ML20206S103

Exam Rept 50-285/OL-86-02 on 860318.Exam Results:One Reactor Operator & Senior Reactor Operator Failed Written Exam & All Remaining Six Operators Passed

ML20206S103
Person / Time
Site:	Fort Calhoun
Issue date:	06/26/1986
From:	Cooley R, Mccrory S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20206S101	List: ML20206S094 ML20206S103
References
50-285-OL-86-02, 50-285-OL-86-2, NUDOCS 8607070290
Download: ML20206S103 (81)

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I FCS EXAMINATION REPORT-

, No. 50-285/0L-86-02 1

Docket No: 50-285 License No.: DPR-40 Licensee: Omaha Public Power District 1623 Harney Omaha, Ne 68012 i Examinations administered at Fort Calhoun Station i

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Chief Examiner: [17:,

4. lk McCrory, Lead Exami r Dete /

V Approved by: 3- d c/fd

'R."A. Cooley,Sectionpief Q6te /

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Summary

Examinations conducted on March 18, 1986.

, Written examinations were administered to two (2) Reactor Operators and six (6)

Senior Reactor Operators. Operating (Oral)examinationswereadministeredto two(2)SeniorReactorOperators. One (1) Reactor Operator and one (1) Senior Reactor Operator failed the written examination all others passed the examinations.

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2 Report Details

1. Examination Results SR0 Requalification R0 Requalification Total Pass Fail % Total Pass Fail  %

6 5 1 83 2 1 1 50

2. Examiners S. L. McCrory, Chief Examiner

3. Examination Report This Examination Report is composed of the sections listed below.

A. Examination Review Comment Resolution B. Exit Meeting Minutes C. General Comments D. Requalification Program Evaluation Report E. FCS Examination Key (SR0/R0 Questions and Answers)

Performance results for individual examinees are not included in this report because examination reports are placed in the NRC Public Document Room as a matter of course. Individual results may be retained in the NRC Region office during the period that the facility is evaluated as unsatisfactory but are not subject to public disclosure.

A. Examination Review Comment Resolution In general, editorial comments or changes made during the examination, the examination review, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive comments made during the examination review. The modifications discussed below are included in the master examination key which is provided elsewhere in this report as are all other changes mentioned above but not discussed herein. Attachment 1 is the facility comments on the examination. Unless otherwise indicated in this section, the facility comments were incorporated into the answer key.

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COMMENTS j (1) 1.05 The question does not state if the pressurizer relief is open or closed. An alternate acceptable answer is

ambient temperature since the candidate may have assumed i that the valve remained closed. Since statements of l assumptions was not requested, the answers in the key and

an answer of containment ambient temperature for both ,

j parts a and b are acceptable.  !

, Resp. REJECT. The valve closed condition is the trivial case

! and is not implied by the question. The examination I

instructions read at the beginning of each written examination require that all assumptions made by an i examinees be stated in their answer.

1.09c There is a second correct answer: There would be no (2) l effect on the cooldown rate since the Steam Generator i

tubes would remain sufficiently covered to support heat j

removal for a significant amount of time following valve closure.

i Resp. REJECT. This coment was not provided with justifying reference. Additionally, with feed lost to one i generator, the cooldown will be affected even before the

tubes become uncovered. With no cooler water entering i the steam generator, the heat transfer rate will

! decrease.

(3) 1.10c The following additional answers should be considered ,

{ correct:

i a. Steam Generator pressure stable or decreasing.

b. Steam Generator pressure controlled between 850 i and 1000 psia.

c. Steam Generator pressure consistent with setpoints on steam dump'and bypass system.

Resp. REJECT. Steam generator pressure stable or decreasing is insufficient by itself since the generator could be i decoupled and still 'show these conditions. The 1000 psia j upper limit for the steam bypass system is incorrect.

l Part c is essentially the same as a previously accepted

answer, t

! (4) 2.01/ The question and answer key as written concern a normal j 2.02/ operating procedure. As such, the question does not 6.03/ examine the operator's knowledge of the facility which is 6.04 to be covered in Category 2 (6). Procedure questions should be covered in Category 4 (7).

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4 The question requires memorization, rather than explanation, of the precautions / prerequisites in this normal operating procedure.

Resp. REJECT. Operating procedures are derived from knowledge of the design, function, and integration of the plants various systems. The system operating precautions and prerequisites in operating procedures reflect specific needs of the system based on its design and function.

Therefore, questions on operational precautions and prerequisites legitimately examine an operator's knowledge of a system.

Memorization of the operational precautions or prerequisites is not required since verbatim response is not demanded and significantly less that 100% of applicable precautions or prerequisites is not requested.

The operator's knowledge of the system design and operation should be adequate to answer the question.

(5) 3.07 The question simply asks what effect a temperature inversion has on the operation of RM-061. It does not ask for a discussion of Radon gases. This item should not have to be discussed for full credit. Full credit should be awarded for a discussion of the need to adjust the monitor setpoints using the normal / inversion switch in the control room or that the observed monitor count rate increases.

Resp. REJECT. The question asks for an explanation. The presence of Radon gases is the only reason that a thermal inversion has radiological significance under otherwise normal conditions.

(6) 4.01/ Operators are not required to memorize normal procedures 4.02/ but only those where immediate action is necessary.

7.01/ Procedure A0P-23 has no immediate action and is a 7.02 carefully and methodically followed procedure for reset of engineered safeguards.

Resp. REJECT. Operators were only asked to recall 3 of 6 overall conditions for securing safety injection. The question is examining operator familiarity with a procedure affecting significant safety issues. Verbatim (memorized) response was neither expected nor required.

(7) 8.02 Any outline inside the boundary given in the key is an acceptable answer because an operator would be at the controls while being inside that outline.

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5 Resp. REJECT. If the operator perceived a more confined boundary than the actual one, he or she may fail to respond to equipment and controls outside the perceived boundary.

(8) 8.03a This question could be interpreted as asking about equipment tagging in a more general sense. Therefore, a discussion of the independent verification required per paragraph 4.2.1.c of Standing Order 0-20 would also be an acceptable answer.

Resp. REJECT. The question clearly asks for actions after initial tagging.

(9) 8.06 Two additional tags which are commonly used on annunciators removed from service are "OUT TO CAL" and

" ANNUNCIATOR STATUS". These should also be considered correct answers. The last three blanks include only examples so all reasonable answers should be accepted.

Other correct answers would include: initials, time, date, etc.

Resp. REJECT. The tags given as alternate answers are not authorized by any existing procedure. The examples in the last three blanks indicate that the data provided should be informative as to the nature of the problem or corrective action being taken. The alternate answers provided do not fit this intent.

B. Exit Meeting Summary At the conclusion of the exan.ination period, the examiner met with the members of the facility staff to discuss the results of the examinations. The following personnel were present for the exit meeting:

NRC UTILITY T McCrory J. Fluehr P. Harrell J. Gasper M. Meador M. Core NRC informed the facility staff that all oral examinations were clear passes and no generic weaknesses were identified during the course of these examinations.

6 C. General Comments

On May 15, 1986 the facility administered written examinations to the

! two operators who failed the March 18, 1986 written examination.

These examinations were reviewed and approved by NRC prior to administration. After subsequent grading, both operators passed the

, examination and were allowed to resume licensed duties as of May 22, 1986. NRC reviewed the grading of these examinations and found it to

be satisfactory.

As a result of the March 18 and May 15 examination results. FCS is deemed to have a satisfactory remedial training program. The

- facility is allowed to reexamine operators who are found deficient as a result of NRC prepared and administered requalifications. The l conditions for license renewal stated in Examination Report 1

50-285/0L-86-01 will remain in effect until the next scheduled overall program evaluation in November 1986.

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D. Requalification Program Evaluation Report 4

Facility: Fort Calhoun Station Examiner: 5. L. McCrory Dates of Evaluation: 3/18/86 & 5/15/86 j Areas Evaluated: X Written X Oral Simulator Written Examination

1. Evaluation of Examination: SATISFACTORY

2. Evaluation of Facility Examination Grading

SATISFACTORY Oral Examination l 1. Overall Evaluation: SATISFACTORY

2. Number Observed: Number Conducted: 2 i Overall Program Evaluation Satisfactory: Marginal: Unsatisfactory: X (List major deficiency areas with brief Descriptive coments.)

The FCS requalification training program is satisfactory only for conducting remedial training of NRC identified deficient operators.

Submitted: Forwarded: Approved:

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/ Branch Chief

~5ecti~on Chief [

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8 E. FCS EXAMINATION KEY Date Administered: 3/18/86 Exam Type: Senior Reactor Operator Requalification Reactor Operator Requalification e

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U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION Facility: Ft. Calhoun Reactor Type: PWR-CE Date Administered: 3/18/86 Examiner: S. L. McCRORY Candidate: __

INSTRUCTIONS TO CANDIDATE:

READ THE ATTACHED INSTRUCTION PAGE CAREFULLY. THIS EXAMINATION REPLACES THE CURRENT CYCLE FACILITY ADMINISTERED EXAMINATION FOR DEMONSTRATION OF OPERATOR PROFICIENCY AND LICENSE RENEWAL. FAILURE OF THIS EXAMINATION WILL REQUIRE RETRAINING UNDER THE CURRENT FACILITY REQUALIFICATION TRAINING PROGRAM AND MAY REQUIRE RE-EXAMINATION BY NRC. Points for each question oro indicated in parentheses after the question number. The passing grade requires at least 70% in each category and a final grade of at least 80%.

Excmination papers will be picked up FOUR (4) hours after the examination oterts.

% of Cctcgory  % of Candidates's Category Value Total Score Value Category 15.0 23.44 1. Principles of Nuclear Power Plant Operations, Fluias, and Thermodynamic 16.0 25.0 2. Plant Design Including Safety and Emergency Systems

_16.5 25.78 3. Instruments and Controls 16.5 25.78 4. Procedures - Normal, Abnormal, Emergency, and Radiological Control 64.0 TOTALS Final Grade

• All work done on this examination is my own. I have neither given nor recoived aid.

Candidate's Signature

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

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

2. Restroom trips are to be limited and only one candidate at a time may leave.

You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

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

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

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

6. Use only the paper provided for answers.

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

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

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

10. Skip at least three lines between each answer. ,

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

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

13. The point value for each question is indicated in parentheses after the f

question and can be used as a guide for the depth of answer required. l I 14. Show all calcualtions, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

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

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

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

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

(1) Exam questions on top.

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

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

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

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

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

e. Do not dicuss the examination with other licensee staff personnel until the formal examination review is complete.

l 11__EBINQlELE1_QE_NUQLE88_EQWEB_EL8HI_QEEB8IlQNi PAGE 2 IBEBdQQ1Ned10ft_BE61_IB8NSEEB_8NQ_ELu1D_ELQW QUESTION 1.01 (2.00)

How would xenon concentration change for the power history shown below (Either a sketen or explanation is acceptable.)? Assume power changes are otep changes as shown. Assume xenon-free initially.

I P 100%-1 ************

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(%) 0%-1 * *********************

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1 0 40 80 120 160 200 1

hours I

i QUESTION 1.02 (1.00)

What are the three reasons for establishing regulating group insertion limits?

i QUESTION 1.03 (1.00)

A positive reactivity addition occurs in the core after a trip from power because of the increase in concentration of a certain fissile isotope.

What is the name of this fissile isotope and why does its concentration increase after a trip? (1.0)

QUESTION 1.04 (1.00)

I What are two automatic trips which act to prevent the core from violating DNBR limits?

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Iz__EBINQIELES_QE_ NUCLE 88_EQWEB_ELeNI_QEEB8I1QNt PAGE 3 IBEBdQQ1Nad191t_BE6I_IBeNSEEB_8ND_ELu1D_ELQW QUESTION 1.05 (2.00)

Calculate the pressurizer relief valve discharge temperature if quench tank pressure is 5 psig, there is a steam bubble in the pressurizer, and RCS pressure is:

c. 2035 psig (1.0)

b. 885 psig (1.0)

QUESTION 1.06 (1.50)

What are three indications of pump cavitation? (1.5)

QUESTION 1.07 (1.00)

Explain why the pressure / temperature limit curves must be redrawn periodically to account for increased radiation exposure of the reactor vessel.

