.

c. Xenon production and removal increases linearly as power level increases; i.e., the value of 100% equilibrium

                  .:e n o n is twice that of 50% equilibrium xenon.

d. At low power levels, ::enon decay is the major removal method. At high power levels, burnout is the major removal method.

QUESTION 5.12 5: 1.00) Which of the following is the units of heat flux?

s. Watts / cubic centimeter

b. E:TU / (hr square ft)

c. Calories / gram

d. kW / ft

(***** CATEGORY 05 CONTfNUED ON NEXT PAGE ****m)

                -            ,_                   c                     - , . -   -

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE e 50 QUESTION 5 13 (1.00) Which of the following statements concerning the power defect is correct?

a. The power defect is the difference between the measured power coefficient and the predicted power coefficient.

b. The power defect increases the rod worth requirements necessary to maintain the' desired shutdo'wn margin following a reactor trip.

c. Because of the higher boron concentration, the power defect is more negative at beginning of core life.

d. The power defect necessitates the use of a ramped Tavs program to maintain an adequate Reactor Coolant System subcooling margin.

QUESTION 5.14 (1.00) The reactor is critical at 10,000 cps when a S/G PORV fails open. Assuming BOL conditions, no rod motion, and no reactor trip, choose the answer below that best describes the values of Tavs and nuclear power for the resulting new steady state. (POAH = point of adding heat).

a. Finst Tavs greater than initial Tavs, Final power above POAH.

b. Final Tavs greater than initial Tavs, Final power at POAH.

c. Final Tavg less than initial Tsvs, Final power at POAH.

d. Final Tave less than initial Tavs, Final power above POAH.

(***xx CATEGORY 05 CONTINUED ON NEXT PAGE **xxx)*

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7 QUESTION 5.15 (1.00)

The Moderator Temperature Coefficient (MTC) varies with certain plant conditions. Concerning these variations, whien of the.following i s correct?

a. The MTC becomes more negative as boron concentration is increased.

b. The MTC causes a::ial flux' distribution t'o be tilted towards the top of the core at BOL.

c. The MTC varies as temperature changes because of the non-linear density chan3es of water as temperature changes.

d. The MTC is not permitted by Technical Specifications to be positive in any plant operating modes.

, GUESTION 5.16 (1.00) Which of the following statements concerning Shutdown Margin (SDM) is correct?

a. The maximum SDM requirement occurs at EOL and is based on a rod ejection accident.

b. The maximum SDM requirement occurs at EOL and is based on a steam line break accident.

c. The ma::imum SDM requirem'ent occurs at BOL and is based on having a positive moderator temperature coefficient.

d. The maximum SDM requirement occurs at BOL and is based on a r od withdr awal accident while in the source range.

(xxxxx CATEG0FY 05 CONTINUED ON NEXT PAGE xxxxx)=

                                  -,.-g                               , - w - . , - - - w

. 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6 QUESTION 5.17 (1.00) The reactor is operating at 30% power when one RCP trips. Assuming the reactor does not trip and no turbine load chan3es occur, which of the following parameters will DECREASE?

a. Flow in operating reactor coolant loops

b. Core delta T

c. Reactor vessel delta P

d. Steam generator pressure in affected loop OUESTION 5.18 (1.00)

The Technical Specifications allow operations for a 2-hour time period with a Guadrant Power Tilt Ratio (GPTR). of greater than 1.02. Whici. of the following is the reason for allowing operations for these 2 hours?

a. To allow time for corrective action in the event of xenon redistribution following power changes.

b. To allow time for correction of a dropped or misaligned control rod.

c. To allow time for boron concentration changes to restore the OPTR to less than 1.02.

d. To allow time for correction of RCS flow imbalances.

(***** CATEGORY.05 CONTINUED ON NEXT PACE *****) WO I I

So THEORY OF NUCLEAR F0WER PLANT OPERATION, FLUIDS, AND PAGE 9 QUESTION 5.19 (1.00) During a reactor trip recovery, the initial 1/M data point was 1.0. After a 1-hour delay, rod withdrawal was commenced. Upon stcpping rod withdrawal to take 1/M data, the RO reported that the second 1/M data point was 1.1. Which of the following would explain this increase in the 1/M value?

a. This is NOT possible, the.R0 must have made an error when taking count rate data.

b. The buildup of xenon during the 1-hour delay added more negative reactivity than the rod' withdrawal added positive.

c. Compensating voltage on the IR instruments fails low.

d. An inadver, tent dilution is in progress.

QUESTION 5.20 (1.00) During a reactor startup in a suberitical reactor, the first reactiyity addition caused count rate to increase from 10 cps to 16 cps. The second resetivity addition caused count rate to increase from 16 cps to 32 cps. Which of the following statements describing the relationship between the first and second reactivity additions is correct?

a. The first reactivity addition was larger.

b. The second reactivity addition was larger.

c. The first and second reactivity additions were equal.

d. There is not enough data given to determine relationship between reactivity values.

(xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) , -. - - _. ., + ._

                                                                                                              .-,n,m--4      - - - . . ~ , .

THEORY OF HUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 10 + 5. gg g-------------------------------------- 00E5 TION 5.21 (1.00) Which of the following is defined as 'the fractional change in neutron population per generation *? ,

a. keff

b. Delta k o

c. Reactivity .

d. Delta reactivity QUESTION 5.22 (1.00)

Which of the following can be defined as 'the number of neutrons causing fission that were originally born delayed divided by the tota,1 number of neutrons causing fission'?

a. Lambda effective

b. Rho

c. E: eta ef f ective

d. Tau GUESTION 5.23 (1 00)

Which of the following e:: presses the relationship between differential rod worth (DRW) and integral rod worth (IRW)?

a. DRW is the slope of the IRW curve at that location.

b. DRW is the area under the IRW curve at that location.

c. DRW is the square root of the IRW at that location.

d. There is no relationship between DRW and IRW.

4 (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) i 1 I

          - ._ . _ _ . . . _ - _ - -            .,    .- . _ . -    . _ -. - - ,     ,    ,~ . . - , . -.
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So THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 11 QUESTION 5.24 (1.00) Which of the following conditions would result in the hijhest available net positive suction head?

a. Pump with heat source downstream and heat sink upstream with surge tank on its discharge line.

b. Pump with heat source downstream and heat sink upstream with sur3e tank on its suction-line. '

c. Pump with heat sink downstream and heat source upstream with l

surge tank on its discharge line.

d. Pump with heat sink downstream and heat source upstream with surge tank on its suction line.

QUESTION 5.25 (1.00) Which of the following statements concerning samarium reactivity effects is correct?

a. The equilibrium (at power) value of samarium is dependent upon power level. The peak value of samarium following a shutdown i s dependent upon power level prior to shutdown.

b. The equilibrium (at power) valve of samarium is dependent upon power level. The peak value of samarium following a shutdown is independent of power level prior to shutdown.

c. The equilibrium (st power) value of samarium is independent of power level. The peak value of samarium following a shutdown i s dependent upon power level prior to shutdown.

d. The equilibrium (at power) value of samarium is independent of power level. The peak value of samarium following a shutdown is independent of power level prior to shutdown.

(xxxxx CATEGOPY 05 CONTINUED ON NEXT PAGE x*xxx) e e

50 THEORY OF NUCLEAF POWEF PLANT OPERATION, FLUIDS, AND PAGE 1; QUESTION 5.26 (1.00) The reactor startup procedere requires that the critical rod position be taken at 10E-8 amps on the intermediate range. If, curing a nenon free reactor startup at MOL, the operator ' overshoots' 10E-8 amps and instead leveled off at 10E-7 amps, which of the following statements is correct?

a. At 10E-7 amps, there are little or no ef.fects from nuclear heati but, since the reactor is a decade higher in power, the critical rod position would be higher than at 10E-8 amps.

b. At 10E-7 amps, there are little or no effects from nuclear heati therefore, the critical rod position should be the same as at 10E-8 amps.

c. At 10E-7 amps, ther e are substantial ef f ects from nuclear heati therefore, the critical rod position will be higher than at 10E-8 amps.

d. At 10E-7 amps, nuclear heat, xenon, and being a decade higher in power level will result in a higher critical rod posi-tion than at 10E-8 amps.

GUE5 TION 5.27 (1.00) The reactor trips from full power, equilibrium xenon conditions. Sin hours 1 ster the reactor is brought critical at 10E-8 amps on the inter-mediate rsnge. If power level is maintained at 10E-B amps which of the following statements concerning rod motion requirements for the next two hours is correct?

a. Rods will have to be withdrawn since xenon will closely follow its normal build-in rate following,a trip.

b. Rods will have to be inserted since nenon will closely follow its normal decay rate following a trip.

c. Rods will have to be rapidly inserf.ed since the critical reactor will cause a high rate of furnout,

d. Rods will have to be rapidly, withdrawn since the cri,tical reactor will cause a higher than normal rate of build-in.

(<**** CATEGORY 05 CONTINUED dN NEXT PAGE ***xx) 6 I 1 I I

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 13 GUESTION 5.28 (1.00)

Which of the following is a reason for shifting the SI mode from cold les recirculation to hot les recirculation approximately 24 hours after e LOCA?

a. Increase recirculation flow through the core.

b. Ensure boron precipitation does not cause long-term cooling problems.

c. Allow sampling RCS to determine amount of core damage.

d. hinimize the hydrogen concentration in the coolant.

QUESTION 5 29 (1.00) Which of the following would be an indication of a loss of natural circulation flow?

a. RCS delta T less than full power delta T.

b. S/G pressure decreasing rapidly as S/G 1evel is increased.

c. RCS subcooling decreasing.

d. Cold les at saturation temperature for S/G pressure.

QUE3 TION 5.30 (1.50) Indicate whether the following statements are TRUE or FALSE.

s. A positive 100 pcm reactivity addition from criticality low in the intermediate range will produce the same constant startup rate throughout core life. (0.5)

b. A delayed neutron has a higher probability of causing fission than does a prompt neutron. (0.5)

c. If a reactor is supercritical, the fraction of delayed neutrons shifts to the shorter lived precursors and the value of the average decay constant (lambda) decreases. (0.5'

(***** CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) L

S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 14 3 QUESTION 5.31 (1.00) When performing a reactor S/U to full power that commenced five hours after a trip from full power equilibrium conditions, a 2%/ min ramp was used. How would the resulting xenon transient vary if instead a 0.5%/ min ramp was used?

a. The nenon dip for the 0.5%/ min ramp would occur sooner and be smaller. .

b. The xenon dip for the 0.5%/ min ramp would occur later and be smaller.

c. The xenon dip for the 0.5%/ min ramp would occur sooner and be larger.

d. The xenon dip for the 0.5%/ min ramp would occur later and be larger.

QUESTION 5.32 (1.00) Which of the following conditions would cause a 1/M curve to predict criticality earlier than it will actually occur?

a. Source located too near detector.

b. Fuel assemblies loaded too far from detector.

c. Highest worth fuel assemblies loaded first.

d. Control rod located between fuel assemblies loaded and detector.

QUESTION 5.33 (1.00) Which of the following will result in the largest INCREASE in the concentration of dissolved gases in a quantity of water? (Assume the changes in temperature and pressure below are of equal magnitude)

a. '.ncreasing the pressure and lowering the temperature.

b. Decreasing the pressure and lowering the temperature.

c. Increasing the pressure and raising the temperature.

d. Decreasing the pressure and raising the temperature.'

(xrxxx CATEGORY 05 CONT,INUED ON NEXT PAGE xxxxx) e m . . - - - y-- w-

THEORY OF i JC. EAR POWER PL ANT OPERATION, FLUIDS, AND PAGE 15 U. ______________.3 , .____________________________________ THERMODYNAS1(d . QUESTION 5.34 (2.00) If steam goes through a throttling process, indicate whether the following parameters will INCREASE, DECREASE, or REMAIN THE SAME. (0.5)

a. Enthalpy (0.5)

b. Pressure (0.5)

c. Entropy .

(0.5)

d. Temperature GUESTION 5.35 (2.00)

A motor driven centrifugal pump is operatingHow at (INCREASE, rated flow.DECREASE, You then start closing down on the discharge valve. or REMAIN THE SAME) will each of the following be affected? (0.5)

c. Flow ,

(0.5)

b. DischarSe Pressure (0.5)

c. Available NPSH Motor Amps (0.5) d.

(x*xxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx) t l l l l  ?  !

3. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 16 ggg--------------------------------------

QUESTION 5.36 (2.00) Answer TRUE o.c FALSE to the following.

c. The use of a ramped Tave program allows the plant to operate with a higher thermodynamic efficiency than does a constant Tave program. (0.5)

b. Increasing condensate depression (subcooling) will cause BOTH a decrease in plant ef ficiency *AND an increas'e in condensate (hotwell) pump available NPSH. (0.5)

c. During a RCS heatup, as temperature Sets higher, it will take a smaller letdown flow rate to maintain a constant pressuri:er level. (0.5)

d. The difference between pump su tion pressure and the saturation pressure of the fluid being pumped is referred tc as net posi-tive suction head. (0.5)

QUESTION 5.37 ( .50) TRUE or FALSE? For heat exchangers with the same heat transfer area, the same mass flow rates and with incoming fluid temperatures the same for both the cooling and cooled mediums, a counterflow heat exchanger will transfer more heat than a parallel flow heat exchanger. (***** END OF CATEGORY 05 *****) 4 I l I I

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

6. PLANT SYSTEh5 DESIGN, CONTROL, AND INSTRUMENTATION PAGE 17 QUESTION 6.01 (1.00)

Which of the following conditions is NOT required for automatic swapover of the LHSI pumps to the Recirculation Mode following an SI?

a. RWST Lo-Lo Level

b. A LHSI pump recire isolation MOV closed for each pump

c. SI signal present

d. SI Recire Mode signal present QUESTION 6.02 (1.00)

, Which statement below regarding pressuri er spray is correct?

a. The position of the auxiliary spray valve can be adjusted from fully shut through the fully open position by the operator at the main control board.

b. Each of the normal spray valves has a 1 spm bypass line to prevent cooldown of the spray lines.

c. The ma::imum spray flow rate is designed to prevent PORVs from opening during a 10% step DECREASE in power.

d. Au::ili a r y spray utili:es a separate spray no::le due to the large temperature differential that exists when auxiliary spray is used.

QUESTION 6.03 (1.00) 9hich of the follcwing functions.is NOT performed by the hydraulic power system associated with the Containment Personnel Air Lock?

a. Operation of the Equali:ing Valve

b. Door Lock ins Ring Rotation

c. Interior Door Operation

d. E::t e r io r Door Operation (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx).

60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE la GUESTION 6.04 (1.00) PT-402 and 403 measure RCS pressure upstream of RHR suction isolation valves MOV 1700 and 1701. Which of the following statements concerning the effect(s) of PT-402 failing upscale (high) is correct?

a. If both MOV 1700 and MOV 1701 are open they will shut.

b. If open, MOV 1700 will shut and if closed, NOV 1701 will not be able to be opened.

c. If both MOV 1700 and MOV 1-701 are shut, they will not be able to be opened. -

d. If open, MOV 1700 will shut, but MOV 1701 can be positioned as desired by the operator.

e. MOV 1700 can be positioned as desired by the operator, but MOV 1701 will not be able to be opened if it is shut.

QUESTION 6.05 (1.00) Which of the following statements concerning the operation of the letdown isolation valves (LCV-1460A and B) is correct?

a. Neither of the tetdown isolation valves ccn be opened if the containment letdown isolation valve (TV 1204) is shut.

b. Shutting all orifice isolation valves will automatically shut the letdown isolation valves

c. All orifice isolation' valves must be shut in order to open the letdown isolation valves

d. High temperature on the letdown outlet of the regenerative heat exchanger will automatically shut the letdown isolation valv'es (vrxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx) t

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 19 GUESTION 6.06 (1.00)

Which of the following describes the purpose of the time delay (195 sec) in starting the Inside Recirculation Spray (IRS) pumps on a CDA signal?

a. Prevents overloading of the emergency buses as loads are sequentially energized during accident conditions.

b. Allows time for the containment sumps to collect sufficient fivid to provide NPSH for the pumps.

c. Enhances core cooling by increasing refl'ood rate af ter a LOCA as the pressure drop between core exit and the break is reduced with a higher containment pressure.

d. Allows time for the fluid collecting in the containment sumps to cool, to avoid flashing in the RS heat exchangers as the fluid is cooled by Service Water.

GUESTION 6.07 (1.00) Which cf the following describes the Service Wa.ter System automatic actions on a single unit SI?

a. All SW pumps start, that unit's spray header isolation NOVs receive an 'open' signal

b. Only the SW pumps supplied from that unit's emergency buses start, that unit's spray header isolation MOVs receive an 'open' signal

c. All.SW pumps start, all spray header isolation HOVs receive an

              'open' signal

d. The SW pumps supplied from that unit's emergency buses start, all spray header isolation MOVs receive an 'open' signal GUESTION 6.08 (1.00)

Which location below is the discharge point for the pressurizer head vent?

a. Containment fuel canal

b. Upper region of containment below quench spray rings

c. Pressurizer Relief Tank

d. Suction side of containment Hydro 3en Recombiners (xxxxx CATEGORY 06 CONT,INUED ON NEXT PAGE xxxxx)

                                                  -    -      f _           _ _ _

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20 QUESTION 6.09 (1.00)

Which of the following signals, acting independently, will automatically CLOSE the Main Feedwater Regulating Valves (MOV-FW-1154A, B and C)?

a. Hi-Hi S/G level in any S/G (2 out of 3 detectors)

b. Low Tavs (2 out of 3)

c. Reactor Trip

d. Phase B Isolation QUESTION 6.10 (1.00)

Which graph on the attached page represents the change in the internal pressure of a fuel rod versus fuel burnup?

a. A

b. B

c. C

d. D

e. E GUESTION 6.11 (1.00)

The pressuri:er safety valves are designed to limit RCS pressure to within the safety limit of 2735 psig following a complete loss of turbine generator load while at rated thermal power. Which of the following citigating events is zisumed to have occurred in the analysis for this design basis?

a. Reactor trip due to turbine loss of load

b. PORV operation

c. Hi PZR pressure trip

d. Both spray valves actuate

e. Relief valves on all 3 S/Gs actuate GUESTIJN 6.12 (1.00)

Which of the following is a load using Service Water as a BACKUP source of heat removal?

a. Service Air Compressors

b. Instrument Air Compressors

c. Control Room Air Conditioning Units .

d. Containment Air Recire Coolers (xxxxx CATEGORY 06 CONT'INUED ON NEXT PAGE xxxxx)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 21 QUESTION 6.13 (1.00)

Which statement below regarding the North Anna offsite power system is correct?

a. 22 KV from offsite can be used to power BOTH the Reserve Station and Station Service Transformers.

b. A Reserve Station Transformer consists of a bank of three single-phase transformers.

c. All three Reserve Station' Transformers c'an be supplied, via switching circuits, from EITHER of two 500/34.5 KV transformers.

d. North Anna Power Station is connected to the VEPCO grid by three 500 KV lines of which a MINIMUM of two are required to carry the output of BOTH units.

QUESTION 6.14 (2.00) Indicate whether the following situations will cause the Steam Dump System to ARM ONLY, ARM AND ACTUATE or have NO EFFECT. (Assume all switches are in normal positions unless otherwise indicated) a) 60% power, 3 %/ min load reject, Tavs> Tref by 8 degrees F, Mode Selector Switch in Tavg. b) Turbine trip from 25% power, Tavg=545 degrees F, Mode Selector Switch in Tavg. c) Hot Zero Power, Tavg=550 degrees F, Mode Selector Switch in STM PRESS, 1020 psis setpoint selected.. d) 40% power, 5%/ min load increase, Pimp CH-III fails low, Tavg=560 degrees F, Mode Selector Switch in Tavs QUESTION 6.15 (1.50) Indicate whether the following statements apply to Fire Pump 1-FP-P-1, Fire Pump 1-FP-P-2 or to BOTH. a) 2500 GFM, diesel driven. pump b) Recirculates its discharge to the Unit 1 A-well J c) The pump can only be stopped locally , (xxxxx CATEGORY 06 CONT,INUED ON NEXT PAGE xxxxx)

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6. FLANT SYSTEMS DESIGN, CONTROL, AND INSTRUhENTATION FAGE  ;;

QUESTION 6.16 (1.50) Answer the following questions regarding the Encore NIS TRUE or FALSE. a) An Intermediate Range detector that is overcompensated could result in the P-6 setpoint never being reached on a shutdown. b) Flacing the Source Range instrument in ' Bypass' at the instrument drawer will prevent a reactor trip signal if either the Instrument or the Control Power Fuses are pulled. c) Neither the Source Range or the Intermediate' Range 'Hi Flux' trips are taken credit for in accident analysis. QUESTION 6.17 (1.50) Indicate whether the following containment isolation valve combinations are ALLOWED or NOT. a) One LOCKED CLOSED isolation valve INSIDE and a simple check valve used as an AUTOMATIC isolation valve OUTSIDE b) An AUTOMATIC isolation valve INSIDE (not a simple check valve) and a LOCKED CLOSED isolation valve OUTSIDE containment. c) AUTOMATIC isolation valves BOTH INSIDE and OUTSIDE containment with the INSIDE valve a simple check valve. QUESTION 6.18 (1.50) Indicate whether the following controls for 1H EDG are transferred to/from the 1H diesel room using the ' Control Room Emergency'- (CRE) switch or the

 ' Diesel Control Isolation' (DCI) switch.

a) Control and synchronization for the 4.16 KV Emergency Bus emergency supply breaker. b) Alternate governor contro.1.s by means of the Droop /Isochronous Selector switch. c) Diesel Start /Stop control (***** CATEGORY 06 CONTINUED ON NEXT PAGE **xxx)

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

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 23 6. 1 QUESTION 6.19 (1.50) Match the Containment Atmospheric Cleanup System component in Column A with the correct location in Column B. COLUMN A COLUMN B a) H2A-HC-200 H2 Analyzer 1) Aux Bldg near Drumming Area b) H2A-HC-201 Remote Analy:er Panel 2) Below Rod Drive Room c) 1-HC-HC-1 H2 Recombiner 3) Aux Blds near Domin Alley

4) Unit 2 Instrument Rack Roon

5) By High Rad Sampling Panels QUESTION 6.20 (1.50)

Match-the Main Steam Valve description listed in Column A with the cppropriate valve listed in Column B. Column A Column B a) Valve disc is independent of stem 1) Main Steam Trip Valve and floats on steam flow (TV-MS-101 A,B,C) b) Valve disc is held out of position 2) Equalizing Trip Valve by use of air pistons (TV-MS-113 A,B,C) c) Air is supplied to both the top and 3) Non-Return Valve bottom of valve actuator (NRV-MS-101 A,B,C)

4) AFW Pump Steam Supply Valve (TV-MS-111 A,B)

5) Equali:ing Non-Return Valve (NRV-MS-103 A,B,C)

GUESTION 6.21 (1.50) With the reactor at 50% Power, S/G A's steam pressure input to the Steam Generator Water Level Control System fails high. Explain how this will affect the Level Control System for S/G A, and why this action occurs. (mmmum CATEGORY 06 CONT,INUED ON NEXT PAGE *****)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 24 QUESTION 6 22 (2.00)

Attached is a drawing of the Au::iliary Feedwater System. List the valves downstream of the AFW pumps which are closed in a normal MODE 1 lineup. QUESTION 6.23 (1.50) What are the three groups of RCS piping penetrations made BELOW the horizo,ntal centerline of the piping? (penetrations with a similar purpose are to be treated as one group) OUESTION 6.24 (1.00) Unit 2's Moisture Separator Reheaters (HSRs) have a hexagonal tube bundle which has almost twice as many tubes as Unit l's MSRs, but the Unit 2 MSRs are wider and shorter than Unit l's. What effect, if any, does this difference in MSR construction have on the Differential Pressure across Unit 2's MSRs and the plant's thermal power capabilities? (Indicate both direction and approximate magnitude of any differences noted) GUESTION 6 25 (1.75) With the exception of indication, list all the protective, control and permissive circuits which receive an input from PT-1446 and PT-1447 (turbine first stage pressure channels III and IV). Identify circuits by nomenclature where applicable. (eg P-6) QUESTION 6.26 (1.00) Explain why the load shedding feature associated with a degraded or undervoltage condition is defeated after an Emergency 4160 VAC Bus i s' isolated from the Reserve Station Transformer and the diesel output breaker closes. QUESTION 6.27 (1 25) Describe how the High Steam Line Flow SI input varies and the parameter on which this program is based. (xxxxx CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 25 GUESTION 6.28 (1.50)

List all the methods, including the actuating device, by which the contents of the Lube Oil /EHC Pump Facility

• Deluge
• system can be dumped. Include any appropriate setpoints.

QUESTION 6.29 (1.00)

List all the electrical inputs to the starting circuitry that must be satisfied to successfully start a Reactor Coolant Pump. QUESTION 6.30 (1.50) Describe the improvements _ _...: and reliability of the electrical distribution system following a Unit 1 Trip due to the installation of Unit 1 Main Generator Breaker G12. QUESTION 6.31 (2 00) Answer the following questions regarding a DEGRADED voltage condition on an Emergency 4160 VAC Bus. a) A degraded voltage condition occurs when ____ out of ____ relays (.75) detect <_____% nominal voltage and initiates a ____ second timer. (Fill in the blanks) b) Describe the sequence of events from the timing out of the timer (1.25) until all required loads are reenersi:ed on the Emergency Bus. (Be sure to list any loads NOT shed) (xxxxx END OF CATEGORY 06 xxxxx) b t

• 6

7. PROCEDURES - NORMAL, AE:N O R M AL , EMERGENCY AND PAGE 26

~~~~R d656E65fC E 66 TR6C--------~~~------~~~---- QUESTION 7.01 (1.00) Which of the following is NOT required operator action upon' loss of power range channel N-41 during Mode 1 operations?

a. Place the test switch for the OT delta T to the TEST position for the failed channel.

b. Place or verify the control rods are in , MANUAL control.

c. Place the test switch for the OP delta T to the TEST position for the failed channel.

d. Remove the ' Control Power' fuses for N-41 at the power range drawers.

QUESTION 7.02 (1.00) Which of the following is an INAPPROPRIATE operator action ife during refueling, the containment gas and particulate radiation monitors alarm?

a. Stop the containment purge supply and exhaust fans.

b. Close the purge and exhaust fan MOVs.

c. Stop the containment air recirculation fans.

d. Start the containment iodine filtration fans.

(m**** CATEGORY 07 CONTINUED ON NEXT PAGE manas)

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

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 27

 ~~~~E 653E6616 E~6 NTR6E~~~~~~~~~~~~~~~~~~~~~~~~

GUESTION 7.03 (1.00) Given the following procedural steps, select the choice that provides the proper sequence for paralleling a diesel generator to a live bus.

