ML20206R923

Exam Rept 50-315/OLS-86-02 on 860519-22.Results of Exam:All Senior Reactor Operator Candidates & Seven Reactor Operator Candidates Passed

ML20206R923
Person / Time
Site:	Cook
Issue date:	06/30/1986
From:	Burdick T, Higgins R, Isaksen P, Jensen N, Picker B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206R882	List: ML20206R877 ML20206R923
References
50-315-OLS-86, 50-315-OLS-86-0, NUDOCS 8607070257
Download: ML20206R923 (96)

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U.S. NUCLEAR REGULATORY COMISSION i

REGION III

[ Report No. 50-315/0LS-86-02 4

Docket No. 50-315 License No. DRP-58 No. 50-316 License No. DRP-74 Licensee: Indiana and Michigan Electric Company Facility Name: D. C. Cook huclear Plant, Unit-1 and 2 Examination Administered At: D. C. Cook Nuclear Plant, Unit 1 and 2 i Examination Conducted: Ma 19-22, 1986 W *'f g/flp#

Examiners: R. iggins .

Date

{ N Date

, en e IN Date k bl-Mlfd l Date

Approved By

/ t Chief T.j' urdick, NM

Operator Licensing Section Date

Examination Summary Examination administered on May 19-22, 1986 (Report No. 50-315/0LS-86-02)

Administered written exams to eight upgrade senior reactor operator. candidates and ten reactor operator candidates. Administered oral exams to eight senior.

reactor operator candidates and eleven reactor operator candidates.

Results: All senior reactor operator candidates and seven reactor operator candidates passed.

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REPORT DETAILS

1. Examiners R. Higgins, Chief Examiner T. Burdick-P. Isaksen N. Jensen B. Picker

2. Examination Review Meeting i

No longer conducted. A copy of both the SR0 exam and R0 exam, along with each respective answer key, were left with the facility. The specific facility comments, followed by the NRC responses are listed in the following paragraphs.

Facility Comment 2.02 Item 2 in the keyed answer is commonly referred to as Thrust Bearing Wear.

In Unit 1 this is measured by the Thrust Bearing Oil Pressure, in Unit 2 it is measured by a feeler foot. We request that either " Thrust Bearing Wear" or " Thrust Bearing Oil Pressure" be accepted for full credit.

Reference:

Description Article DDC-2-1-14 page 2, R0-C-PG10-TP-8 NRC Resolution Answer key modified to accept " Thrust Bearing Wear" as an alternate i correct answer for item 2. Additional facility-supplied reference added to original reference.

r Facility Comment 2.12.a The flow path is verified each shift in Modes 1, 2, and 3 per OHP-4030.STP.030. The position is verified by checking annunciator drops provide independent valve position indication. We request that this position verification be accepted as an alternate correct answer.

Reference:

0HP-4040.STP.030 Data /Signoff Sheet 6.3, Item 9 OHP-4024.106.029 NRC Resolution Answer key modified to accept " flow path is verified once each shift by checking annunciator drops to verify valves are still fully open" as an alternate correct answer. Additional facility-supplied references added to original reference.

Facility Comment 2.13 The "use" stated in part 3 of the keyed answer is true following a 2

Containment Spray and not necessarily "following a Safety Injection".

Additionally, statements describing the "uses" of RHR could include " Cold Leg Injection", " Cold Leg Recirculation", and " Hot Leg Recirculation".

We request that these statements be accepted as alternate correct answers.

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Reference:

ECCS System Description HP-111 page 5. Applicable statements highlighted and noted.

NRC Resolution Existing item 1 on answer key is " Cold Leg Injection". " Cold Leg Recirculation" and " Hot Leg ReEirculation" added to answer key as items 4 and 5, and any 3 required for full credit. Additional facility-supplied reference added to original reference.

Facility Comment 3.03 Page 10 of R0-C-NS06 incorrectly identifies the 7% alarm as a " Low-Low".

This is a " Low" alarm as indicated on OHP-4024.109 Drop 49.

Reference:

0HP-4024.109 Drop 49 NRC Resolution Answer key modified to identify the 7% alarm as a " Low" alarm vice

" Low-Low" alarm. Additional facility-supplied reference added to original reference.

Facility Comment 3.04 This question is open-ended as described in NUREG 1021 ES-202 part E-18 in that it asks "when" but is keyed for two (2) specific sets of conditions. Since the question does not elicit multiple responses, we request that either condition be accepted for full credit. (Note also U-2 temperature requirements for PORV operability is 152 F.)

NRC Resolution Answer key modified to accept:

1. "WhenoperatinginMode1,25 or 3" or

2. " Mode 5, "< 170 F U-1 ( < 152 F U-2),"Tf a PORV is required to be operable .

Either answer worth full credit.

Facility Comment 3.05 b.