QUESTION 1.08 (1.00)

With the reactor critical at 10 EE-04 %, rod withdrawal is used to increase power to 10 EE-03 %. Select the statement that correctly describes the position of rods after the power is stabilized at 10 E-3%. (1.0)

e. The rod position will be higher than at 10 EE-04% because more fuel must be exposed to the available neutrons to maintain the higher power level.

b. The rod position will be higher than at 10 EE-04% to overcome the power defect.

c. The rod position will be the same. The outward rod motion needed to achieve a given startup rate equals the inward motion needed to reduce the startup rate to zero,

d. The rod position will be lower than at 10 EE-04% due to the increased delayed neutron population associated with the higher power level.

(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

It__EBINGIELES_QE_NUGLE88_EQWEB_EL8HI_QEEB8IIQNt PAGE 4 IHEBBQQ1Nad1GSt_HE81_IB8NSEEB_8NQ_ELu1D_ELQW QUESTION 1.09 (3.00)

What effect would each of the following failures have on a natural circulation cooldown which is underway at 490 F. Explain your answers and consider each failure independently.

a. The steam dump valve which is being used to control cooldown rate fails open.

b. Level is lost in the pressurizer.

c. The auxiliary feedwater valve to one of the SG's fails shut.

QUESTION 1.10 (1.S0)

For each of the parameters listed below, what is the desired indication or trending that would be expected for natural circulation cooling?

a. RCS Hot Leg Temperature. (0.3)

b. Subcooling. (0.3)

c. Steam Generator Pressure. (0.3)

d. Steam Generator Level. (0.3) i o. Pressurizer Level. (0.3) 4 i

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2t__ELaNI_QES10N_INQLWQ1NQ_18EEI1_8NQ_EdEBGENQ1_S11IEd1 PAGE 5 QUESTION 2.01 (1.00)

A precaution in OI-AFW-01 permits the Auxiliary Feedwater Actuation System to be placed in the manual mode during power operation (mode 1), only if

= hat two (2) conditions are met?

i QUESTION 2.02 (1.00)

What prerequisite concerning the radiation monitoring system must be oatisfied prior to using auxiliary steam to supply heating stesm?

j QUESTION 2.03 (1.00)

Why does the " Compressed Air System - Normal Operation Operating Instructions," OI-CA-1, require that cooling water outlet valves be throttled down?

QUESTION 2.04 (3.50)

Place the following eight (8) steps, required per 01-CH-1 to establish i charging and letdown flow, in the proper order.

o. Open letdown stop valve TCV-202.

b. Ensure volume control tank has been vented and purged with nitrogen.

c. Reference OI-RC-2 and 3 for establishing and maintaining reactor coolant system pressure and pressurizer level control.

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d. When the letdown pressure oscillations have subsided, and if desired, place PIC-210 in automatic control.

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o. Ensure volume control tank is filled to normal level with proper concentration of boron and hydrogen overpressure is established.

f. Start one charging pump and manually adj ust the position of PCV-210 '

to maintain reactor coolant system pressure at 200 psig.

g. When pressurizer level is in AUTO control and letdown heat exchanger '

outlet temperature in stabilized at 120 degrees F, initiate flow through the boronometer and process radiation monitor by placing  !

controller HC-211-1 in AUTO. l

h. Ensure that the intermediate letdown backpressure controller switch PIC-210 controller setpoint is set at 200 psig.

(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

2t__ELeNI_QESIGN_INCLUQ1NQ_18EEI1_8NQ_EMEBGENGl_SYSIEd1 PAGE 6 QUESTION 2.05 (2.00)

Fill in the blanks in the following precautions taken from OI-DG-1. Blanks may contain one or more words or numbers.

c. In an EXTREME EMERGENCY the engine can be shutdown by pulling _____.

b. Damage to the turbocharger may result if a diesel is started or shutdown without _____.

c. Placing the Master Emergency Switch 183/MES at AI-133A/B out of the

" Normal" position will _____.

d. Do not load the diesel until _____ temperature is greater than 120 degrees F.

o. Allow the diesel generator to run at _____ after removing load.

QUESTION 2.06 (3.00)

Fill in the blanks below dealing with the Safety Injection system.

c. PUMP BUS VOLTAGE

1. Low press. pump SI-1A _____ _____ (0.3)

2. Low press. pump SI-1B _____ _____ (0.3)

3. High press. pump SI-2A _____ _____ (0.3)

4. High press. pump SI-2B _____ _____ (0.3)

5. High press. pump SI-2C _____ _____ (0.3)

b. 1. Whenever the reactor coolant system cold leg temperature is less than 330 degrees F and the reactor vessel head is installed,

_____ shall be _____. (0.5)

2. Whenever the reactor coolant system cold leg temperature is less than 320 degrees F and the reactor vessel head is installed,

_____ shall be _____. (0.5)

3. Whenever the reactor coolant system cold leg temperature is less than 282 degrees F and the reactor vessel head is installed,

_____ shall be _____. (0.5)

QUESTION 2.07 (2.50)

Technical Specification 2.2 on the Chemical and Volume Control System roquires that there be at least one flow path to the core for boric acid inj ection whenever fuel is in the reactor. Describe two of the three flow paths which meet this requirement per Tech Specs. (2.53

(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)

Zi__EL8HI_DE11GN_INCLUQ1NE_18EEI1_8NQ_EMEBGENQ1_111IEd1 PAGE 7 QUESTION 2.08 (2.00)

What are five (5) electrical systems which must be operable, per Tech Specs, prior to reactor heatup?

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2t__INSIBUNENI1_8NQ_CQNIBQL1 PAGE 8 QUESTION 3.01 (2.50)

With respect to normal operation of the wide range nuclear instrumentation:

a. When (at what power on which range (s)) is the High Rate of Change Reactor Trip activated? (1.0)

b. When (at what power on which range) will the zero power mode bypass trip be removed? (0.5)

c. What startup rate CDPM) will cause a Rod Withdrawal Prohibit? (0.5)

d. When is the audible count rate circuit in operation? (0.5)

QUESTION 3.02 (1.00)

OI-NI-1 requires comparison of the power range channels and wide range channels each shift. State the allowed deviation for en acceptable comparison of the:

a. wide range channels. (0.5)

b. power range channels. (0.5)

QUESTION 3.03 (3.50)

For each of the Reactor Protective System trips listed below, state the censed parameter, the Limiting Safety System setting, and what each trip protects against. Only the LSSS that would apply at 100% power is required.

c. High power level. (1.0)

b. Low reactor coolant flow. (1.0)

c. Steam generator differential pressure. (1.0)

d. Containment high pressure (Sensed parameter is NOT required). (0.5) l l

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2t__INSIBudENIS_8NQ_GQNIBQLS PAGE 9 QUESTION 3.04 (2.50)

Explain which, if any, Engineered Safety Features would initiate due to the conditions below. Assume each case is independent and all parameters other than those specified are normal.

c. Pressurizer pressure at 1600 psia. (0.5)

b. SIRW level at 18 inches above tank bottom. (0.5)

c. Steam generator pressure at 475 psia. (0.5)

d. Steam generator water level at 28% of wide range tap span. (0.5)

e. Containment pressure at 5 psig. (0.5)

QUESTION 3.05 (3.00)

c. What are three of the four ways that the electric motor driven fire pump may be started? For manual starts, give switch location. For automatic starts, give sensed parameter and setpoint. (1.5)

b. What are three of the four ways that the diesel driven fire pump may be started? For manual starts, give switch location. For automatic starts, give sensed parameter and setpoint. (1.5)

QUESTION 3.06 (2.00)

Assume the plant is operating at 100% power, with all systems normal, and the backup pressurizer heaters are energized.

c. How and why will pressurizer pressure change? (0.5)

b. How and why will pressurizer level change? (0.5)

c. How will control systems respond to the changes in (a) & (b)? (1.0)

QUESTION 3.07 (2.00)

Explain the effect a temperature inversion (where atmospheric temperature rises with increasing altitude) has on the operation of RM-061. (2.0)

(***** END OF CATEGORY 03 *****)

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4 t__EBQQEQUBER_ _NQBd6Lt_aRNQBdekt_EdEBGENCY_8NQ PAGE 10 BaQ10LQQlC6L_QQNIBQL QUESTION 4.01 (3.00)

The procedure for reset of engineered safeguards, AOP-23, lists six different conditions to be checked or satisfied before SI can be reset.

List three of these conditions. (Do not list the various parameter checks needed to verify a particular condition.)

QUESTION 4.02 (1.00)

Fill in the blanks in the following caution statement from AOP-19, loss of shutdown cooling.

A LPSI pump must not be allowed to operate at a flow rate of ___. If the cause of the flow stoppage cannot be determined immediately,___. ,

I QUESTION 4.03 (1.00)

During the heatup phase of a plant startup, shutdown cooling is secured and after satisfying the temperature and pressure limits for RCP operation, 3 RCPs are started between 1315 and 1330 to resume heatup. At 1345, CCW is lost and the RCPs are secured at 1350. At 1400, CCW is restored. How long uill it take to restore RCP forced flow neglecting the time it takes to otep through the pump starting procedure? Explain.

QUESTION 4.04 (2.50)

E0Ps 2-6 each require that two entry conditions be met before entering the procedure. The first entry condition requires that the immediate actions for reactor trip have been completed. The second entry condition is based on observation of at least one of the symptoms provided in a list. For each of the casualties listed below give two symptoms included in the list to satisfy the second entry condition.

A. ELECTRICAL EMERGENCY B. LOSS OF COOLANT ACCIDENT C. STEAM GENERATOR TUBE RUPTURE D. UNCONTROLLED HEAT EXTRACTION E. LOSS OF ALL FEEDkfATER l

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l st__EBQQEQUBER_:_NQBdeLi_8HNQBd8Lt_EMEBGENQ1_8NQ PAGE 11 l 88DIQLQGlQeL_GQNIBQL l QUESTION 4.05 (2.00)

Steps 3.3 and 3.4 of E0P-20, FUNCTIONAL RECOVERY, read as follows:

3.3 Check status of safety functions using Safety Function Status Check by identifying success path (s) currently in use for each safety function and then checking the appropriate acceptance criteria,

! 3.4 Assess all safety functions before going to Resource Assessment  !

Trees.

j Discuss what is done to " assess" all safety functions without using the assessment trees. Include the types of actions taken and the procedure (s)

.j to be used.

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I QUESTION 4.06 (1.50)

Revision 4 to RRP-VII-8-20, RADIATION WORK PERMITS, introduced a new type of RWP to the existing Standing RWP and Job Specific RWP. Name and describe the new RWP.

QUESTION 4.07 (3.00)

When frisking out of a controlled radiation area using a hand held probe and ratemeter:

A. What is the maximum background countrate allowable?

B. What reading above background requires decontamination?

i C. Describe the two acceptable methods for contacting HP personnel i

if contamination is detected.

I QUESTION 4.08 (2.50) l

For each of the following parameters, give the initial condition required by OP-3, Plant Startup From Hot Standby to Minimum Load.

A. Reactor power B. Pressurizer level C. Pressurizer pressure D. Steam Generator normal water level

E. Reactor Coolant temperature

(***** END OF CATEGORY 04 *****)

[ (************* END OF EXAMINGTRMM ***************h

f5tC LICENSE EXAMINATION HANDOUT EQUATIONS, CONSTANTS, /JID CONVERSIONS ,

6=rii*C*deltaT p 6=U*A*deltaT t

P = P *e /T SUR = 26/T P = Po*10sur*(t)

T=1*/p+(p-p)/Ip T=1/(p-5) T = ($-p)/Xp P " (Keff-1)/Keff = deltaKeff/Keff p=1*/TKeff+hff/(1+X.T)

A = In2/tg = 0.693/tg K = 0.1 seconds-1

' -ux I = Ie*e CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water Parameters _

1 gallon = 8.345 lbm = 3.87 liters 1 ft3 = 7.48 gallons Density 9 STP = 62.4 lbm/ft3 = 1 gm/cm3 Heat of vaporization = 970 Btu /lbm Heat of fusion = 144 Btu /1 N 1 atmosphere = 14.7 psia = 29.9 inches Hg.