1. Turn the synchronizer switch on.

2. Start the emergency diesel generator.

3. Turn off the synchronizer switch.

4. Operate the governor control switch so the synchroscope moves slowly in the fast direction.

5. Close the generator output breaker when the synchroscope needle passes the 11 o' clock position.

6. Operate the e:: citer control switch to match incoming voltage to running voltage.

a. 1,2,4,5,6,3

b. 2,1,6,4,5,3

c. 2,6,1,4,5,3

d. 1,2,6,5,4,3

e. 2,1,4,5,6,3 GUESTION 7.04 (1.00)

What is the criterion used to verify the RCS is solid when placing the RCS in a solid water condition per OP-3.4?

a. A high pressure spike occuring on the pressuri:er pressure instruments.

b. When spraying the pre'ssurizer no longer will decrease RCS pressure.

c. When indication of flow to the PRT through the pressuri:er PORVs is verified.

d. When an increase in letdown flow has been verified.

(msx*r CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx) _ - ~ _ _ _ _ - _ _ . , - --y,- __

7. PROCEDURES - NORMAL, ABNCRMAL, EMERGENCY AND PAGE 28

                                 ~
 ~~~~dddfdLddIddL dd TddL'~~~~~~~~~~~~~~~~~~~~~~~

QUESTIGN 7.05 (1.00) Which of the following is the PREFERRED method of providing a means of decay heat removal during a refueling outage with the reactor vessel head removed if all RHR pumps are lost?

a. Using natural circulation by feeding the SGs with AFW and bleeding them using SG blowdown.

b. By lining up the LHSI pumps to recircula'te flow through.the RWST and the reactor vessel.

c. By lining up the refueling purification system to recirculate water from the reactor cavity to the spent fuel pit using the spent fuel coolers for cooldown.

d. By feeding the RCS using a LHSI pump and bleeding it from a SG primary manway.

QUESTION 7.06 (1.00) Which of the following indications require SI reinitiation following a spurious SI that has been secured?

a. RCS pressure at 1950 psis.

b. RCS subcoolin3 at 35 degrees F.

c. Pressurizer level at 17%.

d. All steam generator levels at 15%.

QUESTION 7.07 (1.00) Prior to transferring rod control from MANUAL to AUTO during a reactor startup, Tave and Tref should be within ________ degree (s) F.

a. 1

b. 3.5 c .. 5

d. 10

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

     . - - -                                              .               f . ._   m - -   y-,,_

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29

   --- EI616E6655EL 56 TE6L'~~~~~~~~~~~~~~~~~~~~~~~

QUESTION 7.08 (1.00) How are the trip bistables of a failed power range detector placed in the trip condition?

a. Place the applicable bistable test switch in the ' Test' position in the Reactor Protection Cabinet.

b. Remove the applicable control and instru, ment power fuses on the power range drawers.

c. Place the applicable Power Hismatch Bypass switch to the failed position at the Miscellaneous Control and Indication' Panel.

d. Place the applicableJCneparator Channel Defeat switch to the failed channel pcs.4cion at the Detector Current Comparator Panel.

QUESTION 7.09 (1.00) During a natural circulation cooldown, what is the preferred method of reactor coolant system depressupi=ation?

a. Open the normal spray valve.

b. Open the auxiliary spray valve.

c. Open the pressurizer PORVs.

d. Open the reactor vessel vent valves.

(xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)

• t t

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30

                      ~

~~~~Ed616E6556IL 56NTR6E~~~~~~~~~~~~~~~~~~~~~~~~ QUESTION 7.10 (1.00) How do you eliminate a Hydro 3en buble that has formed in the reactor vesael?

a. Increase pressuri:er temperature above ccre thermocouple readings.

b. Inject oxygen into the reactor coolant system via the chemical and volume control system.

c. Maximize coolant flow by running all reactor coolant pumps, increase letdown flow to 120 spm, and place the cation bed demineralizer in service in parallel with the mixed bed deminerali:er.

d. Vent the reactor vessel head.

QUESTION 7.11 (1.00) Which of the following is a 10 CFR 20 exposure limit?

a. 5 rem / year-whole body.

b. 1 rem / quarter-whole body.

c. 18.75 rem / quarter-hands.

d. 7 rem / quarter-skin of whole body.

QUESTION 7.12 (1.00) If you are in a 100 mrad / hour samma field for 45 minutes, what is your dose in MREM after 45 minutes?

a. 45

b. 75

c. 450

d. 750 (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE *****)

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                                        +
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7. PROCEDURES - NORMAL, A E:N O R M A L , EMERGENCY AND PAGE 31

                                       ~~~~                         -        ~~~~~~~~~~~~~~~~~~~~~~~~

R BIBL53iE%L E5NT5DL QUESTION 7.13 (1.00) Which of the following radiation e::posures would inflict the greatest biological damage to man?

a. 1 Rem of GAMMA. ,

b. 1 Rem of ALPHA.

c. 1 Rem of NEUTRON. '

d. NONE of the abovei they are all equivalent.

QUESTION 7.14 (1.00) Which action below should the operator perform if criticality is NOT cchieved before the upper limit of the ECP is reached on a reactor startup?

a. Insert Bank D control rods to the bottom.

b. Stop control bank withdrawal and approach criticality via boron dilution.

c. Proceed with the startup utilizing an Inverse Count Rate Plot from the time the ECP opper limit is reached until the time criticality is reached.

d. Proceed with the s*.a" tup utilizing a pull-and-wait approach to criticality.

QUESTION 7.15 (1.00) When is control rod cooling air required to be operating?

a. Only if the control rod mechanism is energi:ed when the plant temperature is above 350 F.

b. Whenever the plant temperature is aoove 350 F, regardless rf the rod mechanism status.

c. During any plant cooldown evolution below 350 F.

d. Whenever the individual rod position indication system or any rod drive mechanism is energized.

(xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)

7. FROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32

                             ~            ~~~~~~~~~~~~~~~~~~~~~~~~
   ~~~~R 5656L 5fE5L E6 TR6L QUESTION        7 16               (1.00)

How is the baron concentration equali:ed between the pressuri:er and the rest of the reactor coolant system?

a. Charsi ng pure water or borated water to the PZR through the aux.

spray line as required.

b. Charging a blended flow equal to the Reactor Coolant System boron concentration into the Reactor Coolant system to force more of the RCS inventory up the surge line and into the PZR.

c. Taking MANUAL control of the Pressurizer Level Control System and then increasing and decreasing the pressurizer level as required in order to get an exchange between the RCS and PZR water inventory.

d. Operating'the pressuri:er heaters and spray valves as required.

QUESTION 7.17 (1.00)

 . The administrative differential temperature limit between the pressuri:er end its spray is ______ degrees F delta T.

a. 150

b. 200

c. 260

d. 320 QUESTION 7.18 (1.00)

It is required to exit a radiation area and report to HP whenever your 0-200 mr self-reading dosimeter reaches ______ mr.

a. 100

b. 125

c. 150 .

d. 175 (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx)

De

                                  -w-                                        + J.-  - , y-      e -

70 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33 l ~~~~E 655E6535 E~5dNT56E~~~~~~~~~~~~~~~~~~~~~~~~ QUESTION 7.19 (1.00) An RWP number is necessary to obtain which of the following?

a. A dose card.

b. A TLD.

c. A self-reading dosimeter.

d. Your current quarterly dose.

QUESTION 7.20 (1.00) During recovery from a refueling outage, what tool is utilized to relatch the RCC elements to their drive shafts?

a. The same tool used for unlatching the drive shafts.

b. A special drive shaft latching tool.

c. A new fuel handling tool.

d. A spent fuel handling tool.

QUESTION 7 21 (1.00) Which of the following actions constitutes entry into Mode 6?

a. Increasing the boron concentration to 2000 ppm.

b. Decreasing reactor coolant temperature to less than 140 F.

c. Detensioning the first reactor vessel head stud.

d. Decreasing keff to less than 0.95.

(xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx) h O

7o PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34

                        ~          ~~~~~~~~~~~~~~~~~~~~~~~~
 ~~~~5E655L655C L CUNTE5L QUESTION       7.22           (1.00)

If one of the circulating water pumps trips during a radioactive liquid waste release, which action below is required?

a. The release must be terminated immediately.

b. No action is required if there is at least one other circulating water pump running. .

c. Attempt to restart another circulating water pump. If this fails, then terminate the release.

d. Two circulating water pumps must be restarted for dilution in order to continue the release.

QUESTION 7 23 (1 00) A fire found in the 4160 switchgear is best extingui.hed with a fire extinSuisher marked with which symbol below?

a. A

b. B

c. C

d. 0 QUESTION 7.24 (1.00)

Which Gai-Tronics channel is reserved for emergencies?

a. channel 1.

b. channel 3.

c. channel 5.

d. channel 7.

(****r CATEGORY 07 CONTINUED ON NEXT PAGE *****) . e 4 a n

7. PROCEDURES - NORMAL, AE:N O R M AL , EMERGENCJ AND PAGE 35

                           -        ~~~~~~~~~~~~~~~~~~~~~~~~
   ~~~~R I656L6555 L 56NTR6L GUESTION        7.25          ( .50)
 !  For a reactor startup, it is required by the procedure to go critical while the power indication is still in the source range. TRUE or FALSE?

OUESTION 7.26 ( .50) All radioactive gaseous releases must be made by'a Licensed Senior Reactor Operator. TRUE or FALSE? 00ESTION 7.27 ( .50) Either the Health Physics Department OR the Operations Department may initiate a radioactive waste discharge record for a gaseous radwaste release. TRUE or FALSE? OUESTION 7 28 ( .50) Will adverse containment conditions cause affected control room indications to indicate HIGHER or LOWER than actual conditions? QUESTION 7.29 ( .50) A rod control logic cabinet internal failure can be verified in the control room. TRUE or FALSE? GUESTION 7.30 (1.00) What is the immediate make-up lineup upon loss of refueling cavity level during refueling? QUESTION 7.31 (1.00) What should the operator do about an SI signal if it e::ists during a loss of all AC power? (xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE xxxxx) . n e 1 i __ _I __ E -m . - .-

7o PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PACE 36

                            ~         ~~~~~~~~~~~~~~~~~~~~~~~~
   ~~~~5I555L5555EL 55NTR5L QUESTION        7.32            (1.00)

What two conditions constitute adverse containment conditions? QUESTION 7.33 (1 00) What are the required operator actions if both source range instruments fail when power is >P-6 and <P-10 during a power , increase to 100% power? GUESTION 7.34 (1.00) List the immediate operator actions for an Anticipated Transient Without Trip. QUESTION 7.35 (3.25) List the immediate operator actions for the Reactor Trip or Safety Injection procedure. Q$ESTION 7.36 (1.00) List the whole body administrative limits per calender quarter that can be achieved (a) without any additional approval and (b) with the highest level of approval. QUESTION 7.37 (1.00) List four of the critical conditions required to be recorded during a startup when 1X 10E-8 amps is attained. QUESTION. 7.38 (1.00) List five pieces of information an RWP will tell you what you will need to know in order to work in the applicable restricted area. (muzza CATEGORY 07 CONTINUED ON NEXT PAGE ***x*) 4 i

70 PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 37

    -~~ EI555L55iEIE C5siE5L QUESTION     7.39        (1.25)

List five surveillance items that must be done by operators to ensure Tech Spec requirements are met in the event of a loss of both units

     -PROD AC-250 Computer s .

QUESTION 7.40 (1.00) If instrument air is lost outside the containment'and the plant is solid, what are your required immediate actions? (xxxxx END OF CATEGORY 07 xxxxx) 4 OO l f ., l _. . . _ _ - _ . _ _ . _ _ _ _

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38 8. QUESTION 8.01 (1 00) In performing an Emergency Work Request outside r croal work hours, which cne of the fo11owin3 is NOT a Shift Supervisor function?

a. Deciding who will perform the emergency maintenance based on exposure records, work requests and manpower availability.

b. Attaching an Equipment History / Failure Analysis form to the Emergency Work Order.

c. Identifying the emergency item'. '

d. Entering the emergency iter.into the Work Planning and Tracking System.

e. Notification of the SRO on call and the craft foreman on shift.