Since the question states that the plant is at 100% power, Tave would be the full power Tave of 567 (574) for Unit 1 (2). The candidate may not consider this an "0THER c.ondition". We request that "Tave > 541 F not be required for full credit. Since only one " arming signal" is required to unblock some or all steam dumps, we request that any one of the following conditions be accepted for full credit.

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10% load rejection (C-7A) or 50% load rejection (C-78) or Turbine Trip NRC Resolution The plant may (or may not) actually be at rated Tavg for 100% power, but this condition does not preclude the requirement that the Tavg interlock be satisfied for unblocking the arming valves. Although a step load decrease > 10% or turbine trip is required to complete the arning signal, neither is required if the other exists, so both parts must be ' cluded for full credit. Answer key remains unchanged.

Facility Comment 3.10 Revision 0 (7-18-85) for R0-C-NSO4 states that Control Bank D withdraw stop (item #7 in answer key) is 225 steps. In the interim Plant Modifications were completed that changed the setpoint to 230 steps.

Attached are copies of the annunciator drops for both units which reflects the latest change. We request that either 225 or 230 steps be accepted for full credit.

References:

1-0HP-4024.110.040, CS-8 2-0HP-4024.210.040, CS-8 NRC Resolution Answer key modified to 230 steps reflecting actual plant configuration.

225 steps not accepted since this is no longer current. Additional facility-supplied references added to original reference.

Facility Comment 3.12 The stated reference does not provide the requested information. Rod insertion limit circuitry does not use Auctioneered High Tave as stated in the key. Loop gT is used. We request that " Rod insertion limit comparators" be deleted from the key and item 1 under Individual Loop Tave be changed to OT Delta-T.

Reference:

R0-C-NS04, Rev. 9, page 11 NRC Resolution Answer key modified to delete " rod insertion limit comparators".

Typographical error in item 1 under individual Loop Tavg corrected to "0T Delta-T". Point values redistributed to reflect deletion of " rod insertion limit comparators. Reference corrected to R0-C-NS04, Rev. 0, page 11. The facility is advised to update its literature by correcting erroneous information.

Facility Comment 3.13 Since DNB is a condition which is a result of an accident and not an accident itself, we request it not be required for full credit.

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Additionally, P-11 which allows manual block of low pressure SI should be included.

Reference:

R0-C-NS12, Rev. 1, page 15 Functional Diagram 98503 NRC Resolution Low pressure SI helps to ensure DN8 limits are not exceeded under conditions in addition to LOCA and steam break. However, since the possibility exists for confusion, DN8 protection is removed from the answer key. The P-11 permissive is added, and point values remain the same.

Facility Comment 4.01 The last phrase of the keyed answer was taken from the AIDS column of the lesson plan. This information is intended as amplifying information for the instructor and should not be expected in a response. We request that "and the overcurrent trip on the RCP tie is set lower than the diesel" not be required for full credit.

References:

R0-C-AS10, Rev. 9, page 36 Tec. Spec. Clarification No. 15, Rev 2 NRC Resolution If the overcurrent trip setpoints were the same, or the normal feed breaker setpoint was higher than the DG breaker, then the precaution would be meaningless. To assure full understanding of the reason for this precaution, the response must include correct reference to the different trip setpoints. Answer key remains unchanged.

Facility Comment 4.02

a. Section 4.2 of the procedure states that the flow "should be equal to or less than design flow 5000 gpm". The fact that 5000 gpm is the design limits does not reflect a REASON. We request that the value alone be accepted for full credit for the MAXIMUM CCW flow.

b. The question strongly suggeste that a VALUE is expected for the MINIMUM flow, when in fact no value is specified. This portion of the question should be deleted since the wording leads candidates to come up with a number and justify it.

Reference:

0HP-4021.027.002 Step 4.2 and note.

NRC Resolution

a. If no other specific limit applies, and plant literature makes reference to ." design limits" or " design criteria", then it is commonly accepted industry-wide practice to utilize " design limits" as reasons for maximum and minimum values. . Answer key remains unchanged.

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b. A "value" may be a fixed numerical value or one which varies based on other parameters, as is the case here. It is extremely important that an operator understand the minimum flow rate required, and the reasons for it, as specifically spelled out in the procedure in question. However, will accept for full credit the reason for the minimum flow, and its "value" will not be required for full credit.

Points will be deducted if the value which the candidate gives for minimum flow is greater than or equal to the maximum flow value.

Facility Comment 4.07 This question / answer requires the candidates to have memorized a value of hydrogen concentration identified in a procedural step which is not an immediate action. We request that > 6.0% not be required and answers which indicate high hydrogen concentration be accepted for full credit.

Reference:

1-0HP-4023.E-1 Step 11.a.

NRC Resolution Answer key modified to accept " hydrogen concentration is in the explosive range (> approximately 4% and < approximately 90%)."

Facility Comment 4.08 Item 4 of the answer key is not elicited by the question. The question requires a response for each color but does not ask for requirements for addressing the other paths after initiating action. We request that item 4 not be required for full credit.