Miscellaneous Conversions 1 curie = 3.7 x 101u disintegrations per second 1 kilogram = 2.21 lbm 1 horsepower = 2.54 x 103 Btu /hr 1 mw = 3.41 x 106 Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Btu = 778 ft-lbf

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_P__O_WE_R__P_L_A_N_T_O_P_ER_A_T_I_O_N_,

PAGE 12 ISEBdQQ186dICSz_SE61_IBedSEEB_6NQ_ELUIQ_ELQB 1

ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

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ANSWER 1.01 (2.00)

• = xenon concentration {d} /Xe peak 6"8 hrs / trip Xe pk-l ------------------------**----------***---------------

I * * *{f}x X 100%-1 -------------------****---*-------*----*--------------

e I * {c} * * *

/ I {a} * * *

• e 50%-1 ---------*-*--*---------------***-------*-------------

q l * ** {e}

• 1 * {b}

• 1 *

• Xe-free e i E * /60"80 hrs / trip

(%) 0%-1 * {s} ****

I 0 40 80 120 160 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> Grading: Locations marked ({}) should be shown or described at 0.286 each.

a. Xenon builds in almost to equilibrium value.

b. Xe. drops due to burnout increase when flux increases with <

iodine decay increasins more slowly.

c. Xe. builds in almost to equilibrium value.

d. Xe. builds to peak since burnout drops sharply with iodine decay dropping more slowly.

e. Xe. drops to equilibrium value.

f. Xe. increases to peak after trip. . Time and value are '

dependent on power history ("SQRT of % power).

3 About 60"80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> to negligible xenon. i REFERENCE C-E POWER SYSTEM NSS SERIES LECTURE NOTES, Figure VIII-87  ;

l 1

ANSWER 1.02 (1.00)

1. Acceptable power peakin3 factors are maintained.

2. Maintain minimum available shutdown margin.

3. Limit consequences of a CEA efection accident.

(0.333 ea). i l

l

12__EBINGIELEH_9E_SUGLE68_E9BEB_EL6NI_9EEB6IIDN2 PAGE 13 IBEBd99XN85IGSz_UE61_IB6NSEEB_6NQ_ELUIQ_ELQB ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

REFERENCE Technical Specification 2.10.2 ANSWER 1.03 (1.00)

Plutonium 239 (which builds up after a trip due to decay of Neptunium).

REFERENCE Basic Reactor Theory (CAF)

ANSWER 1.04 (1.00)

Any 2/5 of the following at 0.5 each

1. High power

2. Low RCS flow

3. TMLP

4. Axial Power Distribution Trip

5. Asymetric Steam Generator Transient Trip (differential pressure trip)

REFERENCE Technical Specification 1.3 ANSWER 1.05 (2.00)

a. 230 oF (0.5 for concept, 0.5 for math / steam table use)

b. 310 oF-(value as above.)

REFERENCE Steam Tables Basic Heat Transfer and Thermodynamics

1___EBINCIELES_gE_NyCLE68_EgBEB_EledI_ GEE 86IIgN2 PAGE 14 IHEBdggIN6dICSz_UE61_IB6BSEEB_edD_ELyIQ_ELQB ANSWERS -- FT. CALHOUN -86/03/18-HCCRORY, S.

ANSWER 1.06 (1.50)

Any 3 at 0.5 pach of tha followinst

1. excessive noise.

2. excessive vibration.

3. low suction pressure.

4. fluctuating pump amps.

5. Pluctuatins discharge pressure.

6. excessively low flow.

REFERENCE West. Thermal Hydraulic Principles and Applications to the PWR II, p. 10-54 ANSWER 1.07 (1.00)

Fast neutron irradiation of the reactor vessel will raise its reference transition temperature.

REFERENCE Technical Specification 3.1.2 ANSWER 1.08 (1.00)

c. Rod position will be the same.

REFERENCE Basic Nuclear Physics (CAF)

ANSWER 1.09 (3.00)

a. Increase cooldown rate (0.4) since more energy is beins removed from the primary. (0.6)

b. May interrupt natural circulation (0.4) since hot less maybe voided. l (0.6) OR No effect (0.4) if hot less are not voided (0.6). l 1

c. Decrease cooldown rate (0.4) since SG tubes will become uncovered reducing heat removal. (0.6) I l

, , , ,,- ,, n -- ,

12__EBINGIELES_gE_NyGLEe8_EgME8_ELeNI_QEEBBIIQN1 PAGE 15 IBE8dgp!NedIggi_UEeI_IBeBSEE8_6NQ_ELQIQ_ELg8 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

REFERENCE Basic Reactor Theory (CAF)

ANSWER 1.10 (1.50)

a. Th stable or decreasins.

b. Subcoolin3 constant or increasins. (>20 des F subcool margin)

c. SG Pressure trackins Tave (or Tc) saturation pressure. (SG Press 850-925 psia).

d. SG 1evel-in Narrow Ranse, stable. (level 35-85% narrow ranse) i l

e. PZR level trendin3 with charsins/ letdown. (level 10-70%)

REFERENCE General Physics; Heat Trans., Thermo., and Fluid. Flow Fundamentals, p. 356 EOPs 1, and 2 4

l.

4 I

1 1

4 l

2 1

. , - . . . . . . . . - - . . , . . . . - , . . - , . . . . . c- , - , . . . _ - - , _ . - , , , . , - . , ,. , . . , , , . , . . n.. . , . . ,,

l 22_._EgedI_DgSIgN_INggUDINg_S$EgIY_8ND_gNERGEHQY_SXSIENS -PAGE 16 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

I b

ANSWER 2.01 (1.00)

(1) The main feedwater system is not operating. (0.5)-

(2) The operator is manually directing AFW flow. (0.3)

REFERENCE FCS OI-AFW-1, R1, p. 1 ANSWER 2.02 (1.00)

Condenser Eveguation Radiation Monitor (RM057) must be properly operatins (calibrated, setpoints checked, & in operation).

REFERENCE FCS DI-AS-1, R6, p.1 f ANSWER 2.03 (1.00)

If the outlet valves are wide ~open, the standby compressor will trip (on l low cooling water pressure) if required to start (0.3) and the operating compressor may trip if the valves are,not properly throttled ( 0. 5 ) .-

i REFERENCE '

FCS OI-CA-1, p. 1 l ANSWER 2.04 (3.30) l Correct sequence ist b r' er he ar fr de er-s per the OI. j Items bresh may be in-any order but ALL THREE must be BEFORE items arfades.  !

Item e may be placed anywhere BEFORE item s. Items a and f may be reversed j IF the candidate states that the plant is. HOT. j i

l Grade at -0.5 per step out of sequence.

REFERENCE FCS OI-CH-1r_p. 3

21__eL681_DESIGU_ INCLUDING _S6EEIl_6dD_EdEBGENCY_SYSIEd3 PAGE 17 ANSWERS -- FT. CALHOUN -86/03/18-HCCRORY, S.

ANSWER 2.05 (2.00)

a. the manual injector control lever out (away from engine).

b. the oil circulating pump in operatica. (lube oil)

c. de-enersize the 4160 volt busses. (load shed emergency bus, isolate control circuits from control room and shift DG control to local)

d. Jacket water. (ensine coolant)

e. 900 rpm for at least 5 minutes.

(0.4 PT EACH)

REFERENCE FCS DI-DG-1, p. 1 ANSWER 2.06 (3.00)

a. 1. 1A3, 4160V

2. 1A4, 4160V

3. 183A, 480V

4. 184C, 480V

5. 183A, 480V

b. 1. at least i HPSI pump, placed in pull-stop (or disabled).

2. at least 2 HPSI pumps, placed in pull-stop (or disabled).

3. all 3 HPSI pumps, placed in pull-stop (or disabled).

(a; bus & volt G 0.15 ea.; b 1.-3. 0 0.25 ea, answer)

REFERENCE ,

FCS DI-SI-1, p. 1, 2 TS 2.3 (3) l l

l l

I 1

21__EL$NI_QESIGN_INCLUQINg_S$[EIy_6NQ_EMEBGEdgy_SySIEMS PAGE 10 ANSWERS -- FT. CALHOUN -86/03/18-NCCRORY, S.

i ANSWER 2.07 (2.50)

1. The boric acid pumps (0.5) can deliver boric acid tank (0.25) contents to the chargins pumps (0.5).

2. The safety injection pumps (0.5) can take suction from the SIRW (0.25) and inject into the core (0.5).

3. The chargins pumps (0.5) can take suction by gravity from the SIRW Lank (0.75).

4. The charging pumps (0.5) can take suction by gravity from the boric acid tanks (0.75).

(Any 2/4 at point values shown)

REFERENCE FCS TS 2.2 Basis, p. 2-18, 19 l

l 1

1 1

l l

22__EkedI_ DESIGN _INggUDIHg_S6EEIY_6BD_EdEBGENgy_SYSIENS PAGE 19 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 2.08 (2.00)

1. Unit aux. Power transformers TIA-1 or TIA-2.

2. House service transformers TIA-3 or TIA-4.

3. 4160V ESF buses 1A3 and 1A4.

4. 4160/480V transformers TIB-3A/3B/3C/4A/4B/4C.

5. 480V distribution buses 183A, 183A-4A, 184A, 1D3B, 1838-4D, 184D, 1B3C, 183C-4C, and 184C.

6. MCC 3A1, 381, 3A1, 3A2, 3C1, 3C2, 4A1, 4A2, 4C1, & 4C2.

7. 123VDC buses 1 &2 (EE-8F & EE-8G).

8. 125VDC distribution panels A1-41A & A1-41B.

9. Four instrument buses A-D.

10. AI-42A & AI-42B.

11. Two station batt. w/ >/e 1 chr. per bus.

12. Both EDG's ...

(Any 5/12 9 0.4 es.; component i not required - description only req'd)

REFERENCE FCS TS 2.7, p. 2-32

32__INSIBUdENIS_AND_CgBIBgkg PAGE 20 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 3.01 (2.50)

a. > 10EE-4% on wide range, < 15% on power range. (4 answers 0 0.25 ea.)

b. > 10EE-4% on wide range. (2 answers 0 0.25 ea.)

c. 1.5 DPM (0.5)

d. When the reactor is shutdown. (modes 3-5 and refueling) (0.5)

REFERENCE FCS OI-NI-1, p. 3 ANSWER 3.02 (1.00)

a. 1/2 decade.

b. 1%.

REFERENCE FCS DI-NI-1, p. 5 ANSWER 3.03 (3.50)

a. neutron flux (thermal backup) is sensed. (0.3)

LSSS 9 100% is 107% of rated power. (0.2)

Prevents clad damage due to reactivity excursions ~too rapid to be detected by pressure / temperature trips. (0.5)

b. RCS differential pressure across each SG is sensed.- (0.3)

LSSS 9 100% is 95% of max flow. (0.2)

Protects the core against DNB. (0.5)

c. Secondary differential pressure between ea SG is sensed. (0.3) i LSSS 9 100% is 135 psid. (0.2)

Protects against DNB and high linear heat rate. (0.5)

d. LSSS G 100% is 5 psis. (0.1) ,

Assures reactor is tripped whenever SI initiates. (0.4) l l

l REFERENCE FCS TS 1.3, p. 1-7 to 1-10 )

J

22__INSIBUUENIS_6BD_gDNIROLS PAGE 21 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 3.04 (2.50)

a. SIAS, CIAS, VIAS, DG START, Sequencers, DG Breaker Protect Override

b. No ESF would initiate due to low level in the SWIRT alone.

c. SGIS

d. AFW Actuation

e. SGIS, SIAS, CIAS, VIAS, DG START, Sequencers, DG Breaker Override REFERENCE FCS TS 2.14, Table 2-1, p. 2-64, 65 ANSWER 3.05 (3.00)

a. Auto start at 109 psi pump discharge pressure.