QUESTION 8.02 (1.00) While in mode 3 on Unit'2 a planned boron dilution is completed at 0200. At 0230 boron dilution valves 2-CH-140 and 2-CH-160 are found to have been closed but not locked (not secured). Which one of the following is NOT a required action?

a. Suspend all operations involving positive' reactivity changes,

b. Lock, seal or otherwise secure the valves in the closed position.

c. Verify that the shutdown margin is greater than or equal to 1.77%

delta k/t.

d. Initiate and continue boration at greater than or equal to 10 spa of 20,000 ppm boron solution.

1 (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE **xxx) s

                                )
                                            /
                                                         .  ,                    1

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39 QUESTION 8.03 (1.00)

Which one (1) of the following is NOT a ' Limiting Safety System Settin3' trip setpoint?

a. Reactor coolant pump bus underfrequency '>' or '=* to 56.1 H.

b. Turbine trip - low trip system pressure '>' or '=' to 45 psis.

c. Fower range, neutron flux, high negative rate '<' or '=' to 5% of rated thermal power with a time constant,'>' or '=' to 2 seconds.

d. Low pressuri:er water level '>' or '=" to 18% of instrument span.

GUESTION 8.04 (1.00) Tech Spec 3.8.1.1 requires two separate and independent diesel generators. It further specifies certain diesel support equipment as e limiting-condition for operation. Which of the following is NOT part of that LCO?

a. Each with a separate 125-volt battery bank, charger and inverter.

b. Each with a separate day tank containing a minimum of 750 gallons of fuel.

c. A fuel storage system containing a minimum of 45,000 gallons of fuel.

d. A separate fuel transfer pump.

(*xxx* CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) _. ~_ .

i 80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 4C QUESTION 8.05 (1.00) If one overte*Perature delta T instrument channei ts INOPERABLE and placed in the trip condition, which statement below is correct?(In Mode 4,3 Loops)

a. Startup may proceed but NO mode change is permitted.

b. Startup may proceed but no mode changes beyond mode 2 are allowed.

c. Power operation may proceed until performance of the next required channel functional test. '

d. Fower operation may procee'd as long as the Quadrant Power Tilt Ratio is monitored at least once per 12 hours.

e. Startup and power operation may proceed if thermal power is restricted to less than or equal to 75% of rated thermal power.

QUESTION 8.06 (1.00) With one Unit 2 rod position indicator inoperable in mode le the non-indicating rod is moved from its last known position more than 24 steps. As an alternate to reducing powere the operator may determine that rods' position usin3 the movable incore detectors if done within a certain time interval after the last rod movement. What is that time interval?

a. immediately

b. 15 minutes

c. 30 minutes

d. 1 hour

e. 8 hours (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) t i

80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 41 GUESTION 8.07 (1.00) The Unit 1 boron injection tank will be INOPERABLE under which of the following conditions?

a. If it contains 950 gallons of appropriate water.

b. If it is borated to 25,000 ppe.

c. If its temperature is 160 des F.

d. If Hydrogen overpressure is 13 psig.

• QUESTION 8.08 (1.00)

Controlled leakaSe as defined by Unit 1 Technical Specifications refers to'

a. Letdown flow.

b. Liquid radwaste release flow.

c. Reactor coolant pump seal water flow.

d. Excess letdown flow.

e. S/G tube leakage.

QUESTION 8.09 (1.00) The Unit i reactor coolant system pressure exceeds 2735 psig when in mode 3. According to technical specifications, oressure must be restored within acceptable limits within what time frame given below?

a. 5 minutes.

b. 15 minutes.

c. 30 minutes.

        .d . one hour.

(xxxxx CATEGORY 03 CONTINUED ON NEXT PAGE xxxxx) t e

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                                                      +   --v.,   ,   e v --   . -- < ,      --e    e

80 ADMINISTRATIVE PRGCEDURES, CONDITIONS, AND LIMITATIONS PAGE 42 GUESTION 8.10 (1.00) The technical specification requiring the Unit 1 accumulator isolation valves be opened applies in which mode (s) listed below?

a. In mode 1 only.

b. In mode 1 only and only above 50% power.

c. In modes 1 and 2 only.

d. In modes 1,2 and 3 if pressurizer pressure is above 1000 psis.

QUESTION 8.11 (1.00) If control power is lost to a Unit 2 pressurizer power operated relief valve while in mode 1, which statement below is correct?

a. Tech specs require no action provided another PORV is operable and all pressuri:er code safety valves are operable.

b. Tech specs require the power supply to be removed from the associated block valve after verifying it to be open, if the PORV is not operable within 1 hour and continuous operation is desired.

c. Tech specs require the associated block valve to be shut and its power removed if the PORV is not made opercble within one hour and continuous operation is desirable.

d. Tech specs require action to be initiated within one hour to place the plant in at least hot standby within the following hour if the PORV is not made operable.

QUESTION 8.12 (1.00) What are the technical specification maximum heatup rates in any one hour period for the reactor coolant system (including the pressurizer)?

s. 100 degrees F for units 1 & 2.

b. 100 degrees F for unit 1 and 200 degrees F for unit 2.

c. 200 degrees F for unit 1 and 100 degrces F for unit 2.

d. 200 degrees F for units 1 & 2.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) v ,

I i 80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 43 l l l QUESTION 8 13 (1.00) If the lowest operating loop Tave drops below ______ in mode 1, Unit 1 &2 tech specs allow ______ to restore Tave within the limit before proceedin3 to place the unit in hot standby.

a. 541 degrees F, 30 minutes

b. 541 degrees F, 15 minutes

c. 547 degrees F, 30 minutes ,

d. 547 degrees F, 15 minutes OUESTION 8.14 (1.00)

Turbine overspeed protection is required by technical specifications because of which situation below?

a. Turbine components become missles which may penetrate the turbine casing & allow higher than allo ~wed off-site doses with an assumed maximum allowable fuel cladding and primary-to-secondary leak.

b. Turbine components become missles, which may damage cafety-related equipment.

c. The reactor thermal power exceeds limits in the unit's license.

d. The main steam trip valves exceed the closure time assumed in the accident analyses.

QUESTION 8.15 (1.00) Which action below is required when the fire supression spray / sprinkler system for the auxiliary building, component cooling pump area, elev. 244'6 is declared INOPERABLE?

a. Commence a unit shutdown for the applicable unit within one hour.

b. Establish an hourly fire watch patrol for the affected area with backup water fire supression equipment within one hour.

c. Establish a continuous fire watch for the affected area with backup water fire suppression equipment within one hour.

d. Log ar.bient temper atur e readings for the affected area hourly anc route an equivalent capacity fire hose to the area.

(***** CATEGORY OS CONTINUED ON NEXT PAGE *****) D

                                           - ~ , ,             - - -

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 44 )

l QUESTION 8 16 (1.00) Which one of the following is CORRECT concerning Unit 1 tech spec treatment of Fq(:)?

a. Since Fq(:) is not measursable, limits are placed on AFD.

b. Measureable limits are placed on Fq(:) only above 50% rated thermal power.

c. The limit placed on Fq(:) by tech specs varies as a function of core height.

d. The limit placed on Fq(:) by tech specs varies as a function of region average burnup.

QUESTION 8.17 (1.00) Unit 2 tech spec 3.-1.3.5 requires all shutdown rods to be fully withdrawn. Which one of the following is CORRECT concerning this specification?

a. It is applicable in modes 1,2 and 3 only.

b. Fully withdrawing all but one of the s'hutdown rods that were

. partially inserted will satisfy the action statement.

c. If one rod is not and~cannot be fully withdrawn it must be declared inoperable.

d. The actions required by the action statement must be satisfied in 15 minutes or less.

QUESTION 8.18 (1.00) A Unit 2 control rod is determined to be INOPERABLE in mode 2 as a result of excessive friction. Tech Specs require which action below in 1 hour?

a. Be in hot standby.

b. Restore the rod to operable status.

c. Position the remainder of the rods in that group to within

              '+' or    12 steps of the inoperable rod.

d. Determine that the tech spec shutdown margin requirement is satisfied.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)

80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LINITATIONS PAGE 43 QUESTION 8.19 (1.00) A quarterly tech spec surveillance requirement may be extended up to ______ days without declaring the component inoperable due to the surveillance testing not being performed.

a. 9

b. 23

c. 32 ,

d. 41 QUESTION 8.20 (1.00)

ADM 5.7 entitled ' Correcting Data on Completed Procedures' makes which of the following provisions?

a. It is permissible to make corrections to data which has been recorded. This must be performed in the following manner: draw a line through the erroneous data and record correct data. Initial and date the correction.

b. The cognizant supervisor shall list any available substantiating information.

c. The individual who conducted the test and the co3ni: ant supervisor shall attest to the corrected data by affixing their signature and the date.

d. The station nuclear safety and operating committee shall review the circumstances and the chairman shall sign and date the form when their review is completed.

(xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) i 6 I l

So ADMINISTRATIVE PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE 46 OUESTION 8.21 (1.00) According to Unit 2 tech specs, which of the following statements regarding i the site fire brigade is correct?

a. Shall be composed of at least 10 members.

b. Shall not include any of the minimum shift crew required by table l 6.2-1 (minimum shift manning).

c. Shall be responsible for the control room

• command function until the fire emergency is secured.

d. Hay NOT be less than the minimum requirement even to accomodate unexpected absences.

QUESTION 8.22 (1.00) Which of the following is true of tagging of systems and/or components according to ADM 14.0?

a. Special order tags, form 625.6 are grey tags.

b. The operator shall be responsible for placing all tags.

c. Maintenance personnel to whom the ta99ing record is issued will independently review the tag selection and authorize the operator to place the tags.

d. When more than one department is to work on the same equipment, a separate tag will be placed for each department unless a fifteen minute headway tas has been authorized.'

(***** CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) t 9

                                                          ,ya a e-- -   ,   t  -
                                                                                      ++-- -   -- --

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 47 QUESTION 8.23 (1.00)

Which one of the following statements is true of jumpers as controlled by ADM 14.1, ' Jumpers (temporary modifications)'?

a. Temporary hose connections necessary to perform a test are defined as cor.stituting a jumper if their use is described in the text or drawings of the FSAR.

b. In order to. control their use, ADM 14.1 only permits the use of jumpers as described in either an MOP or' an MMP.

c. Jumpers not controlled by any approved procedure shall only be used with the shift supervisor's prior knowledge and approval.

d. For those jumpers that are installed by an approved procedure, a jumper los form shall be initiated and all pertinent data associated with the installation recorded.

QUESTION 8.24 (1.00) Assume both Units are in cold shutdown for refueling and maintenance. Incore instruments have been withdrawn to their storage position and the incore instrumentation thimble retraction has just been completed. All access doors to the reactor cavity are locked. Flooding of the refueling cavity in preparation for refueling was started six hours ago. It has just been determined that the water level in the refueling cavity is decreasing slowly. Assume you are the senior licensed SRO on shift. You have decided to send an au::ilaiary operator into the reactor cavity in an effort to locste the source of leakage. Which of the following would be the anticipated dose rate range in the cavity?

a. < 30 mr/hr.

b. 30-300 mr/hr.

c. 3-30 R/hr.

d. 300-3,000 R/hr.

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

1 I i o

                                                                                                      ,4 .-

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 48 QUESTION 8.25 (1.00)

Which one of the following is a condition requiring stoppa3e of all work and immediate evacuation of containment accordin3 to the precautions and limitations in 1-OP-4.1, ' Controlling Procedure for Refueling *?

a. The 'hi f lu:: at shutdown' alarm is actuated during fuel movement.

b. Loss of audible neutron count rate (< two tones per minute) after offloading 3/4 of the reactor core.

c. The station evacuation ala'rm sounds.

d. Declaration of an ' alert'.

QUESTION 8.26 (1.00) Which of the following conditions for refueling must be met if refueling is to continue?

a. The flow rate of reactor coolant through the reactor coolant system shall be '>' or '=" to 3000 spa whenever a reduction in RCS baron concentration is being made.

b. Valve 1-CH-217 (PG to blender isolation) shall be locked in the open position to assure an adequate supply of borated water in the event emergency boration is required.

c. The following baron injection flow paths shall be operable (1) a flow path from the boric acid tanks via a boric acid transfer pump through a' charging pump to the RCS, and (2) the flow path from the refueling water storage tank via a charging pump to the RCS.

d. Two boric acid transfer pumps shall be operable if neither.

charging pump is operable. (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx), I l i _ _ f __.

8. ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 49 GUESTION 8.27 (1.00)

Which of the following is the principal candidate for relief of the Interim Station Emergency Manager in the event of an emergency at the station.

a. Assistant station manager (ops)

b. Superintendent of operations

c. Station manager

d. Manager nuclear operations and maintenance.

QUESTION 8.28 (1.00) Which of the following require activation of both the TSC and OSC?

a. Either an unusual event, alert, site area emergency or general emergency.,

b. Only an alert, site area emer3ency, or general emer3ency.

c. Only a site area emergency or general emergency.

d. Only a general emergency.

QUESTION 8.29 (1.00) If an operator is returning to shift after a three week vacation, he is required to read a'nd initial the logs for the previous ______. (choose one)

a. 1 day

b. 7 days

c. 14 days

d. 21 days (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx)

O l t l i

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                                                                     ,f- , - . _ .   - ,         .,

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 50 80 QUESTION 8.30 (1.00) Following a reactor trip where the reason for the trip is clearly understood and corrected and no significant malfunctions of safety-related or important equipment occurred, who is responsible for making the decision to restart the reactor?

a. Shift supervisor.

b. Superintendent of operations (or SRO on call).

c. Assistant station manager'(S & L). .

d. Station manager.