NRC Resolution Answer Key modified to only require original item numbers 1, 2, 3, and 5 for full credit. Point values redistributed so each correct response is worth 0.5 (vice 0.4) points.

4 Facility Comment 4.14 1

Part b requires candidate to state when the turbine must be tripped. The keyed answer reflects the Unit 2 procedure. The Unit 1 procedure does not specify a value. Since the question did not specify a unit, we request that " prior to the low vacuum trip" or equivalent statements be accepted for full credit.

Reference 1-0HP-4022.053.001 2-0HP-4022.053.001 NRC Resolution Answer key modified to read: "21.9" HG and still decreasing (U-2), prior to low vacuum trip (U-1), OR: Prior to low vacuum trip (no unit specified)." Either response accepted for full credit.

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Addition NRC Resolution (s) of Answer Key 2.15 Added choice number 7 to existing list of 6 correct responses: "7.

Output breaker not in lockout."

3.05 Item 1.3: Placed "--Unit 1 only" in parenthesis, because the modification requiring an output voltage from CRID II is being installed currently on Unit 2.

3.13 Item 4.b: Deleted "0R high steamline flow coincident with low low Tavg" from the answer key. This was incorrect.

2.03 Added #5 as an additional correct answer: " Provide vent path during RCS fill and vent, and for degasing." Added referenced OHP-4021.002.001 and OHP-4021.003.005.

2.04 Added " Lose of Instrument Air" as an additional correct answer.

Added reference drawing number OP-1-5129-12.

Facility Comment 5.17 The question asks for an approximate value, but the answer key specifies a band of acceptable values only i 2%. We request that the band be expanded to at least i 5% which would allow answers 245 i 12 psig.

NRC Resolution Answer key modified to accept +/-10 psig, any broader acceptance band could preclude corrections for gauge to absolute pressure.

Facility Comment 6.01 Item 2 in the keyed answer is commonly referred to as Thrust Bearing

Wear. In Unit 1 this is measured by the Thrust Bearing 011 Pressure, in Unit 2 it is measured by a feeler foot. We request that either " Thrust Bearing Wear" or " Thrust Bearing Oil Pressure" be accepted for full i credit.

Reference:

Descriptive Article DDC-2-1-14 page 2, R0-C-PG10-TP-8.

NRC Resolution Answer key modified to accept " Thrust Bearing Wear" as an alternate correct answer for item 2. Additional facility supplied reference added to original reference.

Facility Comment 6.04 The "use" stated in part 3 of the keyed answer is true following a Containment Spray and not necessarily "following a Safety Injection".

Additionally, statements describing the "uses" of PHR could include " Cold Leg Injection", " Cold Leg Recirculation", and " Hot Leg Recirculation". We request that these statements be accepted as alternate correct answers.

Reference:

ECCS System Description HP-111 page 5. Applicable statements highlighted and noted.

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NRC Resolution Existing item l'an answer key is " cold leg injection". " Cold Leg Recirculation" added to answer as items 4 and 5 respectively and 3

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4 required for full credit. Additional facility supplied reference

material added to original reference.

! Facility Comment 6.08 b Since the questions states that the plant is at 100% power, Tave would be the full power Tave of 567 (574) for Unit 1 (2). The candidate may not-consider this an "0THER CONDITION". We request that "Tavg > 541*F not be required for full credit. Since only one " arming signal" is required to 4

unblock some or all steam dumps, we request that any one of the following conditions be accepted for full credit.

10% load rejection (C-7A) or 50% load rejection (C-78) or Turbine Trip NRC Resolution I

The plant may (or may not) actually be at rated Tavg for 100% power, but i this condition does not preclude the requirement that the Tavg interlock be satisfied for unblocking the arming valves. Although a step load

decrease > 10% or turbine trip is required to complete the arming signal,

neither is required if the other exists, so' both parts must be included for full credit. Answer key remains unchanged.

J Facility Comment'6.11 Revision 0 (7-18-85) for R0-C-NS04 states that~ Control Bank D withdraw stop (item #7 in answer key) is 225 steps. In the interim Plant 1 Modifications were completed that changed the setpoint to 230 steps.

3 Attached are copies of the annunciator drops for both units which l reflects the latest change. We request that either 225 or 230 steps be j accepted for full credit.

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References:

1-OHP-4024.110.040, CS-8 2-0HP-4024.210.040, CS-8 l NRC Resolution Answer key modified to 230 steps reflecting actual plant configuration,

) 225 steps not accepted since it is no longer correct. Additional

, supplied facility reference added to original reference.

J i Facility Comment 6.13 i

Since DNB is a condition which is a result of an ' accident and not an l accident itself, we request that it not be required for full credit.