Manual start from the control room.

Manual start from the screenhouse local pushbutton station.

Auto start if transformer spray temp. bulb trips deluge valve open.

b. Auto start at 99 psi pump discharge pressure.

Manual start from the control room.

Manual start from the diesel (fp) control panel.

Auto if electric pump fails to restore pressure within ~10 seconds.

(for a a b; any 3/4 0 0.5 each)

REFERENCE FCS DI-FP-1, p. 1, 2

. . l I

22__INSIBUMENIS_eND_C0dISOLS PAGE 22 j ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

l ANSWER 3.06 (2.00)

a. Pressure will increase (0.2) due to the water phase chanSe (0.3).

b. There will be no significant level change (0.5).

OR

! Level will decrease (0.2) due to the removal of liquid mass from the pressurizer and the reduced volume of the remaining mass at higher pressure (0.3)

! c. 1. Pressurizer pressure increase will cause the spray valves to modulate open to reduce pressure. They will modulate shut as

pressure. nears setpoint.

2. Pressuriner level decrease, if occurs, will decrease letdown flow to restore level.

~

3. No other changes will occur.

J Point values for part c depend on answers in a & b. Accept any answer consistent with answers given in parts a a b.

REFERENCE FCS DI-RC-7, p. 4; i

l ANSWER 3.07 (2.00)

Radon 3ases (which are naturally emitted from the ground) normally rise and i are dissipated into the atmosphere. During an inversion, the gases are trapped near the surface (1.0). This causes a a rise in indicated background count rate (0.73) which may require setpoint adjustment (0.25).

i REFERENCE FCS DI-RM-1, p. 1 i

l 1

i 1

i i

4...,_. . . , _ ._ _ , - - _ - - . . . , , _ . , , , , , , _ , , ~ , , , . . . , . . . . . , . , _ . . __.._m., _ _ , _ , . ~ _ , , . . . . , , , , , , . , ,. - . . - _ - , _ , ~ _ .

S2__EBggEQQBES_ _dgBd661_6BdgBd6L1_EdE89 Edgy _6dp PAGE 23 B60I9699I966_998I896 ANSWERS - FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 4.01 (3.00)

(ANY THREE AT 1 PT EACH)

1. RCS subcooling > 200F

2. PZR level > 45% and not decreasins

3. At least one S/G available for RCS heat removal

4. RVLMS indicates CORE is covered

5. Adequate core heat removal (forced or natural circulation flow)

6. PZR pressure within the pressure temperature limits (of figure one).

REFERENCE AOP-23 STEPS 4&6 ANSWER 4.02 (1.00)

(0.5 EACH)

<200 spa stop the pump REFERENCE AOP-19 ANSWER 4.03 (1.00) ,

RCP flow can be restored immediately using any of the four RCPs up to a total of three. One of the pumps was never started and can be used with-any consideration of restart criteria. Since the other three pumps had- )

been running for at least 20 minutes, restarting them would not be considered a ' consecutive ' start in the context of the precaution of OI-RC-7.

REFERENCE l OI-RC-3, DI-RC-9, AOP-11 l l

l l

l l

I a

32__PBggEQUBES_;_UgBd6L2_6Byggd6L1_EME89ENCY_6NQ PAGE 24 B69I96ggIC66_CggIBg6 ANSWERS -- FT. CALHOUN -86/03/10-MCCRORY, S.

ANSWER 4.04 (2.50)

(.25 PT EA - 2.5 TOTAL)

SEE ATTACHED LISTS If symptoms other than those listed are given, they will be credited if they are unique to the event covered by the E0Pr observable in the control room, and can be readily verified by the examiner /srader.

REFERENCE E0Ps 2 through 6 ANSWER 4.05 (2.00)

The operator will determine which safety function acceptance criteria are being satisfied or can be met by implementins the operator actions section of E0P-20 for the safety functions whose acceptance criteria are not beins met.

, Question value is two points with no partial credit since the candidate's response will :,how that he/she either does or does not understand how the l facility expects this assessment to be accomplished.

REFERENCE E0P-20 AND FACILITY RESPONSE ANSWER 4.06 (1.50)

ACCESS CONTROLLED RWP (0.5) - used to control access to radiation 1 controlled areas when the RWP has prerequisite requirements which must be documented as being completed prior to entry. (1.0) (More detailed directions are contained in parasraphs IV.J.2 and V.B. of the procedure which are also acceptable in the answer.)

REFERENCE RRP-VII-8-20

1 TITLE: ELECTRICAL EMERGENCY 1.0 PURPOSE This procedure provides the operator actions which must be accomp-lished in the event of an Electrical Emergency (AC or DC) which causes a reactor trip. These actions are implemented after per-

~

forming the immediate actions in the Reactor Trip procedure and an electrical emergency has been diagnosed. The actions in this pro-cedure are necessary to ensure that the plant is placed in a stable.

l safe condition.

2.0 ENTRY CONDITIONS

a. The Reactor Trip Immediate Actions have been performed.

AND l

b. Plant conditions indicate that an electrical emergency has occurred. Any one or more of the following indications may be present:

i. Poactor trip.

ii. Loss of one or more RCP's.

I iii. Diesel generator (s) auto start.

j iv. 4.16 KV and 480 V bus low voltage / load shed.

v. Extensive loss of indications.

vi. DC bus under voltage.

vii. Loss of forced circulation caused by sustained loss of CCW.

l E0P-02 j ISSUED '

Page 2 of 32 JAN 0 G 1986 R0 01-06-86

TITLE: LOSS OF COOLANT ACCIDENT 1.0' PURPOSE This procedure provides the operator actions which must be accom-

~

plished in the event of a Loss of Coolant Accident. These actions are implemented after performing the immediate actions in the Reactor Trip procedure-and a loss of Coolant Accident has been diagnosed.

The actions in this procedure are necessary to ensure the plant is placed in a stable, safe condition.

i 2.0 ENTRY CONDITIONS j

a. The Reactor Trip Immediate Actions have been performed )

AND

}

b. Plant conditions indicate that a Loss of Coolant Accident has occurred. Any one or more of the following indications may be i

present:

1. Abnormal change in PZR level, I

11. Decreasin'g PZR pressure, iii. SI actuated automatically.

iv. Rising containment pressure.

v. High containment radiation.

vi. High quench tank level, pressure, and temperature.

! vii. Abnormal increase in containment sump level.

viii. Decreasing VCT 1evel.

l l

l E0P-03 ISSUED Page 2 of 38 J AN 061986 R0 01-06-86

TITLE: STEAM GENERATOR TUBE RUPTURE i 1.0 PURPOSE

This procedure provides the operator actions which must be accom-plished in the event of a Steam Generator Tube Rupture. These actions are implemented after performing the immediate actions of the Reactor Trip procedure and a steam generator tube rupture has been diagnosed. The actions in this procedure are necessary to ensure that the plant is placed in a stable, safe condition.

2.0 ENTRY CONDITIONS 1

a. The Reactor Trip Immediate Actions have been performed AND

b. Plant conditions indicate that a steam generator tube rupture has occurred. Any one, or more, of the following indications may be present:

1. Steam generator blowdown radiation alarm.

ii. Condenser off-gas high radiation alarm.

iii Unbalanced charging and letdown flows.

t iv. Decreasing pressurizer level and pressure.

v. High activity and conductivity in S/G liquid sample. 1 vi. Increasing S/G level, t vii. Decreasing VCT level.

! l l

l E0P-04 1SSUED page 2 of 24 JAN 061986

=

R0 01-06-86 1

TITLE: UNCONTROLLED HEAT EXTRACTION 1.0 PURPOSE This procedure provides the operator actions which must be accom-plished in the event of an Uncontrolled Heat Extraction. These actions are implemented after performing the immediate actions in the Reactor Trip procedure and an uncontrolled heat extraction has been diagnosed. The actions in this procedure are necessary to ensure the plant is placed in a stable, safe condition.

2.0 ENTRY CONDITIONS i

a. The Reactor Trip Immediate Actions have been performed.

AND

b. Plant conditions indicated that an Uncontrolled Heat Extraction has occurred. Any one, or more, of the following indications may be present:

1. Decreasing pressure in one or both S/G's (possible SGLS).

ii. Decreasing RCS average temperature caused by increased RCS heat removal, iii. Increase in main feedwater flow until feed header isolated on SGIS.

iv. Possible CIAS caused by high containment pressure with increases in containment temperature, humidity, and sump level.

v. High noise level (local).

E0P-05 ISSUED Page 2 of 28 JAN 0 01986 R0 01-06-86

t TITLE: LOSS OF ALL FEEDWATER 1.0 PURPOSE

' This procedure provides the operator actions which must be accomplished in the event of a Loss of All Feedwater. These actions are implemented after performing the immediate actions of the Reactor Trip procedure and a loss of all feedwater has been diagnosed. The actions in this proce-dure are necessary to ensure that the plant is placed in a stable, safe condition.

2.0 ENTRY CONDITIONS

a. The Reactor Trip Immediate Actions have been performed AND

b. Plant conditions indicate that a loss of all feedwater has occurred. Any one or more of the following indications may be present:

i. Decreasing steam generator water level (narrow range may be off scale low).

11. Main feedwater pump trip.

iii . No feedwater flow.

iv. Low main feedwater pump suction pressure.

i v. Reactor trip.

i i

E0P-06 ISSUED Page 2 of 24

. JAN O G 1986 R0 01-06-86 i

. . 1 i

l S___EBQQEQQ8ES_;_dg8d6L2_6BNQBd66 _EdE89 Edgy _6dp PAGE 25 Bo919699IGoL_G9 BIB 96 ANSWERS -- FT. CALHOUN -86/03/18-HCCRORY, S.

ANSWER 4.07 (3.00)

A. DELETED B. DELETED C. (1.5 pt each)

1. If another individual is present at the friskins station, have him contact HP via the Gai-Tronics System. Do not leave the frisking station until directed by HP tech.

2. If no personnel are present at the frisking station, don a pair of shoe covers and sloves and proceed to the nearest Gai-Tronics Station in the controlled area and contact HP.

l Other answers which may be considered reasonable but are not specified in the applicable procedure will be given 1/2 credit.

l 4

! REFERENCE RRP-HP-11 ANSWER 4.08 (2.50)

(0.5 each)

A. Between 10-4% and <2%

B. 48 + or - 4%

C. 2100 +50, -25 psia D. 65 + or - 10%

E. 515 - 5400F (or 532 des F for hot standby)

REFERENCE OP-3 l

)

U.S. NUCLEAR REGULATORY COMMISSION ,

SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION Facility: Ft. Calhoun Reactor Type: PWR-CE

! Date Administered: 3/18/86 Examiner: -S. L. McCRORY

, Candidate:

4

INSTRUCTIONS TO CANDIDATE

READ THE ATTACHED INSTRUCTION PAGE CAREFULLY. THIS EXAMINATION REPLACES THE CURRENT CYCLE FACILITY ADMINISTERED EXAMINATION FOR DEMONSTRATION OF i OPERATOR PROFICIENCY-AND LICENSE RENEWAL. FAILURE OF THIS EXAMINATION WILL AEQUIRE RETRAINING UNDER THE CURRENT FACILITY REQUALIFICATION TRAINING

PROGRAM AND MAY REQUIRE RE-EXAMINATION BY NRC. Points for each question i are indicated in parentheses after the question number. .The passing grade requires at least 70% in each category and a final grade of at least 80%.

Excmination papers will be picked up FOUR (4) hours after the examination

j. starts.