QUESTION 8.31 (1.00) Match the following tech spec defined leakage with the associated leak rate limit. Assume normal operating reactor coolant system temperature and pressure. The limits may be used more than once. Leakage Leak rate limit a) Pressure boundary leakaSe. 1) 0 SP* b) Controlled leakage. 2) 1 spa c) Identified leakage. 3) 5 spm d) Unidentified leakage. 4) 10 spa e) Primary-to-secondary leakage 5) 30 spa through one SG. 6) 50 sph

7) 500 spd GUESTION 8.32 (1.00)

To prevent entering a technical specification action statement on Unit 1, the quadrant power tilt ratio shall not exceed ______ when reactor power is above 50%. If this limit is exceeded, then an extended temporary QPTR limit of ______ is allowed during efforts to restore GPTR to normal levels. (xxxxx CATEGORY 08 CONTINUED ON NEXT PAGE xxxxx) S l E_.___. .__

                                                                                '    8

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 51 QUESTION 8.33 (1.00)

Complete the following statements concerning refueling operations by filling in the correct number. a) At least ______ feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. b) During movement of irradiated fuel in the reactor pressure vessel the - reactor shall have been suberitical for at least ______ hours. QUESTION 8.34 (1.00) Provide the minimum number of individuals required by tech specs for the following positions to operate both units at full power. a) ______ shift supervisor (s) b) ______ SR0(s).(NOT including the shift supervisor (s)) c) ______ R0(s) d) ______ A0(s) e) ______ STA(s) QUESTION 8.35 (2.00) Fill in the blanks below: With the shutdown margin less than ______%, immediately initiate and continue boration at ____spm or greater of ____ ppm boric acid solution or equivalent until _______________________. uuESTION 8.36 (2.00) ADM 19.3 (shift conduct, relieving the shift) states, 'the shift supervisor shall, perform the following:'. (Supply the six items listed in the procedure for the unit 1 SRO to REVIEW and SIGN at the beginning of the shift) (***** CATEGORY 08 C0r4TINUED ON NEXT PAGE *****), e p ,. ., , , ,,_e. . _

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 52 GUESTION 8 37 (1.00)

When a system, subsystem, train, component or device i s determined to be inoperable solely because its emergency power source i s 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 two conditions are met. List these TWO conditions. d QUESTION 8.38 (1.00) ADM 20.9 indicates that entry into containment during reactor cperation exposes personnel to four distinct hazards. List these FOUR hazards. (xxxxx END OF CATEGORY 08 xxxxx) (************* END OF EXAMINATION xxxxxxxxxxxxxxx) t

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 53 gg gggggg--------------------------------------

           -ANSWERS -- NORTH ANNA 182                               -85/10/15-DEAN, WH ANSWER            5.01            (1.00) c REFERENCE General Physics, HT & FF, Section 3.2 WBN, HT A FF,          p. 13              .,                               ,

002/000-K5.01 (3.1/3.4) ANSWER 5.02 (1.00) d REFERENCE Surry, TS 3.12-16 and 17 HBR, TS, p. 3.10-13 NAPS, TS, p. B 3/4 2-4 ANSWER 5.03 (1.00)

           'c REFERENCE General Physics, HT & FF, pp. 239 and 240 ANSWER            5.04            (1.00) a REFERENCE DPC, FNRE, pp. 200 and 221 ANSWER            5.05            (1.00) c REFERENCE NUS, Vol        4,  pp G-8 9,
                                                 - . - -      . - -   . , ~ . , - . - . -   -   , , .  - , - . . _ -

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 54 q q g g g g - - - - - - - - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - -

ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WM ANSWER 5.06 (1 00) d REFERENCE NUS, Nuclear Energy Training - Reactor Operation VEGP, Training Text, Vol. 9, Ch. 21 DPC, Fundamentals of Nuclear Reactor Engineering' 001/010-K5.13 ANSWER 5.07 (1.00) b REFERENCE NUS, Nuclear Energy Training - Reactor Operation 4 ANSWER 5.08 (1.00) D REFERENCE NUS, Nuclear Energy Training - Reactor Operation, p. 9.2-5 ANSWER 5.09 (1.00) d , REFERENCE General Physics, Heat Transfer Thermodynamics and Fluid Flow, PP. 145 - 148 ANSWER 5.10 (1.00) d REFERENCE NUS, Nuclear Energy Training - Plant Performance, pp. 10-1.8 & 9 t 4 9

                                               *eo*

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 55 g gggqq7 g--------------------------------------

ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WH ANSWER 5.11 (1.00) d REFERENCE Westinghouse Reactor Physics, pp. I-5.63 - 76 HBR, Reactor Theory, Sessions 38 and 39 DPC, Fundamentals of Nuclear R/ actor Engineering,'Section VI 001/000-K5.39 (3.5/4.1) ANSWER 5.12 (1.00) b REFERENCE General Physics, HT & FF, p. 229 002/000-K5.01 (3.1/3.4) ANSWER 5.13 (1.00) L. REFERENCE Westinghcuse Reactor Physics, pp. I-5.26 & 27 ANSWER 5.14 (1.00) d REFERENCE Westinghouse Reactor Rhysics, Section I-5, HTC and Power Defect DPC, Fundamentals of Nuclear Reactor Engineering 002/000-K5.02 (3.3/3.6) t MO l 4M

• I

50 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 56 THERMODYNAMICS ANSWERS -- NORTH ANNA 112 -85/10/15-DEAN, WM ANSWER 5.15 (1.00) c REFERENCE Westinghouse Reactor Physics, pp. I-5.2 - 16 o , ANSWER 5.16 (1.00) b REFERENCE WBN, TS, p. B 3/4 1-1 HER, TS, p. 3.10-10 CAT, TS, p. B 3/4 1-1 NAPS, TS, p. B 3/4 1-1 ANSWER 5.17 (1.00) c REFERENCE General Physics, HTFF - Fluid Flow Applications for Systems and Components ANSWER 5.18 (1.00) b REFERENCE WBN, TS, p. B 3/4 2-1 HBR, TS, p. 3.10-17 NAPS, TS, p. B 3/4 2-5 ANSWER 5.19 (1.00) b s N

I.

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 57

                     - - g g--------------------------------------

ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WH REFERENCE HBR, Reactor Theory, Session 42, pp. 4 & 10 ANSWER 5.20 (1.00) C REFERENCE e HBR, Reactor Theory, Sessions 41 and 42 DPC, Fundamentals of Nuclear Reactor Engineering, pp. 121 and 122 004/000-K5.08 (2.6/3.2) , ANSWER 5.21 (1.00) e REFERENCE NUS, Reactor Theory ANSWER 5.22 (1.00) c REFERENCE NUS, Reactor Theory ANSWER 5.23 (1.00) a REFERENCE DPC, Fundamentals of Nuclear Reactor Engineering, p. 138 001/000-K5.02 (2.9/3.4) ANSWER 5.24 (1.00) b t 9 N

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G. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 58 g ggg g g-------------------------------------- ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WH REFERENCE General Physics, HT & FF, p. 320 ANSWER 5.25 (1.00) C REFERENCE ., , DPC, Fundamentals of Nuclear Reactor Engineering, p. 170 001/000-K5.13 (3.7/4.0) ANSWER 5.26 (1.00) b REFERENCE Westinghouse Reactor Physics, Sects 3 and 5 NUS, Nuclear Energy Training, Units 6 and 8 NA OP 1.4 ANSWER 5.27 (1.00) a REFERENCE Westinghouse Nuclear Training Operations ANSWER 5.28 (1.00) b REFERENCE NAPS, Instructor Guide NCRODP-95.2, Hitigating Core Damage ANSWER 5.29 (1.00) b t

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 59 y gg g--------------------------------------

ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WH REFERENCE NAPS, Instructor Guide NCRODP-95.2, Mitigating Core Damage l ANSWER 5.30 (1.50)

a. FALSE (0.5) i

b. FALSE ,,

(0.5)

c. FALSE (0.5)

REFERENCE Westinghouse Nuclear Training Operations, pp. I-3.9 - 15 ANSWER 5.31 (1.00) d - REFERENCE Westinghouse Nuclear Training Operations, p. I-5.76 . ANSWER 5.32 (1.00) c REFERENCE Westinghouse Nuclear Training Operations, pp. I-4.19 ,- 21 ANSWER 5.33 (1.00) s REFERENCE General Physics, HT & FF, Chapter 1 < 4 h

5. THEORY OF NUCLEAR POWER PLANT' OPERATION. FLUIDS, AND PAGE 60 y gggg--------------------------------------

ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WH ANSWER 5.34 (2 00)

a. REMAIN THE SAME (0.5)

b. DECREASE (0.5)

c. INCREASE (0.5)

d. DECREASE (0.5)

                                           .e                     ,

REFERENCE Steam Tables 010/000-K5.02 (2.6/3.0) ANSWER 5.35 (2.00)

a. DECREASE (0.5)

b. INCREASE (0.5)

c. INCREASE (0.5)

d. DECREASE (0.5)

REFERENCE General Physics, HT & FF, pp. 319 - 334 ANSWER 5.36 (2.00)

a. TRUE (0.5)

b. TRUE (0.5)

c. FALSE (0.5)

d. TRUE (0.5)

REFERENCE NUS, Nuclear Energy Training - Plant Performance, p. 5-3.2 General Physics, HT & FF, pp. 155, 319, and 320 VEGP, Guestion Bank, 4752-a and 4 336 t , l 1 l

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5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 61 ANSWERS -- N' ORTH ANNA 1&2 -85/10/15-DEAN, WM ANSWER 5.37 ( .50)

TRUE REFERENCE General Physics, HT & FF, p. 176 002/000-K5.01 e , 4 ( I r

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s 60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 62 ANSWERS -- NORTH ANNA 1&2 -85/10/15-0EAN, WM I ANSWER 6.01 (1.00) c REFERENCE NA NCRODP 91.1, 'ESF' ANSWER 6.02 (1.00) c REFERENCE NA NCRODP 88.1, 'RCS-PZR and Press. Relief' ANSWER 6.03 (1.00) d REFERENCE NA NCRODP 91.2, ' Containment and Containment Systems' ANSWER 6.04 (1.00) d REFERENCE FNP, RHR Lesson Plan, pp. 8&9 NA NCRODP 80.2, 'RHR System' 005/000-K4.07 (3.2/3.5) ANSWER 6.05 (1.00) c REFERENCE Farley SD, 'CVCS', pp 13 NA NCRODP 88.3, 'CVCS' t

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 63 f.NSWERS -- NORTH ANNA 112 -85/10/15-DEAN, WH ANSWER 6.06 (1.00) c GML b REFERENCE NA NCRODP 91 1, 'ESF-Recire Spray System' ANSWER 6.07 (1.00) ,, ,

o REFERENCE NA NCRODP 92.2, " Service Water System' ANSWER 6.08 (1.00) e REFERENCE NA NCRODP 88.1, 'RCS-PZR and Press Relief' ANSWER 6.09 (1.00) i 3 \ REFERENCE NA NCRODP 89.4, 'Feedwater Systems-Main Feed' s AN 6.10 .00) REF NCE s NCRODP 88.1, 'RCS-Rea r Vessel / Core Construction

• ANSWER 6.11 (1.00) -

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 64 ANSWERS -- NORTH ANNA 112 -85/10/15-DEAN, WH  ;

J i REFERENCE NA NCRODP 88 1, 'RCS-PZR and Press. Relief  ; 6 12 i ANSWER (1.00)  ; L d r i REFERENCE ' NA NCRODP 92.2, " Service Water,, System' , i ANSWER 6.13 (1.00)  ; c , REFERENCE  ! NCRODP 90.1, ' Basic Electrical Distribution *

    -ANSWER          6.14       (2.00)                                                       j a)   No effect    (+.5 ea)                                                           .

b) Art only f - c) Arm alid Actuate - d) No effect REFERENCE i NA NCRODP 93.11, ' Steam Dumps

• i a

ANSWER 6.15 (1.50) a) 1-FP-P'-2. 0+.5 ea)  ! b) 1-FP-P-1 c) Both  ; REFERENCE { NCRODP 92.1, ' Fire Protection' t N . i n r I 4W

l

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 65 l 1

ANSWERS -- NORTH ANNA 1&2 -85/10/15-DEAN, WM i ANSWER 6.16 (1.50) a) False b) False c) True REFERENCE NA NCRODP 93.2, "Excore NIS' pp 14-22 1 o . . ANSWER 6.17 (1.50) a) Not allowed (+.5 ea) b) Allowed c) Allowed REFERENCE NA NCRODP 91.2, ' Containment Isolation' ANSWEP 6.18 (1.50) a) DCI (+.5 ea)

b) DCI c) CRE REFERENCE NCRODP 90.4, 'EOG' Handout H-2.4 ANSWER 6.19 (1.50) a) 3 (+.5 ea) b) 4 c) 2 REFERENCE NA NCRODP 91.2, ' Containment Atmospheric Cleanup System' t

e 9 1 1

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 66 ANSWERS -- NORTH ANNA 1&2 -85/10/15-DEAN, WH ANSWER 6.20 (1.50) e) 3 (+.5 ea) b) 1 c) 4 REFERENCE NCRODP 89.1, ' Main Steam System'