Additionally, P-11 which allows manual block of low pressure SI should be included, j

Reference:

R0-C-NS12, Rev. 1, page 15 Functional Diagram 98503

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l NRC Resolution I

Low pressure SI helps to ensure DNB limits are not exceeded under conditions other than just LOCA and steam break. However-since the

possibility exists for confusion DNB protection is removed from the i

answer key. The P-11 permissive is added and point values remain the I

same.

i Facility Comment 6.15 I

This question is open-ended as described in NUREG 1021 ES-202 part E-18 i _in that it asks "when" but is keys for two (2) specific sets of conditions. .Since the question does not elicit multiple responses, we

request that either condition be accepted for full credit. (Note also i U-2 temperature requirements for PORV operability is 152*F.)

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! NRC Resolution

Answer key modified to accept either condition for full credit.

Facility Comment 7.01 j OHP-4022.013.001, " Source Range Malfunction" Step 4.2.1 states "If Reactor Startup is in progress, stop rod withdrawal and insert all Full i Length Control Rods". In that T.S. 6.8 required that written procedures "be established and implemented", we request that (a) be accepted as an

alternate correct answer.

Reference:

OHP-4022.013.001 Step 4.2.1 T.S. 6.8.1 j NRC Resolution Although question clearly states TS actions answer key modified to accept 4

"a" as additional correct response (d or a) since "a" is procedurally '

correct and purpose of the question was to test knowledge and not the source TS or procedure.

! Facility Comment 7.05 ,

For D. C. Cook, ADVERSE CONTAINMENT conditions are:

. Containment Pressure > 1.1 psig 1 Containment Radiation > 510 R/hr 6

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! Containment Integrated Dose > 10 Rad i

Reference:

M. P. Alexich letter AEP:NRC:0785C a Attachment 1, page 1 i i j

NRC Resolution i Answer was a "CAF", above is incorporate into the answer key and j reference since informtion was not included in initial facility-j information sent for exam preparation.

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Facility Comment 8.05 Using Table 6.2-1 of T.S. 6.2 (U-1 and U-2 attached), the correct answer is 1 SOL (SS) and 4 OL. 0HI-4011 (which establishes "the guidelines for shift staffing to meet the requirements of.10CFR50.54 and Technical Specification 6.2.2) requires 1 SS (SR), 2 US (SRO) and 4 R0, which is item (b) of the' key. Since item (c) could be interpreted to be correct per_ T.S. and item (b) reflects requirements of 10CFR50.54 and OHI-4011, we request that 'either (b) or (c) be accepted for full credit.

Reference:

T.S. Table 6.2-1 U-1, U-2 (Attached)

OHI-4011, page 2, Section 3.1.7 10CFR50.54 NRC Resolution Answer key modified to accept either b or c and additional OHI-4011 reference added to original reference. Since original facility supplied reference material (TS) was incorrectly marked-up which led to confusion and need for_ candidates to have to " interpret the correct response.

Facility Comment 8.09 a The Unit 2 T.S. Basis lists two additional reasons: 4) The pressurizer is capable of being in an operable status, and 5) The reactor vessel is above minimum R NDT.

We request that these be accepted as alternate correct answers

Reference:

U-2 T.S. Basis page B 3/4 1-2 NRC Resolution Answer key modified to accept the two additional correct responses, 4 and 5 above and 3 of 5 required for full credit.

Facility Comment 8.13 OHI-4013 paragraph 4.2.p on page 9 states that the Assistant Shift Supervisor assumes the duties of the Shift Supervisor during absences or anytime should the Shift Supervisor become incapacitated. Paragraph 4.3.s on page 12 indicates that a Unit Supervisor would act as the Assistant Shift Supervisor iflneeded.

PMP-2080.EPP.015 Section 4.2.1 delineates personnel who may relieve the OSEC. The following (or similar) answers for part (b) are acceptable for full credit:

The Assistant Shift Supervisor _(A.S.S) or The Most Senior License Available or The personnel delineated in Section 4.2.1-of EPP.015

References:

OHI-4013, page 9 12 PMP-2080.EPP.015, page 2 10

NRC Resolution Answer key modified to accept Assistant Shift Supervisor or most senior SR0 available, since this was a CAF. Thequestionclearlycallsforthe INITIAL person if the SS is incapacitated not who can relieve the OSEC (SS). Therefore, the personnel delineated in section 4.2.1 of EPP.015 are not acceptable as a correct response.

Additional NRC Resolution of Answer Key 5.01 Modified value of Beff to BOL value of 700 PCM changing answer to 0.96 DPM vs 1.20 DPM since question stated BOL.

6.02 Accepted loss of control air as additional correct response, answer key modified to 5 correct responses 4 required, 0.25 each.

3. Exit Meeting An exit meeting was conducted on May 22, 1986, with facility representatives and the chief examiner to discuss any generic weaknesses noted during the examination as well as any overall impressions of the-facility. The following facility representatives attended the exit meeting.