% of l Cctagory  % of Candidates's Category Value Total Score Value Category i

]

15.0 23.26 5. Theory of Nuclear Power

, Plant Operations,- Fluids,

and Thermodynamics I

16.5 25.58 6. Plant Systems Design, Control and Instruments 16.5 25.58 7. Procedures - Normal, i Abnormal, Emergency,'and j Radiological Control i 16.5 25.58 8. Administrative Procedures Conditions, and Limits j 64.5 TOTALS Final Grade  %

i j All work done on this examination is my own. I have neither given nor j received aid.

i:

Candidate's Signature l

7.,__. _ . - . _ . _ - _ . - - . , _ _ . - . _ _ _ _ _ ,

, ,~,,.. , w _w ,.m.% ,,, w . _ ,...,_

.,% _,m , , , . ,,, ..,.--,,,e v-- .r --

• -'mo -

O .

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

1. Cheating on the examination means an automatic denial of your application and

could result in more severe penalties.

2. Restroom trips are to be limited and only one candidate at a time may leave.

You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

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

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

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

6. Use only the paper provided for answers.

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

4

8. Consecutively number each answer sheet, write "End of Category _" as appropriate, start each category on a new page, write on only one side of the

{ paper, and write "Last Page" on the last answer sheet.

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

10. Skip at least three lines between each answer.

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

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

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

i

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

mathematical problems whether indicated in the question or not.

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND

DO NOT LEAVE ANY ANSWER BLANK.

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

17. You must sign the statement on the cover sheet that indicates that the work is 1

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

18. When you complete your examination, you shall:

a. Assemble your examination as follows:

i (1) Exam questions on top.

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

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

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

j

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

d. Leave the examination area, as defined by the examiner. If after leaving,

) you are found in this area while the examination is still in progress, your

license may be denied or revoked.

e. Do not dicuss the examination with other licensee staff personnel until the j formal examination review is complete.

4

m. _ _ _ -__ . . - . _ _ _ . _ - ..- _ _ _ . - _ - -

Sz__IBEQBY_QE_NUCLEeB_EQWEB_ELeNI_QEEBoI1QNt_ELUIDSt_eNQ PAGE 2 IBEBdQQ1Ned1CS QUESTION 5.01 (2.00)

TRUE or FALSE?

a. Beta-effective is smaller than Beta.

b. The lifetime for a prompt neutron is about 10-14 seconds.

c. A reactor's period is the time required to change power by a factor of 10.

d. The startup rate is the number of decades reactor power will change i r. 1 minute.

QUESTION 5.02 (2.00)

o. Why does Moderator Temperature Coefficient (MTC) become more negative as temperature increases at operating boron concentrations? (1.0)

b. Why does MTC become more negative as a function of core age? (1.0)

QUESTION 5.03 (2.00)

How would xenon concentration change for the power history shown below (Either a sketch or explanation is acceptable.)? Assume power changes are otep changes as shown. Assume xenon-free initially.

I P 100%-l ************

o I *

• w I *

• e 50%-I ************ vb s********

r I *

• l * *

(%) 0%-I * *********************

l________________________________________________________

0 40 80 120 160 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> l l

)

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

51__IBEQBY_QE_NVQLE88_EQWEB_EL8NI_QEEB8Il0Nt_ELUIDSt_8NQ PAGE 3 IHEBdQQYNed191 QUESTION 5.04 (1.00)

What are the three reasons for establishing regulating group insertion limits?

QUESTION 5.05 (1.00)

What are two automatic trips which act to prevent the core from violating DNBR limits?

QUESTION 5.06 (1.50)

What are three indications of pump cavitation? (1.5)

QUESTION 5.07 (1.00)

With the reactor critical at 10 EE-04 %, rod withdrawal is used to increase power to 10 EE-03 %. Select the statement that correctly describes the position of rods after the power is stabilized at 10 E-3%. (1.0)

a. The rod position will be higher than at 10 EE-04% because more fuel must be exposed to the available neutrons to maintain the higher power level.

b. The rod position will be higher than at 10 EE-04% to overcome the power defect.

c. The rod position will be the same. The outward rod motion needed to achieve a given startup rate equals the inward motion needed to reduce the startup rate to zero.

d. The rod position will be lower than at 10 EE-04% due to the increased delayed neutron population associated with the higher power level.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - -_.- l*__* * *

• _ C AT E G 0 RY__ 0 5 C O N T I N U E D _0 N N}X T_ P Q@@ * * * * *1____ __

l St__IBEQBl_QE_NUQLE8B_EQWEB_EL6NI_DEEBol10Nt_ELUIDSt_aNQ PAGE 4 IBEBUQQ188d101 QUESTION 5.08 (3.00)

What effect would each of the following failures have on a natural circulation cooldown which is underway at 490 F. Explain your answers and consider each failure independently,

c. The steam dump valve which is being used to control cooldown rate fails open.

b. Level is lost in the pressurizer.

c. The auxiliary feedwater valve to one of the SG's fails shut.

QUESTION 5.09 (1.50)

For each of the parameters listed below, what is the desired indication or trending that would be expected for natural circulation cooling?

c. RCS Hot Leg Temperature. (0.3)

b. Subcooling. (0.3)

c. Steam Generator Pressure. (0.3)

d. Steam Generator Level. (0.3)

o. Pressurizer Level. (0.3)

(***** END OF CATEGORY 05 *****)

1 Ez__EL6HI_SISIEH1_DE119Nt_GQUIBQL&_880_IN1IBUMENI8IIQN 5 PAGE QUESTION 6.01 (1.00) i Why does the " Compressed Air System - Normal Operation Operating Instructions," OI-CA-1, require that cooling-water outlet valves be throttled down? '*

I QUESTION 6.02 (2.00)

Fill in the blanks in the following sentence teken from the precautions of OI-CC-1, " Component Cooling - Normal Operation." Blanks may contain one or more words or numbers. (3.0) t The component cooling system must be __(a)__ any time __(b)__ is at 4

or above 130 degrees F to furnish cooling water to __(c)__, or whenever __(d)__ is operating to provide cooling water-to __Ce)__.

I i

l l QUESTION 6.03 (3.50) q Place the following eight (8) steps, required per 01-CH-1 to establish charging and letdown flow, in the proper order.

c. Open letdown stop valve TCV-202.

! b. Ensure volume control tank has-been vented and purged with nitrogen.

i 5

c. Reference OI-RC-2 and 3 for establishing and maintaining reactor coolant system pressure and pressurizer level control.

1 d. When the letdown pressure oscillations have subsided, and if. desired, place PIC-210 in automatic control.

I o. Ensure volume control tank is filled to normal level with proper J

concentration of boron and hydrogen overpressure is established.

f. Start one charging pump and manually adj ust the position of PCV-210 to maintain reactor coolant system pressure at 200 psig.

i i

g. When pressurizer level is in AUTO control and letdown heat exchanger outlet' temperature is stabilized at 120 degrees F, initiate flow i

through tho'boronometer and process radiation monitor by placing controller HC-211-1 in AUTO.

i

h. Ensure that the intermediate letdown backpressure controller switch PIC-210 controller setpoint is set at 200 psig.

t I

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****])___

ht__EL6NI_1111 EMS _QE11GNt_GQNIBQLt_eNQ_INSIBUMENIeIIQN PAGE 6 QUESTION 6.04 (2.00)

Fill in the blanks in the following precautions taken from 01-0G-1. Blanks cay contain one or more words or numbers,

c. In an EXTREME EMERGENCY the engine can be shutdown by pulling _____.

b. Damage to the turbocharger may result if a diesel is started or shutdown without _____.

c. Placing the Master Emergency Switch 183/MES at AI-133A/B out of the

" Normal" position will _____.

d. Do not load the diesel until _____ temperature is greater than 120 degrees F.

e. Allow the diesel generator to run at _____ after removing load.

QUESTION 6.05 (3.50)

For each of the Reactor Protective System trips listed below, state the sensed parameter, the Limiting Safety System setting, and what each trip protects against. Only the LSSS that would apply at 100% power is required.

a. High power level. (1.0)

b. Low reactor coolant flow. (1.0)

c. Steam generator differential pressure. (1.0)

d. Containment high pressure (Sensed parameter is NOT required). (0.5)

QUESTION 6.06 (3.00)

a. What are three of the four ways that the electric motor driven fire pump may be started? For manual starts, give switch location. For automatic starts, give sensed parameter and sotpoint. (1.5)

b. What are three of the four ways that the diesel driven fire pump may t be started? For manual starts, give switch location. For automatic starts, give sensed parameter and setpoint. (1.5)

(***** C AT E G O R Y 06 C O N T SNMG@ @GS C00D3V JQ@@ * * * *_*1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _

kr__EL6HI_111IEd1_DE11ENt_QQNIBQLt_8HD_INSIBudENI811DN PAGE 7 QUESTION 6.07 (1.50)

What are five (5) electrical systems which must be operable, per Tech Specs, prior to reactor heatup?

(***** END OF__ CATEGORY 06 *****)

Z1__EBQGEQUBgs_:_NQBd8Lt_6BNQBdekt_EUEBGENGl_6NQ PAGE 8 l 88910LQGlQ8L_QQNIBQL QUESTION 7.01 (3.00)

The procedure for reset of engineered safeguards, A0P-23, lists six different conditions to be checked or satisfied before SI can be reset.

List three of these conditions. (Do not list the various parameter checks needed to verify a particular condition.)

QUESTION 7.02 (1.00)

Fill in the blanks in the following caution statement from AOP-19, loss of chutdown cooling.

A LPSI pump must not be allowed to operate at a flow rate of ___. If the cause of the flow stcppage cannot be determined immediately,___.

QUESTION 7.03 (1.00)

During the hectup phase of a plant startup, shutdown cooling is secured and cfter satisfying the temperature and pressure limits for RCP operation, 3 RCPs are started between 1315 and 1330 to resume heatup. At 1345, CCW is lost and the RCPs are secured at 1350. At 1400, CCW is restored. How long will it take to restore RCP forced flow neglecting the time it takes to etep through the pump starting procedure? Explain.

QUESTION 7.04 (2.50)

E0Ps 2-6 each require that two entry conditions be met before entering the procedure. The first entry condition requires that the immediate actions for reactor trip have been completed. The second entry condition is based en observation of at least one of the symptoms provided in a list. For cach of the casualties listed below give two symptoms included in the list to satisfy the second entry condition.

A. ELECTRICAL EMERGENCY B. LOSS OF COOLANT ACCIDENT C. STEAM GENERATOR TUBE RUPTURE D. UNCONTROLLED HEAT EXTRACTION E. LOSS OF ALL FEEDWATER

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

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

Zi__P8QQEQUBER_:_NQBd8Lt_8BNQBd8Lt_EMEBGENQ1_8NQ PAGE 9 B8DIQLQQ1Q8L_QQNIBQL QUESTION 7.05 (2.00)

Steps 3.3 and 3.4 of E0P-20, FUNCTIONAL RECOVERY, read as follows:

3.3 Check status of safety functions using Safety Function Status Check by identifying success pathCs) currently in use for each safety function and then checking the appropriate acceptance criteria.

3.4 Assess all safety functions before going to Resource Assessment Trees.

Discuss what is done to " assess" all safety functions without using the assessment trees. Include the types of actions taken and the procedure (s) to be used.

t QUESTION 7.06 (1.50)

Revision 4 to RRP-VII-8-20, RADIATION WORK PERMITS, introduced a new type of RWP to the existing Standing RWP and Job Specific RWP. Name and describe the new RWP.

1 1

QUESTION 7.07 (3.00)

When frisking out of a controlled radiation area using a hand held probe and ratemeter:-

A. What is the maximum background countrate allowable?

8. What reading above background requires decontamination?

C. Describe the two acceptable methods for contacting HP personnel if contamination is detected.

QUESTION 7.08 (2.50)

For each of the following parameters, give the initial condition required by OP-3, Plant Startup From Hot Standby to Minimum. Load.

A. Reactor power .

B. Pressurizer level C. Pressurizer pressure D. Steam Generator normal water level l 4 E. Reactor Coolant temperature

(***** END OF CATEGORY 07 *****) _

At__8DMIN11188IIVE_EBQGEQUBEft_GQNDIIIQNit_8NQ_LIMII8IIQN1 PAGE 10 QUESTION 8.01 (1.00)

A. What minimum plant condition (s) require continuous coverage by a Shift Technical Advisor?

B. During off-normal conditions requiring STA assistance or presence,

'the STA must be able to report to the Control Room within how long?