                                         .i                   ,

ANSWER 6.21 (1.50)

       -S/G Press used for density compensation of steam flow, so if pressure fails high, steam flow' input to the SGWLCS will increase (+1.0)
       -Ws>Wf will cause FWRV to open, level will increase until level error counteracts the steam flow / feed flow mismatch (+.5)

REFERENCE NA NCRODP 93.12, 'SGWLC and Protection' ANSWER 6.22 (2.00) MOV-100C (+.25 ea) MOV-100A PCV-1008 PCV-100A 155 149 166 190 See Attached Drawing REFERENCE NA NCRODP 89.4, 'Feedwater Systems-AFW' ANSWER 6.23 (1.50)

       -RHR System Inlet       (+.5 ea)
       -Loop Drain Lines
       -Differential Pressure taps for RCS flow indication t
   +

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6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 67 ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WH REFERENCE NA NCRODP 88.1, 'RCS-Piping and Instrumentation' ANSWER 6.24 (1.00)

  -Lower Delta P (8' vs 12-14' on Uni             1)    ( + . 5 41d"
    =rr'ua. 3 un sein sur u.. . t 2       cc REFERENCE NA NCRODP 89.1, ' Main Stan Sysien* -

• ANSWER 6.25 (1.75)

1) Reference Si3nal for comparison with actual steam flow (+.25 ea)

2) P-13

3) C-7

4) Feed water ces valve

5) C-5

6) Rod Control

7) Tref for comparison w/Tavs REFEREN2E NCRODP 89.1, ' Main Steam System' ANSWER 6.26 (1.00)

Prevents voltage drops during the diesel load sequence (+.5) from causing subsequent load shedding (+.5) negating the loading of the diesel. REFERENCE NCRODP 90.4, 'EDG' ANSWER 6.27 (1.25) , 40% setpoint from 0-20% (+.5) Turbine Power (+.25) and linearly from 40-110% as Turbine Power goes fron. 20-100% (+.5) REFERENCE NCRODP 91.1, 'ESF-SI or ECCS' 1 1 s I _ _,

6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 68 ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WH ANSWER 6.28 (1 50) f>f

1) Auto Heat Actuated >190 d%grees F (+.5)

                    ~

2) Manual release on* Control Room Fire Panel (+.25 ea)

3) Deluge Panel in Track Bay

           *          *

• Break Glass and Pull station next to the facility 4)

5) Deluge Valve REFERENCE s NCRODP 92.1, ' Fire Protection' ANSWER 6.29 (1.00)

    -Loop stop/ Bypass isolation valves positioned to allow flow                             (+.25 ea)
     -Bearing lift pump running with >700 psis discharge
     -no electrical faults on RCP motor
     -locked rotor protection speed switch satisfied REFERENCE NA NCRODP 88.1, 'RCS-RCPs' ANSWER        6.30           (1.50)
     -Allows Unit 1 to have its normal station service busses supplied from its normal station service transformers backfed from 500 KV switchyard for most Unit 1 trips.          (+1.0)
     -This reduces probability of combined loading from both Units normal and emergency buses on the reserve station transformer (+.5)

REFERENCE NA NCRODP 90.1, ' Basic Electrical Distribution' ANSWER. 6.31 (2.00) a) 2 out of 3; 90%; 56 seconds (+.25 ea) b) -diesel starts (+.15 ea correct response, .1 ca for correct order)

         -emergency. bus isolated from RS nfrmr
         -all loads shedded e:: cept Chg pumps, LHSI pumps and 480 VAC loads
         -output bkr shuts when diesel up to speed ano voltage
         -loads sequenced onto bus                                                         .

_ _ _ _ .___.l, ._ , -_ - -.

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4 do PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 69 CNSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WM  ; REFERENCE  ; NA NCRODP 90.4e 'Eners Diesel Gener.ator' i l

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s 70 PROCEDURES - NORMAL, ABNORM)L, EMERGENCY AND PAGE 70

   ~~~~EA6i6L655EIE EdATE5L'~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTN ANNA 182 -85/10/15-DEAN, WH ANSWER 7.01 (1.00) b. REFERENCE NAPS 1-AP-4, pp. 388.

                                           .i                   ,

ANSWER 7'.02 (1.00) c. REFERENCE NAPS 1-AP-5.1, pp.8-9. ANSWER 7.03 (1.00) b. REFERENCE NAPS, 1-OP-6.2, p. 7. ANSWER 7.04 (1.00) d. REFERENCE NAPS, 1-OP-3.4, p. 13. ANSWER 7.05 (1.00) c. REFERENCE NAPS, 1-AP-11, p.9.

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 71 RADIOLOGICAL CONTROL ANSWERS -- NORTH ANNA 1&2 -85/10/15-DEAN, WM ANSWER 7.06 (1.00) b.

REFERENCE NAPS 1-ES-0.2, Foldout page.

                                               .o                               ,

ANSWER 7.07 (1.00) D REFERENCE MNS OP/1/A/6100/01 p. 19. CNS OP/1/A/6100/01, Encl. 4.1, 2.95.10. NAPS 1-OP-2.1, p.8. ANSWER 7.08 (1.00) b REFERENCE MNS AP/2/A/5500/16, Case IV, p.9. CNS AP/1/A/5500/16, Case IV, p.14. NAPS 1-OP-4, p.8. ANSWER 7.09 (1.00) b REFERENCE MNS EP/2/A/5000/1.1, p.5. CNS EP/1/A/5000/1A1, p.4. NAPS 1-ES-0.3, p.5. ANSWER 7.10 (1.00) d t l I

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 72

     ~~~~R d6f6E66fCdE~C6 TR6E~~~~~~~~~~~~~~~~~~~~~~~~

ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WM REFERENCE MNS EP/2/A/5000/16.3 CNS EP/1//A/5000/2F3, p.7. NAPS 1-FRP-I.3A, p.3. ANSWER 7.11 (1.00) e REFERENCE 10 CFR 20.101 ANSWER 7.12 (1.00) b QF=1 for gamma 100(45/60)(1)=75 REFERENCE 10 CFR 20. ANSWER 7.13 (1.00) d REFERENCE 10 CFR 20. ANSWER 7.14 (1.00) a REFERENCE NAPS, 1-OP-1.5, p.10. t 1 4 me l t ~

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 73 RADIOLOGICAL CONTROL ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WH ANSWER 7.15 (1.00) b RE'FERENCE NAPS 1-OP-1.1.

                                                    .i                                  ,

ANSWER 7.16 (1.00) d REFERENCE NAPS 1-0P-8.3, p.5. ANSWER 7.17 '(1.00) c REFERENCE NAPS 1-OP-1.4, Attachment 1. ANSWER 7.18 (1.00) c REFERENCE Vepco GET, p.21. ANSWER 7.19 (1.00) c REFERENCE Vepco'GET, p.41. t we

7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 74

 ~~~ ~Rd6f 6E66f 6d[~66kT R 6E ~ ~~~ ~~~~~~~~ ~~~~ ~~ ~~~ ~~~

ANSWERS -- NORTN ANNA 1&2 -85/10/15-DEAN, WM ANSWER 7.20 (1.00) e REFERENCE NAPS 1-OP-4.1, p.41. a , A N'3 W ER 7.21 (1.00) c REFERENCE NAPS 1-OP-4.1, p.23. ANSWER 7.22 (1.00) B REFERENCE NAPS 1-0P-22.11, p.4. ANSWER 7.23 (1.00) c REFERENCE Vepco GET, p.49. ANSWER 7.24 (1.00) C REFERENCE Vepco GET, p.49.

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 75 RADIOLOGICAL CONTROL ANSWERS -- NORTH ANNA 112 -85/10/15-DEAN, WH ANSWER 7.25 ( .50)

FALSE. REFERENCE NAPS 1-OP-1.5, p.10.

                                                    .o                                   ,

ANSWER 7.26 ( .50) FALSE. REFERENCE NAPS 1-OP-23.2, p.4. ANSWER 7.27 ( .50) TRUE. REFERENCE NAPS 1-OP-23.2, p.3. ANSWER 7.28 ( .50) Hi 3 h er. REFERENCE NAPS 1-EP-0 Foldout page. ANSWER 7.29 ( .50) FALSE. (urgent failure alarm could be from logic or power cabinet failure) REFERENCE NAPS 1-AR-1, 1A-4. i i l me

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 76

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ANSWERS -- NORTH ANNA 1A2 -85/10/15-DEAN, WH ANSWER 7.30 (1.00) From RWST through LPSH into hot legs. REFERENCE NAPS 1-AP-52, p.3. ANSWER 7.31 (1.00) Roset'it. REFERENCE NAPS 1-ECA-0.1, p.2. ANSWER 7.32 (1.00) G 0.5 points each:

1. Containment pressure at 19.7 psia.

2. Containment rad. levels at 40,000 R/hr.

REFERENCE NAPS 1-EP-0, p.2. ANSWER 7.33 (1.00) Place both level trip switches in BYPASS. (Ops may continue.) REFERENCE NAPS 1-AP-4, p.4. ANSWER 7.34 (1.00) For 0.25 points each.

1. Attempt a reactor trip using both manual trip switches.

2. Attempt a turbine trip using both manual trip buttons.

       . Verify all AFW pumps are running.

4. Verify AFW flow to all SGs.

f M$ny 6(.arrTft

                                                                    -]   .        -

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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 77 l 55siE5C------------------------

 --- EK5i5C55fCIC ANSWERS -- NORTN ANNA 142                       -85/10/15-DEAN, WH REFERENCE NAPS 1-ECA-1, pp. 2&3.

ANSWER 7.35 (3.25) For 0.25 points each.

1. Verify reactor trip.

2. Verify turbine trip.

3. Verify AC emergency busses enersi=ed.

4. Check if SI actuated.

5. Check charging /SI pump valve lineup.

6. Verify FW isolation.

7. Verify t.FW flow.

8. Verify charging /SI flow.

9. Verify LHSI pumps running.

10. Verify containment phase A isolation.

11. Verify service water pumps running.

12. Verify RCS heat removal.

13. Check containment pressure.

REFERENCE NAPS 1-EP-0,pp. 1-5. ANSWER 7.36 (1.00) For 0.5 points each.- (a) 750 mr/qtr (b) 2750 mr/qtr REFERENCE NAPS Radiation Protection Manual, p.2.3-7. ANSWER 7.37 (1.00) Any 4 G 0.25 points each

1. Bank C position.

2. Bank D position.

3. Auct. High Tavs.

4. IR N35. .

5. IR N36.

6. RCS boron concentration. .

S W _ __. - _ - . -p , ,, _ . . . . , . _ _ _ , . - ~ _ , - . , .-.