B. Smith, Plant Manager L. Matthias, Administrative Superintendent W. Nichols, Training Manager D. Strasser, Operations Training Supervisor W. Davidson, Replacement Senior Instructor L. Smith, Shift Supervisor The NRC representative was Ronald L. Higgins, the chief examiner.

The topics discussed during the exit meeting are listed in the following paragraphs.

a. The facility was commended for the cooperation accorded the examiners during the conduct of the examinations. The training staff is particularly commended, since INP0 accreditation was being conducted at the same time NRC examinations were being administered, placing added burdens on the training department staff.

b. Plant cleanliness had improved significantly, although the auxiliary building seemed cluttered with low level radioactive waste receptacles and holding areas.

c. The facility was advised regarding the submittal of license

. applications which are not adequately completed and which do not document sufficient experience to allow the applicant to take the exam. Facility representatives acknowledged the problem and stated that most of the difficulty arose because of their unfamiliarity with the newly-revised NRC Form 398. They indicated that such problems would not happen in the future.

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d. The facility was advised regarding the submission of inadequate material in a form which was very difficult to use (unbound, with no tabs or indices). Facility representatives agreed to permit the NRC examiner for D. C. Cook to scrutinize the indexes for all facility literature prior to the next. exam to ensure that the NRC receives adequate material. The facility representatives apologized for the condition of the material sent to the NRC, explaining that much of the material was newly revised and had been hastily assembled and shipped in order to meet the NRC deadline. The facility representatives were informed that if, in the future, examination material was sent to the NRC which was not in binders, tabbed and indexed, the material would be returned to the facility C.O.D.~ and the examinations may be postponed and rescheduled after the NRC received satisfactory material.

e. The intermediate range meter calibration requirements were addressed.

f. Plant personnel were observed disregarding "no smoking" signs and not wearing safety glasses or ear plugs when required.

g. Handheld equipment being removed from the auxiliary building is not surveyed for radioactive contamination by radiation protection personnel unless the individual carrying the equipment requests the survey.

h. One examiner could not find procedures for operating one main feed pump in auto and the other main feed pump in manual. The facility representatives believe that such a procedure is available, but will investigate,

i. The new Emergency Response Guidelines do not specify what " adverse containment" is, and the folders in which they are kept are becoming soiled and torn. Some guidelines are in danger of having pages fall out.

j. The procedure for approving jumpers is unclear. Facility representatives stated that a new jumper authorization procedure was being implemented.

k. Some SR0's seemed unknowledgeable about refueling procedures and requirements.

1. Some SR0's did not know what a " noble gas" was, r what " tritium oxide" was.

m. Some examinees were unfamiliar with the inputs to the subcooling monitor and the operability requirements for channel checks.

n. Some examinees did not know what happens if CRID is lost below P-8.

There is no procedure written to address this situation.

o. Some examinees did not know how rods would respond in automatic if a power range nuclear instrument failed high.

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p. Some examinees were not very knowledgeable about the immediate actions of abnormal procedures.

q. Some examinees demonstrated defic'encie. in the use of Emergency Response Guidelines. Given a h'gh radiation a'ic 7 on the steam jet air ejector, one examinee went ight to E-3 without going to E-0.

Several other examinees forgot to restart ECCS pumps upon a loss of offsite power during a steam generator tube rupture after SI had been reset.

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U. S. NUCLEAR REGULATORY COMMISSION  !

REACTOR OPERATOR LICENSE EXAMINATION

" i= 0vs a FACILITY: COCK ~&2 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 86/05/19 EXAMINER: JENSEN. N.

APPLICANT: ___

INSTRUCTIONS TO APPLICANT:

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

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 804. Examination papers will be picked up six (6) hours after the examination starts.

% OF 4 CATEGORY  % CF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 .1. PRINCIPLES OF NUCLEAR EOWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 3. INSTRUMENTS AND CCNTROLS 25.00 _25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 100.00 TOTALS FINAL GRADE  %

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

APPLICANT'S SIGNATURE

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1.

PRINCIPLES OF NUCLEAR POWER PLANT OPEBATIClL. PAGE 2

, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW MASID COPY CUESTION 1.01 (3.00)

Explain HOW and WHY the following change with moderator temperature:

a. Rod worth

b. Differential boron worth QUESTION 1.02 (1.50)

MULTIPLE CHOICE.

Following a reactor trip from 100% power, how long should it take (cpproximately) before the source range instrumentation should be automatically energized?

a. 9 minutes,

b. 14 minutes,

c. 23 minutes,

d. 40 minutes.

o. 55 minutes.

QUESTION 1.03 (1.50)

Explain why the equilibrium value of Xenon :r.ust increase with power while the equilibrium value of Samarium stays constant with power.

QUESTION 1.04 (2.00) n.

Why does nucleate boiling heat transfer remove more heat than non-boiling heat transfer?