QUESTION 8.02 (1.50)

Using figure 8.1, boldly outline the area in which the operator is said to be "at the controls".

QUESTION 8.03 (2.50)

A. When is an independent verification of equipment tagging required I

after initial tagging?

B. The equipment tagging procedure provides a list of 12 systems which require independent verification of component repositioning after tag I

clearance. List 8 of these systems.

QUESTION 8.04 (1.50)

Fill in the blanks in the following statement taken from S0-0-22, Containment Access and Egress. Blanks may represent single words or phrases.

If the containment Access Log is not in use, each person will be logged in and out of the containment-by ___ . This method will normally be used it. containment entries are ___ and ___ .

QUESTION 8.05 (2.00)

Fill in the blanks in the following statement from S0-0-25, Electrical and Machanical Jumpers and Blocks Control. Blanks may represent single words or phrases.

Prior to installation of a jumper or block, a ___ has been properly authorized in accordance with Standing Order G-17 unless the jumper is required during the performance of ___, ___, ___, special procedure, or ___ that cont rol ___ and ___ of the j umper .

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****D .

St__8DMIN1118611YE_EBQQEQMBEft_QQNDIl10N1t_6NQ_LidlI8110NS PAGE 11 QUESTION 8.06 (2.00)

Fill in the blanks in the following statement from S0-0-40, Annunciator Status Changes. Blanks may represent single words or phrases.

A ___ tag or ___ tag will be placed on each annunciator which is

___. Tags placed on annunciators should include applicable information such as ___, ___, and ___.

QUESTION 8.07 (2.00)

During operations at full reactor power, the emergency feedwater storage tur.k is declared inoperable and repairs are expected to be completed in Ebout three days. What are the Technical Specification requirements concerning operation of the plant? Figure 8.2 is the Technical Specification for Steam and Feedwater Systems.

QUESTION 8.08 (4.00)

While the plant is operating at 100% reactor power, your reactor operator computes a total Steam Generator leak rate of 1.5 gpm. At the same time you receive the radiochemistry report for the RCS which shows RCS activity to be 65 uCi/gm DOSE EQUIVALENT I-131. Before you can decide on a course of action, multiple lightning strikes from a severe electrical storm hit in the switchyard causing a loss of off-site power. While verifying the cutomatic actions for a reactor trip an operator notices that at least one of the Steam Generators indicates pressure below the no load saturation value for hot shutdown.

A. Identify each event which would require declaration of an emergency classification by itself based only on the information available, i

B. Using EPIP-OSC-1, determine a single classification for your current condition and j ustif y your selection f ully.

, (***** END OF CATEGORY 08 *****)_

NRC LICENSE EXMINATION HANDOUT EQUATIONS CONSTANTS, AND CONVERSIONS 6=m*C'*deltaT p 6=U*A*deltaT P = Po*10 sur*(t) P = P *et /T SUR = 26/T T=1*/p+(p-p)/Ip T=1/(p-$) T = ($-p)/X p p = 1*/TKeff + jeff/(1+X.T)

P = (Keff-1)/Keff = deltaKeff/Keff A = In2/tg = 0.693/tg X = 0.1 seconds-1

' -ux I = Io*e CR = S/(1-Keff) 2 R/hr = 6*CE/d feet l

Water Parameters l 1 gallon = 8.345 lbm = 3.87 liters ,

1 ft3 = 7.48 gallons i Density 9 STP = 62.4 lbm/ft3 = 1 gm/cm3 Heat of vaporization = 970 Btu /lbm Heat of fusion = 144 Btu /lb 1 atmosphere = 14.7 psia = m29.9 inches Hg.

• Miscellaneous Conversions I curie = 3.7 x 10iu disintegrations per second 1 kilogram = 2.21 lbm I horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 10b Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Btu = 778 ft-lbf

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2.0 LIMITING CONDITIONS FOR OPERATION 25 Steam and Feedvater Systems Applicability Applies to the operating status of the steam and feedvater systems.

Objective To define certain conditions for the steam and feedvater system necessary to assure adequate decay heat removal.

Specifications The reactor coolant shall not be heated about 3000F unless the following conditions are met:

(1) Both auxiliary feedvater pumps are operable. Cne of the auxiliary feedvater pu=ps may be inoperable for 2h hours provided that the redundant component shall be tested to demonstrate operability.

(2) A minimum of 55,000 gallons of water in the emergency feedvater storage tank and a backup water supply to the e=ergency feedvater storage tank from the Missouri River

.by the fire water system.

(3) All valves, interlocks and piping associated with the above co=ponents required to function during accident conditions are operable. Manual valves that could inter-rupt auxiliary feedvater flow to the steam generators shall be locked in the required position to ensure a flow path to the steam generators.

(4) The main steam stop valves are operable and capable of closing in four seconds or less under no-flow conditions.

Ensis i,

A reactor shutdown from power requires a removal of core decay heat. Immediate decay heat removal requirements are normally satisfied by the steam bypass to the condenser. Therefore, core decay heat can be continuously dissipated via the steam bypass to the condenser as long as feedvater to the steam generator is available. Normally, the capability to supply feedvater to the steam generators is provided by operatien of the turbine cycle feedvater system. In the unlikely event of complete loss of electrical power to the station, decay heat removal is by steam discharge to the atmosphere via the main steam safety and atmospheric dump valves. Either auxi-liary feedvater pump can supply sufficient feedvater for re-coval of decay heat from the plant. The minimum amount of water in the emergency feedvater storage tank is the amount needed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of such operation. The tank can ye r supplied with vater from the fire protection system.\l) e-Amendment No. 49 2-28

EPIP-0SC-1-1 Fort Calhoun Station Unit No.1 Emergency Plan Implementing Procedure EPIP-0SC-1 Operations Support Center EMERGENCY CLASSIFICATION I. PURPOSE To provide descriptions and examples initiating conditions for the four Emergency Classifications: 1) Notification of Unusual Event; 2) Alert; 3)

Site Area Emergency; and 4) General Emergency.

II. PREREQUISITES

1. Occurrence of an incident disrupting normal operation of the plant.

111. PRECAUTIONS

1. Monitor conditions carefully and act conservatively to minimize the impact on the surrounding environment, health and safety of the public as well as plant equipment and personnel. ,

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2. Whenever the initiating conditions are not sufficiently defined to absolutely distinguish between either the ALERT or SITE AREA classification, the determination will be made in accordance with the following general guidance:

a. If in the judgement of the Shif t Supervisor or Site Director a l release of radioactivity has occurred for longer than 15 minutes and is espected to continue for greater than one hour, a Site Area emergency should be declared; or

b. If in the judgement of the Shift Sueprvisor or Site Director the l general safety of the plant could degrade by-gross contamination within the plant, a Site Area emergency should be declared; or

c. If in the judgement of the Shift Supervisor or Site Director the l general safety of the public could be reduced by the unavailability of plant functions needed for protection of the public, a Site Area emergency should be declared.

IV. PROCEDURE

1. Upon recognition of an abnormal plant or site condition the observer shall notify the Shift Supervisor of the situation.

FC/EPIP/04 R7 6-20-85 E . .

l_. -

EPIP-OSC-1-2 IV. PROCEDURE (Continued)

2. The Shift Supervisor shall (using EPIP-OSC-14 as a guide):

a) Evaluate the condition and determine the applicable emergency classification per Figure OSC-1.1, Emergency Action Level (EAL) Criteria. Figure OSC-1.2 should be used to aid in making GENERAL ENERGENCY offsite protective action decisions.

b. Declare the applicable classification and activate the Radiological Emergency Response Plan (RERP) per EPIP-OSC-2.

c. Upon activating the RERP refer to the appropriate Emergency Plan Implementing Procedure (EPIP) for immediate and follow-up actions based upon the following:

1) Notification of Unusual Event Classification Actions, EPIP-OSC-3

2) Alert Classification Actions, EPIP-OSC-4

3) Site Area Emergency Classification Actions, EPIP-OSC-5

4) General Emergency Classification Actions, EPIP-OSC-6

5) Initial assessment of plant parameters and effluent monitors to determine source term, EPIP-OSC-10.

6) Initial dose assessment based on plant instrumentation, EPIP-OSC-11.

d. Monitor response activities and plant conditions and adjust the classifications as necessary.

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EPIP-0SC-1-4 FlGURE OSC-1.2 ,

I FLOW CHART FOR GENERAL EMERGENCY OFFSITE PROTECTIVE DECISIONS The following .ctions wa.i be based on predeterman.d observabl. instewmentation and plant et.tus indicators

( EAle) contained in the .mergency plan and to t have been revie.ed by of fsite of fier l.. Ho.ever. e..oonsible of f site of ficials must dectae on the fe.sicality of implernenting the pect.ctive actions .t the time of the ocident.

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Sz__IHEQBl_QE_NUGLE88_EQWEB_EL8NI_QEEB8IIQNt_ELUIDSt_8NQ PAGE 12 IHEBdQQ1N8 DIGS ANSWERS _- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 5.01 (2.00)

a. True.

b. False.

c. False,

d. True.

REFERENCE C-E POWER SYSTEM NSS SERIES LECTURE NOTES, p. VIII-60 to 67

] ANSWER 5.02 (2.00)

e. Water (as a liquid) expands exponentially as temperature increasesC.67)

This increases the density and boron-related changes to reactivity per degree change in moderator temperature (0.33).

b. MTC primarily changes with core age due to the decrease in boron concentration, which reduces the positive reactivity added by driving boron out of the core.

REFERENCE C-E POWER SYSTEM NSS SERIES LECTURE NOTES, p. VIII-69 l

St__IHEQBY_QE_NUCLEeB_EQWEB_EL8NI_QEEB8IIONt_ELUIDSt_6NQ PAGE 13 IHEBdQQ1N8 MIGS ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

4 ANSWER 5.03 (2.00)

• = xenon concentration {d) /Xe-peak 6~8 hrs / trip Xe-pk-l ----------------------- **--------- ***---------------

1 * * *{f)*

i X 100%-1 ------------------ ****-- *------ *--- *--------------

, e 1 * {c) * * *

/ I {a) * * *

• e 50%-1 -------- * *- *-------------- ***------ *-------------

q l * ** {e)

• 1 * {b) *

! 1 *

• Xe-free e 1 *

• 60~80 hrs / trip 1 ****

(%) 0%-1 * {g) i l________________________________________________________

0 40 80 120 160 200 hourc Grading: Locations marked ({}) should be shown or described at 0.286 i cach.

i a. Xenon builds in almost to equilibrium value.

b. Xe. drops due to burnout increase when flux increases with j iodine decay increasing more slowly.

j c. Xe. builds in almost to equilibrium value.

d. Xe. builds to peak since burnout drops sharply with iodine decay dropping more slowly.

, e. Xe. drops to equilibrium value.

j f. Xe. increases to peak after trip. Time and value are dependent on power history (~SQRT of % power).

g. About 60~80 hours to negligible xenon.

REFERENCE C-E POWER SYSTEM NSS SERIES LECTURE NOTES, Figure VIII-87 4

e ANSWER 5.04 (1.00)

1. Acceptable power peaking factors are maintained.

i

2. Maintain minimum available shutdown margin.

j 3. Limit consequences of a CEA efection accident.

(0.333 ea) 1 4

i

?