70 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 78

   ~~~~

Rd6fEEEEICAE CEETREE~~~~~~~~~~~~~~~~~~~~~~~~ ANSWERS -- NORTN ANNA 182 -85/10/15-DEAN, WH REFERENCE NAPS 1.0P-1.5, p.12. ANSWER 7.38 (1.00) Any 5 9 0.2 points each

1. Job location. .

2. Brief job description.

3. Radiological conditions at job site.

4. Anti C requirements.

5. Respiratory requirements.

6.. Dosimetry requirements.

7. Necessary HP coverage.

8. Any special instructions.

REFERENCE Vepco GET, p.41

 . ANSWER         7.39             (1.25)

Any 5 9 0.25 points each'

1. IRPI within demand position limit. '?. h/ I#

2. Each IRPI operable.

3. AFD within limits.

p* p/ g d , q

4. Containment temperature.

5. Control room temperature.

6. Cire. water temperature.

REFERENCE NAPS 1-AP-42. ANSWER 7.40 (1.00) 0 0.25 points each:

1. Start all available service air compressors.

2. Start all available instrument air compressors.

3. Secure reactor coolant pumps.

4. Secure chargins Pump. .

REFERENCE NAPS 1-AP-28.1, p.3. , h -- - . _ . . - , . . - ,

80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 79 ANSWERS -- NORTH ANNA 1&2 -85/10/15-DEAN, WH l A 8.01 (1.00) (a) REFERENCE ' NA ADM i , pp 5 P  : Responsibilities during Mairstenance (3.5/3.9)

                                         .o                       ,

ANSWER 8.02 (1.00) (d) REFERENCE NA U2 TS 3.1.1.3 004.020; PWG-5 (2.9/4.1) ANSWER 8.03 (1.00) (d) REFERENCE NA U2 TS table 2.2-1. PWG-5! Generic TS Knowledge (2.9/3.9) ANSWER 8.04 (1.00) (a) REFERENCE NA U2 TS 3.8.1.1 064/050; PWG-5 (3.1/4.1) t es e a

8. ADNINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 80 ANSWERS -- NORTH ANNA 142 -85/10/15-DEAN, WM ANSW 8.05 (1.00)

(c) REFERENCE NA TS 3. . 01' 003 PNG-5 (3.2/4.0) ANSWER 8.06 (1.00) (a) REFERENCE NA U2 TS 3.1.3.2 - 014/000; PNG-5 (2.9/3.9) ANSWER 8.07 (1.00) (b).)(c) REFERENCE NA U1 TS 3.5.4.1 006/050; PWG-5 (3.2/4.3) ANSWER 8.08 (1.00) (c) REFERENCE NA U1 TS 1.7 004/020; PNG-5 (2.9/4.1) l j 1

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80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 81 ANSWERS -- NORTH ANNA la2 -85/10/15-DEAN, WH ANSWER 8.09 (1.00) (a) REFERENCE NA U1 TS 2.1.2 010/0008 PNG-5 (2'9/4.1) ANSWER 8.10 (1.00)

     ~

(d) REFERENCE NA U1 TS 3.5.1 006/020; K4.04 (3.8/4.2) ANSWER 8.11 (1.00) (c) REFERENCE NA U2 TS 3.4.4 0102000; A2.03 (4.1/4.2) ANSWER 8.12 (1.00) (a) REFERENCE N4 Ul&2 TS 3.4.9.1 002/020; PWG-5 (2.9/4.1) (

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So ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 82 ANSWERS -- NORTH ANNA 1&2 -85/10/15-DEAN, WH l

                                                                                              )

ANSWER 8.13 (1.00) (b) REFERENCE NA U1 & 2 TS 3.1.1.5 002/020; PWG-5 (2.9/4.1) ANSWER 8.14 (1.00) (b) REFERENCE NA TS bases 3/4.7.1.6 and 7 045/050; PWG-5 (2.7/3.5) ANSWER 8.15 (1.00) (c) REFERENCE NA TS 3.7.14.6 086/000i PNG-36 (2.8/3.7) , 1 ANSWER 8.16 (1.00) (c) REFERENCE NA U1 TS 3.2.2 001/000; K5.46 (2.3/3.6) 6

                                                                         \

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 83 ANSWERS -- NORTH ANNA 182 -85/10/15-DEAN, WM ANSWER 8 17 (1 00)

(c) REFERENCE NA U2 TS 3.1 3.5 001/050; PWG-5 (2 9/4.3)

                                                          .e                                                    ,

ANSWER 8.18 (1.00) (d) REFERENCE NA.U2 TS 3 1.3.1 001/050; PWG-5 (2.9/4.3) ANSWER 8.19 (1 00) (b)

      . REFERENCE NA TS 4.0.2 & 4.0.5 PWG-5: Generic TS Knowledge (2.9/3.9)

ANSWER 8.20 (1.00) (a) REFERENCE ADM 5.7, p 1. PWG-26: Logs / Records (3.3/3.6) I c

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 84 ANSWERS -- NORTN ANNA 112 -85/10/15-DEAN, WH ANSWER 0.21 (1.00)

(b) REFERENCE NA U2 TS 6.2.2 PWG-19: Fire Brigade (3.4/4.2) e

• ANSWER 8.22 (1.00)

(b) REFERENCE ADM 14.0, p 1. PNG-14: Tagging (3.6/4.0) ANSWER 8.23 (1 00) (c) REFERENCE ADM 14.1, p 3. FWG-14: Tagging (3.6/4.0) ANSWER 8.24 (1.00) (d) REFERENCE

1. IE information notice 82-51 'Overeuposures in PWR cavities'.

2. Surry 2 event, april 1979.

034/000; A1.02 (2.9/3.7) i 1

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 85 ANSWERS -- NORTH ANNA 112 -85/10/15-DEAN, WH ANSWER 8.25 (1.00)

(c) . REFERENCE 1-OP-4.1, p 15. EPE-036; EK3.01 (3.1/3.7) ANSWER 8.26 (1.00) (a) REFERENCE l 1-OP-4.1, p 5. 004/020; PNG-7 (3.4/4.1) ANSWER 8.27 (1.00) (c) REFERENCE NA EP 5.2.1.1 PWG-36: E-Plan (2.9/4.7) ANSWER 8.28 (1.00) (b) REFERENCE EPIP 3.02 and 3.03. PWG-36: E-Plan (2.9/4.7) 4 I e 4 i e N

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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS ,PAGE 86 8. ANSWERS -- NORTH ANNA 1A2 -85/10/15-DEAN, WH ANSWER 8.29 (1.00) (b) REFERENCE ADM 19.3, p 1. PNG-26: Loss (3.3/3.6)

                                          .e                        ,

ANSWER 8.30 (1.00) (b) REFERENCE ADN 19.18, P 2. PWG-2: Notification of Plant Supv. and Off-Plant Personnel (2.7/3.8) ANSWER 8.31 (1.00)

a. 1

b. 5

c. 4

d. 2

e. 7 REFERENCE NA U1 TS 3.4.6.2 002/02C; PWG-5 (2.9/4.1)

ANSWER 8.32 (1.00) 1.02 (+.5 ea) 1.09 REFERENCE NA U1 TS 3.2.4 001/050; PWG-5 (2.9/4.3) t e

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8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 87 ANSWERS -- NORTH ANNA 1&2 -85/10/15-DEAN, WH ANSWER 8 33 (1.00)

                                                            ~

a. 23

b. 100 or IN REFERENCE

o. NA U2 TS 3.9.10

b. NA U2 TS 3.9.3 034/000i PNG-5 (2.8/3.7)

ANSWER 8.34 (1.00) D. 1

b. 1

c. 3

d. 3

a. 1 REFERENCE NA Ula2 TS table 6.2-1.

PNG-23: Staffing / Activities (2.8/3.5) ANSWER 8.35 (2.00) 1.77 10 20,000 F Ih 60 the required shutdown margin is restored REFERENCE NA TS 3.1.1.1 004/010; A2.07 (3.8/3.9) s e

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l 1

8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 88 ANSWERS -- NORTN ANNA 112 -85/10/15-DEAN, WH ANSWER 8.36 (2.00)

1) Review and sign the Unit 1 CR0 shift turnover checklist.

2) Review and sign the Unit 1 Turbine Buildin3 shift turnover checklist.

3) Review and sign the outside shift turnover checklist.

4) Review and sign the auxiliary building shift turnover checklist.

5) Review and sign the backboards shift turnover checklist.

6) Review and sign the boiler,s and diesels shift turnover checklist.

REFERENCE ADM 19.3, p 1. PNG-23: Staffing / Activities (2.8/3.5) ANSWER 8.37 (1.00)

1. Its corresponding normal or emergency source is operable, and

2. All its redundant components are operable.

REFERENCE NA TS 3.0 5 062/0007 PWG-5 (3.0/4.0) ANSWER 8.38 (1.00)

1. Ioni=ing radiation.

2. Heat stress.

3. Differential pressure.

4. Potential o::ygen deficiency.

REFERENCE NA ADM 20.9, p 1. 103/000i K3.02 (3.8/4.2) 1 s i e

          , - , - - ,      -- .-      ,     - , - -   ~ - - - - - -     , , . , - . - , . .   ,e ,9 ., .- - ,.--     - - -

g e. , TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE 05.01 1.00 WMD0000260 05.02 1.00 WMD0000267 05.03 1.00 WMD0000277 05 04 1.00 WMD0000289 05.05 1 00 WMD0000130 05.06 1.00 WMD000025.' 05 07 1 00 WMD00007.34 05.08 1.00 WMD0000255 05.09 1.00 WMD0000258 " ' 05.10 1.00 WMD0000259 05.11 1.00 WMD0000261 i 05.12 1.00 WMD0000262 05.13 1.00 WMD0000264 05.14 1.00 WMD0000265 . 05.15 1.00 WMD0000266 05.16 1.00 WMD0000268 05.17 1.00 WMD0000269 05.18 1.00 WMD0000270 05 19 1.00 WMD0000272 05.20 1.00 WMD0000273 05 21 1.00 WMD0000274 05.22 1.00 WMD0000275 05.23 1.00 WMD0000276 05.24 1.00 WMD0000278 05.25 1.00 WMD0000280 05.26 1.00 WMD0000281 05.27 1.00 WMD0000282 05.23 1.00 WND0000283 05.29 1.00 WMD0000284 05.30 1.50 WMD0000283 09.31 1.00 WMD0000286 05.32 1.00 WMD0000287 05.33 1.00 WMD0000288 05.34 2.00 WMD0000256 03.35 2.00 WMD0000263 05.36 2.00 WND0000257 05.37 .50 WMD0000279 40 00 06.01 1.00 WMD0000308 . 06.02 1.00 WMD0000316 06.03 1.00 WMD0000341 ' 06.04 1.00 WMD0000160 06.05 1.00 WMD0000168 06.06 1.00 WMD0000310

• 06.07 1.00 UND0000313 06.08 1.00 WMD0000317 06.09 1.00 u n D 0000 3 2 '3 06 10 1.00 WnD0000329 '

                                                 ,   ,-     ,        I

o n. s TEST CROSS REFERENCE PAGE 2 5 QUESTION VALUE REFERENCE _ 06.11 1.00 WMD0000330 06.12 1.00 WM00000338 06 13 1.00 WMD0000347 06.14 2.00 WND0000295 06.15 1.50 WMD0000351 06.16 1.50 WMD0000293 06.17 1.50 WMD0000342 06.18 1.50 WMD0000349 06.19 1.50 WMD0000343 06.20 1.50 WMD0000346 9 , 06.21 1.50 WND0000297 06.22 2.00 WMD0000324 06.23 1.50 WMD0000333 06.24 1.00 WMD0000336 06.25 1.75 WND0000344 06 26 1.00 WMD0000048 06.27 1.25 WMD0000350 06.28 1 50 WMD0000352 06.29 1.00 WMD0000314 06.30 1.53 WND0000305 06.31 2.00 WM00000307 40.00 07.01 1.00 WMD0000436 07.02 1.00 WMD0000440 07.03 1.00 WMD0000433 07.04 1.00 WMD0000434 07.05 ' 1.00 WND0000435 07.06 1.00 WMD0000439 07.07 1.00 WMD0000442 07.08 1.00 WND0000443 07.09 1.00 WND0000444 07.10 1.00 WMD0000445 07.11 1.00 WMD0000446 07.12 1.00 WMD0000447

  ,       07.13          1.00  WND0000448 07.14          1.00  WM00000451 07.15          1.00  WMD0000452 07.16          1.00  WND0000453 07.17          1.00  WND0000454                                  '

07.18 1 00 WND0000455 07.19 1.00 WMD0000457 07.20 1.00 WND0000459 , 07.21 1.00 WND0000460 07.22 1.00 WM00000461 07.23 1.00 WHD0000464 , 07.24 1.00 WMD0000465 4 07.25 .50 WMD0000450 07.26 .50 WMD0000462 07.27 .50 WMD0000463 ll

TEST CROSS REFERENCE PAGE 3 GUESTION VALUE REFERENCE

 --------    ------       ----------   '-  a s,,

07.28 .50 WMD0000469 07.29 .50 WMD0000470 07.30 1.00 WMD0000458 07.31 1.00 WND0000467 07.32 1.00 WND0000468 07.33 1 00 WMD0000472 07 34 1.00 WMD0000437 07.35 3.25 WMD0000438 07.36 1.00 WMD0000441 07.37 1.00 WMD0000449 ., 07.38 1.00 WMD0000456 07.39 1.25 WMD0000466 07.40 1.00 WMD0000471  : 40.00 08.01 1.00 WND0000395 08.02 1.00 WMD0000411 08.03 1.00 WMD0000413 ' i 08.04 1.00 WMD0000414  ; 08.05 1.00 WMD0000396 , 08.06 1.00 WMD0000397 08.07 1.00 WND0000398 08.08 1.00 WMD0000399 08.09 1.00 WND0000401 08.10 1.00 WMD0000402 08.11 1.00 WND0000403 08.12 1.00 WMD0000404 08.13 1.00 WMD0000405 08.14 1.00 WMD0000406 08.15 1.00 WM00000407 08.16 1.00 WND0000408 .' 08.17 1.00 WMD0000409 08.18 1 00 WMD0000410 - 08.19 1.00 WMD0000412 08.20 1.00 WND0000415 08.21 1.00 WMD0000416 08.22 1.00 WMD0000417 08.23 1.00 WND0000418 08.24 1.00 WMD0000419 08.25 1.00 WMD0000420 08.26 1.00 WMD0000421 08.27 1.00 WMD0000422 08.28 1.00 WMD0000423 . 08 29 1.00 WMD0000424 08.30 1.00 WMD0000425 08.31 1 00 HMD0000420 08.32 1.00 WMD0000400 08.33 1 00 WMD0000426 08.34 1 00 WMD0000427 08.35 2. 00 WMD0000429 1 t I

    ..   .s     e
'                                       TEST CROSS REFERENCE       PAGE 4 QUESTION       VALUE    REFERENCE 08.36       2.00  WM00000430 08.37       1.00  WMD0000431 08.38       1.00  WM00000432 40.00

160.00

' .e , , i l l t 2 i 1 e a

              -                                                ? f

 ,                                                                  Attich cnt I
  ,                                                                 10/17/85 ENCLOSURE 4 Page 1 COMMENTS ON WRITTEN NRC EXAMINATIONS NORTH ANNA POWER STATION Reactor Operator Exam Section 1: No Comments Section 2:

General Comments on Section 2 While many questions in section 2 relate to the operation of the power station, some questions do not. For example questions 2.04, 2.09, 2.16, 2.18, 2.19, 2.20, 2.21 and 2.25 ask for detailed knowledge concerning the specific design of the plant with little emphasis on the operational aspects of the plant. The knowledge required by these questions is not supported by the job task analysis. Recommendations Evaluate operator performance on section 2 to ensure that a lack of knowledge of these detailed design questions will not result in the failure of operators who have demonstrated knowledge and skills on other i sections of the examinations. Question 2.02 Comment: Part c of Column A, answer key has (5) as correct answer. RCS samples come off of Hot and Cold legs of all loops thus all answer on key are correct. Recommendation: Change answer key to indicate all selections in Column B are correct.

Reference:

FM-93A and 89D. Question 2.05 Comment: Answer key has (b) as correct answer, hoeever, 1-OP-14.1 Step 4.1.14 throttles CC outlet MOV's. Step 4.1.25 mentions HCV-1758. Both (a) and (b) are correct in that throttling of CC HOV's determines position of HCV 1758 thus either or both can be throttled to control C/D rate. Recommendation: Change answer key to accept botn (a) and (b) as correct answers.

Reference:

1-OP-14.1 M'

Attcchmint I 10/17/85 Page 2 Question 2.06 Comment: The distractor b is a trick answer in that one word was chanted from the lesson plan (i.e., a total of I gpm vs. 1 gpm per line). The difference is not significant from an operational viewpoint. Thus both answer (b) and (c) are correct. Recommendation: Accept both b and c as correct answers.

Reference:

NCRODP-88.1, RCS Pressurizer and Pressurizer Relief, page 2.12. Question 2.09 Comment: Question not operationally oriented. Asks for memorized portions of operational procedure for plant cooldown. These types of knowledges are specifically not required to be memorized IAW NUREG 1021 ES202. The operation in question is only performed = once/18 months and the maintenance procedure is used which clearly provides the knowledge required by this question. Recommendation: Delete the question.

Reference:

HOP 5.04. Question 2.17 Comment: Answer key has " Hollow O-Ring and Inner and outer seal". These rings are commonly referred to at North Anna as " Seal Rings". Question asks for obscure descriptions and words taken out of context. We feel, upon examination, that all candidates can fully explain the entire vessel head seal system and leak detection system. Completion items should minimize the number of blanks. Although some multiple blank statements seem to measure rather complex reasoning abilities, such responses are more appropriate measures of intelligence than achievement. Recommendation: Delete question.

Reference:

Gronlund, N. E. Measurement and Evaluation in Teaching 2nd Ed. Question 2.18 Comment: Answer key is not complete. Recommendation: Add the following to the answer key:

1. T up through holes to cool upper head. *

2. T to T via discharge nozzle / core panel outlet interface.

h

Reference:

NCRODP 88.1.

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

Attcchm nt I 10/17/85 Page 3 Question 2.19 Comment: The exact shape of the rod pressure vs. life curve is not important to operators since the introduction of high density fuel pellets and use of pressurization. Operators know that in general the rod pressure increas'es through life but the presence of an initial drop or the changes in rate of pressure increase following this is not an operational concern. Recommendation: Delete the question.

Reference:

R0 Job / Task Analysis. Question 2.20 Comunent : Question asks the operator to distinguish between what usually occurs on loss of turbine load while at rated thermal to what is assumed to have occurred in the analysis for this design basis. Whether or not the operator is able to successfully answer this question correctly has not consequence to his safe operation of the plant. Recommendation: Note recommendation for general comment on Section , 2. Question 2.23 Comment: Knowledge of number of demineralizers is 0.K. however, knowledge of D.F.'s not required by operator. D.F.'s are not considered by operators when taking demins in/out of service. This is based on recommendation from Chemistry department. This ., knowledge not identified in RO Task list. Recommendation: Delete portion of question: "The mixed Bed's resin , will provide a minimum decontamination factor of 10 for all fission products except , and .

Reference:

This knowledge not referenced in RO JTA/KSA. Question 2.25 Comment: May be nice to know information but not information which will impact performance as reactor operator. Also second part of question is confusing where asking for effect on plant " thermal power capabilities". Answer key indicates examiner was looking for effect on generator output ("3 HW gain for Unit 2"). Recommendation: Delete question.

Reference:

No reference for this knowledge requirement in R0 Task list or KSA catalog. m

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Attachmsnt I g 10/17/85 Paga 4 Question 2.26 Cominent : Question asks for location - Not clear as to whether examiner is looking for phy:ical location or location within system creating confusion for candidate. Recommendation: Add the following to the answer key as acceptable answers. 1 CC Heat Exchanger /SW side. Basement of Aux Bldg. , East end of CC Heat Exchangers on floor.

2. Discharge t tunnel - Turbine Bldg. SW pit.

3. Discharge to reservoir -

Southeast corner of SW pump house.

4. Recirc Spray Hx's - Basement of Quench Spray pump house.

Reference:

Plant locations / drawings (FB). Question 2.31 c Comment: Location is not under rod drive room as answer key indicates. It is close to this, however, candidate may choose (1) j since it is outside of the drumming area. The correct description is - in the alley west of Aux. Buld, by primary plant gas house. Recommendation: Change answer key to accept (1) or (2) as correct i answers.

Reference:

Actual plant locations. Section 3: Question 3.03 - Comment: The answer key states that accum. discharge valves receive [ a signal to open on an SI signal and when RCS pressure is *2010  ; psig. RCS pressure '2010 psig is incorrect. The Accum. discharge l MOV's received signal to open on actuation of permissive (P-11)  ! which occurs at RCS pressure *2000 psig. Recommendation: Change answer key to reflect correct pressure.

Reference:

T.S. 33.2.1, Page 3/4 3-23. en

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Attechmint I 10/17/65 Page 5 Question 3.06 Comment: Question asks operator to distinguish between reactor trips designed into the reactor protection system and reactor trips taken credit for in the accident analysis. This type of information is beyond the scope of a reactor operator's knowledge and his ability to successfully answer the question has no consequence to plant operation. Recommendation: Delete the question.

Reference:

JTA/KSA. ., Question 3.09 Comment: Knowing the internals of the source range instrumentation has no bearing on the ability of the operator to safely operate the plant. (e.g: ite:.s B, F, G) Recommendation: Note recommendation for general comment on Section 2. Question 3.16 Comment: Answer b. calls for knowledge of maintenance tasks performed by instrumentation technicians. This is not a required operator knowledge and is not supported by the JTAs and KSA for this station. Recommendation: Delete portion 3.16 b. answer requirement.

Reference:

North Anna Operator JTA (KSA). Question 3.27 Comment: Answers (b) and (c) are both correct. Recommendation: Change answwer key to accept either (b) or (c) as correct.

Reference:

FSAR Vol. IX, Pages 6-2-90 and 6-2-100 (6-2-100 deals with NPSH). Section 4 Question 4.12 Comment: 7he question requires candidates to commit subsequent actions to memory (step 5.9.1 of AP-5.1) which is not required by NUREG 1021, ES 202, Section B-4. Recommendation: Delete the question.

Reference:

NUREG 1021, ES 202, Section B-4.

AttEchsent I

• 10/17/85 Page 6 Question 4.14 Comment: The questions requires candidate to commit to memory long j terms actions of AP-42. This is not IAW NUREG 1021, ES 202, Section B-4.

Recommendation: Delete question.

Reference:

NUREG 1021, ES 202, Section B-4. Question 4.38 comment: Answer key does not include all possible answer to the question. Recommendation: Additional answers that can be accepted include:

1. time and date of expiration

2. general instructions

3. respiratory protection requirements

4. radiation levels

5. contamination levels

6. airborne radioactivity levels

Reference:

NAPS Radiation Work Permit Form. Section 2.1 Radiation Protection Manual. I l l

Attachment I 10/17/85 Page 7 Senior Reactor Operator Exam Section 5: No Comments. Section 6: Question 6.02 C_omment: Answers (b) and (c) are both correct. Recommendation: Change answer key to accept both (b) and (c) as correct answers.

Reference:

NCRODP-88.1, RCS Pressurizer and Pzr Relief, Page 2.12. Question 6.06 Comment: Answers (b) and (c) are both correct. .

!                   Recommendation:   Change answer key to accept both (b) and (c) as correct answers.

References:

FSAR Vol. IX, Pages 6-2-90 and 6-2-100 (6-2-100 deals with NPSH). Question 6.10 Comment: The exact shape of the rod pressure vs. life curve is not important to operators since the introduction of high density fuel pellets and use of pressurization. Operators know that in general the rod pressure increases through life but the presence of an initial drop or the changes in rate of pressure increase following this is not an operational concern. Question 6.19 c Comment: Location is not under Rod Drive Room as answer key indicates - This is closest answer but candidate could just as easily choose (1) since it is outside the drumming area. The correct answer which is not one of the choices s "In alley west of Aux. Bldg. by primary plant gas house". Recommendation: Change answer key to accept either (1) or (2) as correct answers.

Reference:

Actual Plant Locations. m wm

Attechsrnt I 10/17/85 Page 8 Question 6.24 Comment: May be nice to know information but not information which will impact performance as reactor operator. Also second part of question is confusing where asking for effect on plant " thermal power capabilities". Answer key indicates examiner was looking for effect on generator output ("3 MW gain for Unit 2"). Recommendation: Delete question.

Reference:

No reference for this knowledge requirement in R0 Task list or KSA catalog. ., Question 6.28 . Comment: Setpoint should not be required especially with a (.25) point value assigned to it. The heat sensors are set / tested / adjusted by the Fire Marshall, not operations department. Thus, knowing that there is an automatic actuation by a heat sensor is acceptable. Recommendation: Delete requirement for setpoints - see comment section. Setpoint knowledge of minimium importance compared to knowledge of locations and methods.

Reference:

None Section 7 - Question 7.02 Comment: The question requires candidates to commit subsequent actions to memory (step 5.9.1 of AP-5.1) which is not required by NUREG 1021, ES 202, Section B-4. Recommendation: Delete the question.

Reference:

NUREG 1021, ES 202, Section B-4. Question 7.30 C_omment: This question asks for long term action memorization. LHSI pumps are not required to be operable in mode 6 by Tech. Specs. 1-0P-4.1 Page 13 of 49, Step 3.1.30 says "must have one LHSI or RP pump operable" thus LHSI may not be operable so question is not as straight forward as 1-AP-52 leads you to believe. Recommendation: Delete question.

Reference:

Tech. Specs, 1-AP-51. NUREG 1021 ES-202 page 3 of 6 (E.2). l l

Attschmint I

  ;                                                                       10/17/85 Page 9 Question 7.38 l

i Comment: Answer key does not include all possible answers to the question.

                                             ~

Recommendation: Additional answers that can be accepted include:

1. time and date of expiration

2. general instructions

3. respiratory protection requirements

4. radiation levels

5. contamination levels

6. airborne radioactivity levels

Reference:

NAPS Radiation Work Permit Form. Section 2.1 Radiation Protection Manual. Question 7.39 Comment: Question asks candidate to repeat long term actions. Not required by NUREG 1021. Recommendation: Delete question.

Reference:

NUREG 1021 ES-202. Section 8 Question 8.01 Comment: Question requires candidate to recall from memory four names that are attached to an Emergency Work Order. Since the reference is available to the SRO, memorization of form names is not appropriate. Also to reduce the workload on the SRO on shift the Operations Maintenance Coordinator process the " Equipment History / failure Analysis" form after the work has been completed and J then attaches it to the work package the next working day. Recommendation: Delete the question.

Reference:

Admin 16.5. Question 8.05 Comment: A difference exists in the action statements between Units 1 and 2. Unit 1 T.S. - Table 3.3 - 1 requires answers (d) or (e). Unit 2 T.S. Table 3.3-1 requires answer (c). Since unit was not specified, question is invalid. Recommendation: Delete the question.

Reference:

Unit 1 and 2 T.S. - Table 3.3-1.

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1

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        .                                                                    Attechm:nt I
    ,                                                                         10/17/85 Page 10 Question 8.07 Comment: All 3 distractors and " correct" answer are incorrect due to design change.

Recommendation: Delete question.

Reference:

Ammendment to License #58 and 66 Question 8.33 b Comment: A recent Tech. Spec. change has been received that changes the shutdown requirement to 150 hours. The contori room copy of T.S. has been revised but training copies have not been updated as of this date resulting in a conflict of candidate references. Recommendation: Change answer key to accept either 100 hours or 150 hours as correct answer.

References:

Unit 1 and 2 Tech. Specs. 3.93. ! Question 8.35 Comment: A recent T.S. change has been received that changes the required boron concentration to 12,950 ppm instead of 20,000. The control . room copy of this T.S. has been revised but training copies have not been revised as of this date resulting in a conflict of candidate references. Recommendation: Change answer key to accept within 20,000 ppm . or j 12,950 as correct answers.

Reference:

Unit I and 2 Tech. Specs. 3.1.1.1. i 3 i i l l l m . . , . _ . _,, y,y -

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