B

b. Why does film boiling remove less heat than nucleate boiling? i 1'

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

1. FRINCIELES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW yASER C:PY QUESTION 1.05 (3.00)

How will the following affect the Moderator Temperature Coefficient?

BRIEFLY EXPLAIN your answer.

a. The BIT is inadvertantly injected into the RCS.

b. The core ages from BOL to EOL.

c. The RCS is cooled down from 550 F to 450 F.

QUESTION 1.06 (1.50) .

a. Which parameter below will have the MOST effect on the shape of a Differential Rod Worth Curve?

1) Core radial flux profile

2) Core axial flux profile

3) Core axial temperature profile

4) Time of core cycle (0.5)

b. What effect does having a bank overlap program have on the differential rod worth curve? (As compared to a differential rod worth curve for a rod program that does not use bank overlap.) (1.0)

QUESTION 1.07 (3.00)

Hot channel factors are measurable and their Technical Specification surveillance frequency requirements are relatively low provided four items are monitored and verified to be within their limits. Provide THREE of these four items (conditions).

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

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1. PEINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 4

, IEEEM0 DYNAMICS. HEAT TRANSFER AND FLUID FLOW NASTl 0:PY QUESTION 1.08 (3.00)

c. A variable speed centrifugal pump is operating at 1/4 rated speed in a CLOSED system with the following parameters:

Power = 300 KW Pump delta P = 50 psid Flow = 880 gpm j What are the new values for these parameters when the pump speed is increased to full rated speed? (1.5)

b. Choose the answer that most correctly completes the sentence.

'In a CLOSED system, two single stage centrifugal pumps operating in parallel will have--(choose-from-below)- , as compared to the same system with one single stage centrifugal pump operating with one pump isolated."

1. a higher head and higher flow rate.

2. the same head and the same flow rate.

3. the same head and a higher flow rate.

4. a higher head and the same flow rate. (0.5)

c. How is the available NPSH affected by an increase in system flowrate? (0.5)

d. Why is cavitation undesirable? (0.5)

QUESTION 1.09 (1.50)

a. Does Beta Effective Increase, Decrease, or Remain the Same, from BOL to EOL7 EXPLAIN YOUR CHOICE. (1.0)

b. For two equivalent positive reactivity additions to a critical reactor, will the SUR be the Same, Larger, or Smaller at EOL as compared to BOL7 NO EXPLANATION IS NECESSARY. (0.5)

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

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1. PRINCIPLES OF NUCLEAR POWER PLAN'T OPERATION. PAGE 5 IBEEMODYNAMICS. HEAT TRANSFER AND FLUID FLOW MASIS CC0Y QUESTION 1.10 (1.00)

High energy neutron exposure increases the possibility of brittle fracture of the reactor vessel by increasing the:

n. nil ductility temperature of the reactor vessel.

b. plastic deformation of the reactor vessel.

c. compressive stress on the reactor vessel.

d. magnitude of the failure stress of the reactor vessel.

QUESTION 1.11 (2.25)

a. In what direction does extraction steam pressure (and flow rate) vary as plant power is increased? How (increase / decrease / remain the same) is plant efficiency affected by this variation? BRIEFLY explain.

(1.25)

b. Explain how increasing condenser circulating water flow may act to improve OR decrease overall plant efficiency. (1.0)

QUESTION 1.12 (1.75)

c. Positive displacement pumps can be classified as either recip-rocating or rotary. What are 2 of the 3 types of Rotary Posit-ive Displacement pumps ? (1.0)

b. The ideal performance curve for a positive displacement pump would be a straight line. Why is this not true in an actual sit-uation ? (0.75)

(***** END OF CATEGORY 01 *****)

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 NASIMOPY QUESTION 2.01 (1.00)

Describe the purpose of the Reheat Stop and Intercept valves installed upstream of the LP turbine.

QUESTION 2.02 (2.00)

List EIGHT of the nine conditions which will cause main feedwater pumps to trip. (Setpoints not required.)

QUESTION 2.03 (1.50)

Describe THREE purposes of the pressuriser power operated relief valves.

QUESTION 2.04 (1.00)

Uhat are FOUR conditions which cause automatic closure of the CVC.c orifice isolation valves QRV-160, -161, and -162?

QUESTION 2.05 (1.00)

What is the function of the RCP oil lift system?

QUESTION 2.06 (1.00)

Describe TWO methods of detecting in-leakage into the component cooling water system from the residual heat removal system.

QUESTION 2.07 (1.00)

State the THREE conditions which cause automatic start of the Essential Service Water pumps, l

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 O

l QUESTION 2.08 ( .50) Wd-During operation at 80% power with control rods at 200 steps, steam.

generator safety valve SV1A-1 fails full open. How much additional load does this place on the reactor?

QUESTION 2.09 (1.50)

List the loads on each SAFEGUARDS train which are cooled by CCW.