. - _ _ . . - _ _ . - _ _ , - . _ . . . _ . . _ . , _ . , , . . , . , , , _ _ ,_ , ,_ , , __,. , . - .-. . _ , , ,er .r._,.-_,, . . . - ~ , , . . . _ , , - - _ - . , , _

.-- ~ .. .

l 52__IHE981_QE_NWGLE88_EQWEB_EL6NI_9EEB6IIQHi_ELUIQ1 _eNQ PAGE 14 IHEBBQQ1NedIGS-ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

REFERENCE' Tochnical Specification 2.10.2 ANSWER 5.05 (1.00)

Any 2/5 of the following at 0.5 each

1. High power l 2. Low RCS flow

3. TMLP

4. Axial Power Distribution trip

5. Asymmetric SG Transient trip (differentimi pressure trip) l REFERENCE Technical Specifications 2.3.1, 2.3.2.and 2.3.4 1

l ANSWER 5.06 (1.50) l Any 3 at 0.5 each of the following:

, 1. excessive noise.

! 2. excessive vibration.

I 3. Iow suction pressure.

4. fluctuating pump amps.

t 5. fluctuating discharge pressure.

! 6. excessively low flow.

i REFERENCE West. Thermal Hydraulic Principles and Applications to the PWR II, p. 10-54 i

k ANSWER 5.07 (1.00) 1

c. Rod position will be the same.

~

REFERENCE Basic Nuclear Physics (CAF) l I l

i 3

i

St__IHEQBI_QE_NWCLE88_EQWEB_EL8HI_QEEB8IIQNt_ELUIQSt_8NQ PAGE 15-IHEBdQQINedIGS i

ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 5.08 (3.00)

e. Increase cooldown rate (0.4) since more energy is being removed from the primary. (0.6)

b. May interrupt natural circulation (0.4) since hot legs maybe voided.

(0.6) OR No effect (.4) if hot legs do not become voided C.6)

c. Decrease cooldown rate (0.4) since SG tubes will become uncovered reducing heat removal. (0.6)

REFERENCE Basic Reactor Theory (CAF) i ANSWER 5.09 (1.50)

a. Th stable or decreasing.

b. Subcooling constant or inc reas ing . (>/= 20 deg F subcooled margin)

c. SG pressure tracking Tave (or Tc) saturation pressure (850-925 psia)

d. SG 1evel in Narrow Range, stable. (35-85% narrow range)

o. PIR level trending with charging / letdown. (10-70% level) j REFERENCE General Physics; Heat Trans., Thermo., and Fluid Flow Fundamentals, p. 356 i

I

62__EL8HI_SYSIEMS_DESIQNt_CQNIBQLt_8NQ_INSIBUMENI8IIQN PAGE 16 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 6.01 (1.00)

If the outlet valves are wide open, the standby compressor will trip Con low cooling water pressure) if required to start. The operating compressor may also trip if the valves are not properly throttled.

REFERENCE FCS OI-CA-1, p. 1 ANSWER 6.02 (2.00)

c. in operation. (0.4)

b. RCS temperature. (0.4)

c. the RCP seals. (0.4)

d. a RCP. (0.4)

o. the motor oil coolers. (0.4)

REFERENCE FCS OI-CC-1, p. 1 ANSWER 6.03 (3.50)

Correct sequence is: b, e, h, a, f, d, c, g per the OI.

Items b,e,h may be in any order but must ALL come BEFORE items a,f,d,g.

Item c may be in any order but must be BEFORE item g.

Items a and f may be reversed if the candidate states that the plant is hot.

Grade at -0.5 per step out of sequence.

REFERENCE FCS OI-CH-1, p. 3

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

i . .

6t__EL6NI_SISIEMS_DESIGNt_GQNIB9Lt_8N0_INSIBudENI8IION PAGE 17

ANSWERS -- FT. CALHOUN -86/03/18-NCCRORY, S.

4 1

i l

ANSWER 6.04 (2.00)

a. the manual injector control lever out (away from engine). (0.4)

b. the oil circulating pump in operation. Clube oil) (0.4)

c. de-energize the 4160 volt busses. Cload shed the emergency bus, i isolate control circuits from control room, shift DG to local) (0.4)

{

i

! d. ,facket water. (engine coolant) (0.4) f o. 900 rpm for at least 5 minutes. (0.4)

(0.4 each)

REFERENCE FCS OI-DG-1, p. 1 i

ANSWER 6.05 (3.50) i .

J a. neutron flux (thermal backup) is sensed. (0.3)  !

I LSSS 4 100% is 107% of rated power. (0.2) l Prevents clad damage due to reactivity excursions too rapid to be -

i detected by pressure / temperature trips. (0.5) i b. RCS differential pressure across each SG is sensed. (0.3)

LSSS 4 100% is 95% of max flow. (0.2)

.! Protects the core against DNS. (0.5) i c. Secondary differential pressure between es SG is sensed. (0.3)

I LSSS e 100% is 135 ps id. (0.2)

) Protects against DN8 and high linear heat rate. (0.5) '

I i d. LSSS e 100% is 5 psig. (0.1) j Assures reactor is tr ipped whenever SI initiates. (0.4)

REFERENCE i FCS TS 1.3, p. 1-7 to 1-10

)

i 4

_-____.____.______.__________._.___.__._________.___.____.__m..

6t__ELeNI_21SIEd1_DESIGNt_QQNIBQLt_8NQ_INSIBudENI6IIQN PAGE 18 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 6.06 (3.00)

m. Auto start at 109 psi pump discharge pressure.

Manual start from the control room.

Manual start from the screenhouse local pushbutton station.

Auto start if transformer spray temp. bulb trips deluge valve open.

b. Auto start at 99 psi pump discharge pressure.

Manual start from the control room.

Manual start from the diesel (fp) control panel.

Auto if electric pump fails to restore pressure within ~10 seconds.

(for a & b; any 3/4 8 0.5 each)

REFERENCE FCS OI-FP-1, p. 1, 2 ANSWER 6.07 (1.50)

1. Unit aux. power transformers TIA-1 or TIA-2.

2. House service transformers TIA-3 or TIA-4.

3. 4160V ESF buses 1A3 and 1A4.

4. 4160/480V transformers TIB-3A/3B/3C/4A/4B/4C.

5. 480V distribution buses 193A, 183A-4A, 194A, 1838, 1938-48, 1848, 193C, 183C-4C, and 184C.

6. MCC 3A1, 3B1, 3A1, 3A2, 301, 3C2, 4A1, 4A2, 401, & 4C2.

7. 125VOC buses 1&2 (EE-8F & EE-8G).

8. 125VOC distribution panels A1-41A & A1-418.

9. Four instrument buses A-0,

10. AI-42A & AI-428.

l

11. Two station batt. w/ >/= 1 chr. per bus.

12. Both EDG's ...

1 (Any 5/12 4 0.3 ea.; component # not required - description only req'd) l I

Et__EL8HI_111IEB1_DE11GNt_CQNIBQLt_8NQ_INSIBWHENI8IION PAGE 19 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

REFERENCE FCS TS 2.7, p. 2-32

(

Z2__EBQQEQUBES_=_NQBd8Lt_etNQBd8Li_EMEBQENQX_6NQ PAGE 20 B8DIQLQQIQ8L_QQNIBQL ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 7.01 (3.00)

(ANY THREE AT 1 PT EACH)

1. RCS subcooling > 20oF

2. PZR level > 45% and not decreasing

3. At least one S/G uvailable for RCS heat removal

4. RVLMS indicates CORE is covered

5. Adequate core heat removal (forced or natural circulation flow)

6. PZR pressure within the pressure temperature limits (of figure one).

REFERENCE A0P-23 STEPS 486 ANSWER 7.02 (1.00)

(0.5 EACH)

<200 gpm otop the pump REFERENCE AOP-19 ANSWER 7.03 (1.00)

RCP flow can be restored immediately using any one of the four RCPs up to e total of three. Since one of the pumps was not started, it can be started without considering res tart cr iter ia. Since the other three pumps were running for at least 20 minutes, the " consecutive" start concern of OI-RC-7 does not apply.

REFERENCE OI-RC-3, 01-RC-9, AOP-11

Z2__EBQQEQUBE1_ _NQBdekt_8BNQBdekt_EMEBGENQX_8NQ PAGE 21 86DIQLQQIQaL_QQNIBQL ANSUERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 7.04 (2.50)

C.25 PT EA - 2.5 TOTAL)

SEE ATTACHED LISTS If answers contain symptoms which are unique to the event, observable from the control room, and readily verif iable by the examiner / grader, give full credit.

REFERENCE E0Ps 2 through 6 ANSWER 7.05 (2.00)

The operator will determine which saf ety f unction acceptance criteria are boing satisfied or can be met by implementing the operator actions section of E0P-20 for the safety function whose acceptance criteria are not being mot.

Question value is two points with no portial credit since the candidate's response will show that he/she either does or does not understand how the facility expects this assessment to be accomplished.

REFERENCE E0P-20 AND FACILITY RESPONSE ANSWER 7.06 (1.50)

ACCESS CONTROLLED RWP (0.5) - used to control access to radiation controlled areas when the RWP has prerequisite requirements which must be documented as being completed prior to entry. (1.0) (More detailed directions are contained in paragraphs IV.J.2 and V.B. of the procedure thich are also acceptable in the answer.)

REFERENCE RRP-VII-8-20 i

Zi__EBQCEDUBES_:_NQBU8Li_8BNQBd8Li_EMEBGENGl_eND PAGE 22 B6DIQLQGIC8L_CQNIBQL ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 7.07 (3.00)

A. DELETED

0. DELETED C. (1.5 pt each)

1. If another individual is present at the frisking station, have him contact HP via the Gai-Tronics System. Do not leave the frisking station until directed by HP tech.

2. If no personnel are present at the frisking station, don a pair of shoe covers and gloves and proceed to the nearest Gai-Tronics Station in the controlled area and contact HP.

If other answers are given which may be considered reasonable but are not address by an appropriate procedure, 1/2 credit will be given.

REFERENCE RRP-HP-11 ANSWER 7.08 (2.50)

(0.5 each)

A. Between 10-4% and <2t B. 48 + or - 4%

C. 2100 +50, -25 psia i D. 65 + or - 10%

E. 515 - 540oF (or 532 deg F no load hot standby)

REFERENCE l OP-3 l l

w

l l

az__eDUINISIB8IIVE_EBQCEQUBESi_CONDIIIQNSi_eND_LIMII6IIQNS PAGE 23 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 8.01 (1.00) -

(0.5 ea)

A. RCS temperature > 210oF Call conditions except cold SD and refuel)

B. 10 minutes REFERENCE S0-G-43 ANSWER, 8.02 (1.50)

SEE ATTACHED FIGURE The question is worth 1.5 pt with no partial credit. An allowable variance on the boundary is shown on the key.

REFERENCE S0-G-9 ANSWER 8.03 (2.50)

A. No condition requires an independent verification of equipment tagging. The craftsman must verify a tagout but may accompany the operator hanging the tags. (0.5)

B. (ANY 8 AT 0.25 EACH)

1. RCS

2. CONTAINMENT ISOLATION

3. DIESEL GENERATORS

4. AUX FEEDWATER

5. CONTAINMENT SPRAY

6. SI LPSI & HPSI

7. ENGINEERED SAFEGUARDS CONTROLS

8. CVCS

9. FIRE PROTECTION

10. RAW WATER

11. CCW

12. WASTE DISPOSAL (Give 1/2 credit for systems which are a subsystem of those above.)

REFERENCE S0-0-20

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f

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1

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I Hi__eDUINISIB6IIVE_EB9CEDUBEli_GONDIII9NSi_6NQ_LIMII8IIONS PAGE 24 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

1 l

l ANSWER 8.04 (1.50) j i sppropriate entries in the control room log infrequent Cemergency conditions) limited to small groups of personnel  :

(0.5 each) i REFERENCE S0-0-22 ANSWER 8.05 (2.00)

1. maintenance order for the installation (0.33)

2. an issued surveillance test (0.33)

3. maintenance procedure (0.33)

4. calibration procedure (0.33)

5. operating procedures (0.33)

6. installation (0.167)

7. removal (0.167)

Items 2-5 may be in any order and items 6 and 7 may be reversed.

Accept "PRC approved procedure" for any ONE of 2 - 5 for full credit.

If "PRC approved procedure" is used to replace TWO or MORE of 2 - 5, allow 1/2 credit for each substitution greater than one.

REFERENCE S0-0-25 ANSWER 8.06 (2.00)

1. " Hold for Inspection"

2. " Card Pulled"

3. removed from service

4. maintenance order number

5. date and time

6. reason removed form service  ;

Items 1 and 2 may be reversed and items 4-6 may be in any order (0.33 ea)  ;

1 REFERENCE S0-0-40 l

a __6DMINISIBeIIVE_EBQCEDUBESt_CQNDIIIQN3i_eNQ_LIMII8IIQNS PAGE 25 ANSWERS -- FT. CALHOUN -86/03/18-MCCRORY, S.

ANSWER 8.07 (2.00)

The " motherhood" tech spec, 2.0.1(1) applies to this situation. The plant must be in hot shutdown in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (1.0) and subcritical and < 300oF in the noxt six hours (1.0). (There is no need to go to cold shutdown since the tech spec does not apply below 3000F. If the candidate says that the plant must be in cold shutdown in an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> deduct 0.5 pt.)

REFERENCE TS 2.0.1(1), 2.5 j ANSWER 8.08 (4.00)

A. 1. Fuel damage indicated by high RCS activity (>00uci/gm) requiring shutdown.

2. Exceeding primary to secondary tech spec leak rate (>1.0 gpm total)

3. Loss of all of fsite power

4. Failure of a S/G safety or relief valve to shut.

(0.5 each)

B. A Site Area Emergency should be declared (1.0) on the basis of the shift supervisor's application of precautions III.2.u-c. of EPIP-OSC-1. Any one of the three guidelines may be used as j ustification as long as the logic used supports the selection (1.0). (If the candidate uses poor log ic to support his selection of guidelines deduct 0.5 pt.)

REFERENCE EPIP-OSC-1, TS 2.1.3, 2.1.4 4

. o ,

ATTACHMENT l M

Omaha Public Power District 1623 Harney Omaha Nebraska 68102 402/536~4000 Mr. Ralph Cooley Operating Licensing U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76011

Dear Mr. Cooley:

NRC procedures provide licensee representatives an opportunity to review reactor operator examination questions and answers immediately following administration of the examinations. Such an opportunity was provided to the District following the license exams administered in Omaha on November 12, 1985. The District's comments were forwarded to you in my letter LIC-85-532 dated November 20, 1985. Your resolution of these comments was discussed at the recent meeting in Arlington.

There is no question but that NRC examiners make every effort to provide a complete and accurate answer key in response to each set of examina-tions. Similarly, letter LIC-85-532 represents the best efforts of numerous licensed District personnel to correct the NRC answer key where appropriate. However, it is inevitable that experienced examinees will respond to examination questions with correct answers that neither NRC ,

nor District representatives had previously considered. Several such  !

answers were revealed during review of the examinations with the accel-erated retraining class. It is therefore requested that the items below be considered for inclusion into the examination answer keys. In addi- I tion, it is requested that the examinations given on November 12, be regraded where appropriate to reflect any revisions to the answer key.

SR0 REQUALIFICATION EXAMINATION I

Question 6.3A The answer key is correct, but does not completely explain how it is pos-sible to operate the turbine-driven auxiliary feed pump. Examination of P&ID No. ll405-M-252 (copy attached) will reveal that valves YCV-1045A, or YCV-1045B, and YCV-1045 must open before steam can be admitted to the turbine. Emergency Procedure EP-12 (copy attached) states that these valves fail open upon loss of instrument air. This should be added as an additicnal correct answer.

i l

I 45 5124 Employment with Ecual opportunity l Male / Female

(

Question 6.38 The answer key lists pump flow as a single parameter. However, pump flow is actually composed of two separate parameters: condensate, or suction, flow via flow elements FE-1369 and FE-1368 (see P&ID 11405-M-254 attached) and feedwater, or discharge, flow via flow elements FE-Il09 and FE-1110 (see P&ID 11405-M-253 attached). An examinee who names both of these flows should get credit for two parameters. In addition, steam generator level stable or increasing should be considered correct answers. Steam generator level is monitored in the control room. If auxiliary feedwater pump suction were inadequate, steam generator level would decrease. While this is not a direct indication of adequate suction, it is unquestionably an indication nonetheless.

Question 6.4C A second abnormality would result in low surge line temperature indication.

The chemical and volume control system at Fort Calhoun Station uses positive displacement charging pumps. Pressurizer level is controlled by throttling level control valves LCV-101-1 and/or LCV-101-2 (see P&ID No. E-23866-210-120 attached). Suppose a control problem occurred that resulted in the operating LCV-101 valve throttling down to the extent that letdown flow was less than charging flow. This could easily happen since the positive dis-placement charging pumps are constant flow rate. Such an abnormality would cause a net mass addition to the reactor coolant system. The increased mass would result in rising pressurizer level and, since the reactor coolant entering the pressurizer from the hot leg would be at a lower temperature than the coolant in the pressurizer, surge line temperature would drop.

Please note that such an event is not a normally analyzed transient. How-ever, it is certainly a plausible " specific abnormality" which would result in low surge line temperature indication. No other immediate indi-cations would occur provided the operating LCV-101 valve did go completely closed. This scenario should also be considered a correct answer.

Question 6.5B According to Technical Specification 1.3(8), -the axial power distribution trip ensures that the maximum linear heat rate will not exceed 21 kW/ft and that DNBR will not be lower than 1.22. DNBR should not be considered an incorrect answer.

Question 6.5E Technical Specification 1.3(3) (copy attached) describes the following function for the high pressurizer pressure trip: "A reactor trip for high pressurizer pressure is provided in conjunction with the reactor and steam system safety valves to prevent reactor coolant system over-pressure". This should be considered a correct answer.*

\

(

Question 7.2C Answer #1 in the key would place the reactor protective system in a 2-out-of-4 trip condition and thus trip the reactor. However, there are numerous types of events which could cause that second channel to be in the trip condition. For example, the second channel may have been put in the trip condition for surveillance testing or maintenance. The second channel may have been in a trip condition due to equipment failure. Finally, a noise spike could have caused a spurious trip of the second channel. Any of these answers are correct and would, in fact, cause a reactor trip. They should be considered correct answers.

Question 7.3A.1 Rapid aepressurization of the RCS to 215 psia in the event of a leak is a result _ of the pressurizer being at 388 F., not the reason for it being at 388 F. The pressurizer is heated to 388 F. for the purpose of forming a steam bubble and providing adequate NPSH prior to start of the first RC pump. This is reflected by paragraph 2 of the OI-RC-2B " PURPOSE", attached.

Indeed, the pressurizer is not maintained at 388 F. during leak testing.

It is maintained at 460 F., which would allow depressurization to 465 psia (see p.10 of OI-RC-28, attached).

Question 7.4B Placing a properly borated ion exchanger on line as described in the answer key is certainly a correct solution. However, it is more likely that the operator would rinse in the existing ion. exchanger as described in 01-CH-2 (attached) and place it back in service. This evolution can be done quickly enough that there would be no need to place a second ion exchanger in service during the interim. Either answer should be considered correct and receive full credit.

Question 7.68 Please refer to P&IO No. E-23866-210-130, attached. There are many correct answers in addition to those in the key. It all depends on the cause be-hind loss of shutdown cooling. Shutdown cooling suction valves HCV-347 and HCV-348 close automatically on high pressure. If a spurious signal closed these valves, appropriate corrective action would be to open them again.

If a low pressure safety injection pump tripped, it should be restarted or the second pump started. If shutdown cooling were lost due to inadvertent closure of HCV-335, HCV-335 should be reopened. The examinee should get full credit for any answer which would restore shutdown cooling since the question does not describe the reason shutdown cooling was lost. Emergency Procedure EP-37B does not consider all possibilities and correct answers should not be limited to only those described in that EP.

( .. .

Question 8.1A The answer key is correct. However, a second answer is also correct.

Step 5.9 of Standing Order M-28 (copy attached), states: "The instru-ment and control and electrical field maintenance supervisor is respon-sible for test equipment calibration ...". Either answer should be accepted.

Question 8.2B One of the answers in the key is " Estimate or calculate the dose received based on conditions". Please review John Tesarek's answer. The procedure he described accomplishes exactly the function stated in the answer key.

As a matter of fact, it is difficult to envision how the dose could be estimated or calculated without performing the procedure Mr. Tesarek describes. It is requested that he be awarded full credit for his answer.

Question 8.2C i

The applicable paragraph in Standing Order T-10 (copy attached) does not allow the individual to stay in the area and continue working once he has reached an indicated dose of 2400 mrems. Paragraph 2.4.3 (copy attached) of the Radiation Protection Manual states that an individual's quarterly exposure must not exceed 2400 mrem. An individual with 2400 mrem on his dosimeter would be considered to have reached this limit until his TLD could be read. He would have to leave the controlled area immediately.

Also, the RWP form (attached) explicitly limits a person's dose to 2400 mrem per quarter. The answer key should be corrected accordingly.

Question 8._3A Paragraph 4.2 of Standing Order G-23 describes a condition in which the procedure has become radioactively contaminated or suffered other damage.

A copy of this paragraph is attached. However, the question describesThe

~

a condition in which the procedure becomes "... wet and unuseable".

examinee could easily interpret this to mean that the data and verifica-tion initials were also illegible. It's certainly reasonable to assume that the ink would run when the page became wet. In such a case, the only acceptable corrective action would be to reperform the procedure.

Since the question does not describe whether the procedure is legible or not, "reperform the procedure" should also be considered an acceptable answer.

Question 8.3C An additional correct answer is Technical Specifications.

I - a Question 8.48 All of the following are required to ensure that the DNBR limit is not violated. (1) Cold leg temperature less than or equal to 545 F.;

(2) Pressurizer pressure greater than or equal to 2075 psia; (3) Reactor coolant flow greater than or equal to 197,000 gpm; (4) Axial shape index less than or equal to Tech. Spec. Figure 2-7; (5) CEA insertion These are described in Technical Specification Sections 2.10.2 and 2.10.4 (copy attached). In addition, the TM/LP trip is provided to limits.

prevent operation when the DNBR is less than 1.22.

This is described in paragraph 1.3(4) of the Technical Specifications (copy. attached).

All six items should be required for full credit. Each answer should be worth 1/6 of a point.

Question 8.6 A sixth correct answer described on page B-8 of the radiological emer-gency response plan (copy attached) is "on duty shift supervisor shall announce the location and nature of the emergency on the public address system (gai-tronics)". Any five of the six correct answers should receive full credit.

R0 REQUALIFICATION EXAMINATION Question 2.10 An additional correct answer should be ERF or Emergency Response Facility Computer. The rod position is displayed on CRT display #341. A copy of this display is attached.

Question 3.1 An additional correct answer should be low steam generator water level.

Reference Technical Specification 2.14(6), copy attached.

Question 3.4A Rotating inertia of the reactor coolant pump motor supplies energy to the corresponding 4160V Bus during the dead time of a fast transfer of elec-trical power. This is described in System Description I-4, paragraph 2.3.1. A copy of this page is attached. This should be considered an additional correct answer.

Question 4.18 An additional correct answer should be: " Decrease in reactor vessel level." Vessel level monitoring capability is a NUREG 0737 requirement.

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f, Question 4.6B Step D.1. states:

The answer key references Emergency Procedure EP-378.

" Determine if component cooling water is available to theLoss shutdown cooling of component heat exchangers. Switch heat exchangers if necessary.

Emergency Procedure EP-8 states in cooling water is discussed in EP-8.""If component cooling water cannot be re paragraph D.5.: raw water for those components so equipped by opening the AC/R valves, as required." Therefore, restoration of component cooling water flow or cross connect to raw water system should be considered correct The applicable answers in addition to those stated in the answer key.

pages of EP-378 and EP-8 have been attached.

Thank you once again for the opportunity to comment on the examinations and answer keys. Please contact Jay Fluehr should you have questions or need additional-information.

Sincerely, R. L. Andrews Division Manager Nuclear Production RLA/JJF:rge cc: J. J. Fluehr III e