QUESTION 2.10 (1.00)

The safety-related function of the condensate storage tank is to store at least gallons of water, to ensure hot standby conditions can be maintained for at least hours following a total loss of offsite power, with steam discharge to atmosphere.

QUESTION 2.11 (3.00)

a. To which of the four steam generators do each of the auxiliary feedwater pumps normally discharge for Unit 1?

b. Describe the design feature which allows auxiliary feedwater makeup to each Unit 1 steam generator, if Unit 1 auxiliary feedwater pumps are disabled.

QUESTION 2.12 (3.00)

Concerning the ECCS accumulators:

O. How is assurance made that a flow path will be available through the discharge isolation valves in mode 1? (0.5)

b. List the limits on contained water volume, boron concentration, and pressure for each accumulator for mode 1 operation, including unit differences. (2.5)

QUESTION 2.13 (1.50)

What are THREE uses of the RHR system following a Safety Injection?

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

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 kASIB CCPY QUESTION 2.14 (2.00)

Please indicate TRUE or FALSE.

a. During safety injection, the RHR pump suction lineup automatically shifts to the recirculation sump when RWST level decreases to

</= 32%.

b. During safety injection, a running RHR pump will stop if RWST level drops to </= 9.1% and its respective suction valve from the RWST (IMO-310 or -320) is open.

c. Following a LOCA, both RHR pumps must operate to provide adequate flow to the reactor for ensuring ECCS design criteria of 10CFR50 are met.

d. During the recirculation phase of safety injection, the RHR heat exchanger inlet crosstie valves are opened to provide assurance that flow can be routed through at least one heat exchanger.

QUESTION 2.15 (2.00)

State FIVE of the six signals required for the diesel generator output breakers to automatically close during an emergency without synchronizing or check synchronizing. Include values and/or setpointo as applicable.

QUESTION 2.16 (2.00)

Mhat design features prevent inadvertent draining of the Spent Fuel Pit? l (FOUR required.)

(***** END OF CATEGORY 02 *****)

S

3. INSTRUMENTS AND CONTROLS PAGE 9 NAH MOY QUESTION 3.01 (1.50)

DESCRIBE the interlock asscciated with the containment spray pump cuction valves IMO-215 and IMO-225, and BRIEFLY EXPLAIN why it is installed.

QUESTION 3.02 (2.00)

Describe TWO pressuri=er heater control functions performed by the pressuriser level instrument, and BRIEFLY EXPLAIN why each is designed into the circuit.

QUESTION 3.03 (1.50)

List the SEVEN alarm and control functions of the VCT level instrumentation cystem. Setpoints required.

QUESTION 3.04 (1.00)

When is reactor coolant wide range pressure instrumentation required to ba operable?

QUESTION 3.05 (2.50)

c. What conditions must be met to activate the Condenser Vacuum Permissive (C-9)? Include setpoints, logic, and unit differences.

(1.0)

b. What OTHER conditions must be met to unblock the steam dump arming valves if the plant is initially operating at 100%

power? Be specific. (1.5)

QUESTION 3.06 (1.00)

STATE the input signals used in the main feed pump speed control circuit, and BRIEFLY DESCRIBE the purpose served by each.

I l

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) I 1

3. INSTRUMENTS AND CONTROLS QUESTION 3.07 (1.50)

a. What parameter (s) determine steam generator program level? (0.5)

b. Describe how program level varies over the range of the controlling variable (s). Include setpoints/ values as applicable.

(1.0)

QUESTION 3.08 (1.50)

Unit 1 is operating at 45% power with all systems in automatic control.

For each condition lisced below, give the initial direction of rod motion, AND state the initial reason for this rod motion.

a. Loop 3 Tc fails high.

b. A main steam power operated relief valve fails open.

c. The turbine is ramped to 100% power at 5% per minute.

QUESTION 3.09 (1.50)

Why is a variable-gain circuit included in the rod control system power mismatch circuitry?

QUESTION 3.10 (2.00)

List all conditions which will prevent rod WITHDRAWAL. Indicate whether cach'is effective in the MANUAL mode, AUTOMATIC mode, or BOTH. Include setpoints as applicable.

QUESTION 3.11 (2.00)

List ALL inputs to RPS/ESF provided by pressurizer pressure instrumentation. Include applicable setpoints and coincidence, and specify unit differences.

i l

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

l I l l \

. 3. INSTRUMENTS AND CONTROLS PAGE 11 wg QUESTION 3.12 (2.00)

Identify the control, protective, and permissive functions which are supplied by the Tavg circuitry. State whether each is supplied by individual loop Tavg or by auctioneered Tavg.

QUESTION 3.13 (3.00)

List all AUTOMATIC SI actuation parameters. Include the accident (s) protected against by each parameter listed, and any permissives cssociated with each. Include unit differences. (Setpoints not required.)

QUESTION 3.14 (2.00)

List ALL automatic starts for the auxiliary feedwater pumps. Include setpoints, coincidence logic, and unit differences as applicable.

e l

(***** END OF CATEGORY 03 *****'

4. PROCEDUEES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 12 RADIOLOGICAL CONTROL

~

k AS"!R CPY QUESTION 4.01 (2.00)

The note preceding step 6.2.1 in OHP 4021.032.001 concerning operation of the Emergency Diesel Generators states: "When paralleled, DGAB is to be paralleled to the TilA bus only and DGCD is to be paralleled to the T11D bus only, ..."

WHY is this limitation imposed?

QUESTION 4.02 (2.00)

With regard to OHP 4021.017.002, " Placing in Service and Operation of Residual Heat Removal Loop":

Supply the VALUE and the REASON for the MAXIMUM and MINIMUM CCW ficw rate through the RHR heat exchanger (s).

QUESTION 4.03 (2.00)

a. Fill in the blank for each unit:

Precaution 4.3 of OHP 4021.002.003 " Reactor Coolant Pump Operation" states, "When any RCS cold leg temperature is </= U1 F

( U2 F), one centrifugal charging pump and the reciprocating charging pump must be racked out as per T.S. 3.1.2.3."

b. WHY is this requirement imposed?

QUESTION 4.04 (1.50)

State all the conditions when a RCP may be operated without seal water, according to OHP-4021.002.003 Reactor Coolant Pump Operation"?

QUESTION 4.05 (1.00)

Per OHP 4021.003.005 "RCS Degassing"' why is there a LOWER limit on pressure in the Volume Control Tank, and what is its VALUE?

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

4. PROCEDURES - NORMAL. AENORITL. EMERGENCY AND PAGE 13

. RADIOLOGICAL CONTROL

,, )

y QUESTION 4.06 (2.00)

According to function restoration procedure FR-S.1 " Response to Nuclear Power Generation /ATWS" Step 2, the operator is to verify turbine trip by observation of turbine steam stop valve closed status lights LIT. If this response is NOT obtained, what steps are specified to manually trip the turbine?

QUESTION 4.07 ( .50)

Acccrding to procedure E-1, " Loss of Reactor or Secondary Coolant" the operator is to " check if hydrogen ignitors should be turned on".

Under what condition (s) should they NOT be turned on?

QUESTION 4.08 (2.00)

Critical safety function status trees include red, orange, yellow, and green paths.

What are the operator's responsibilities with regard to EACH of these paths?

QUESTION 4.09 (2.00)

List the critical safety function status trees, in their ORDER of PRIORITY. (Use No. 1 as the highest priority.)

QUESTION 4.10 (2.00)

What check (s) should be performed on a portable radiation monitoring instrument prior to its use? l

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

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 14

, EADIOLOGICAL CONTROL kAS"3 C:?Y QUESTION 4.11 (3.00)

Indicate whether each of the following statements concerning use of radiation work permits (RWP's) is TRUE or FALSE. (Assume reactor power is 50%.)

c. In certain instances, Radiation Protection personnel may be substituted for a RWP during work performance.

b. A RWP is required for ANY task requiring entry into the Reactor Containment.

c. Except in emergencies, RWP's must be initiated by Radiation Protection personnel.

d. An extended RWP may be authorised for work which is repetitive in nature.

o. An extended RWP may be authorized for work in the regenerative heat exchanger area,

f. Personnel may deviate from the conditions of a RWP if authorized by the Plant Radiation Protection Supervisor or the Shift Supervisor.

QUESTION 4.12 (1.50) ,

The following question refers to OHP 4022.005.002, Emergency Boration:

The rate of reactivity insertion due to emergency boration @ 75 GPM is opproximately 1% delta K/K every 4 minutes. Based on this value, what are the recommended approximate emergency boration times for each of the following situations?

a. Excessive control rod insertion.

b. Failure of a control or shutdown rod to drop following a reactor trip.

c. Uncontrolled reactor cooldown (BOL).

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

l 1

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 15 BADIOLOGICAL CONTROL E R C:PY QUESTION 4.13 (2.00)

The following question refers to OHP 4022.002.009, " Leaking Pressuriser Power Relief Valve", and assumes Mode 1 operation:

Briefly describe the actions the operator will carry out to determine WHICH relief valve is leaking. (Valve numbers not required.)

QUESTION 4.14 (1.50)

a. With regard to OHP 4022.053.001, " Decreasing Condenser Vacuum",

identify FOUR conditions which may cause a slow decrease in vacuum.

b. At what point must the turbine be tripped?

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

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

e

. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 16 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

NAS3 CPY ANSWER 1.01 (3.00)

a. Rod worth increases with moderator temperature [0.5] because as moderator temperature increases, moderator density decreases, which increases the slowing down timelless chance of collision). [0.5]

This allows more neutrons to be captured in resonances. [0.5] (Also since some of the moderator has been removed from the core, the rods see less competition for neutrons. Acceptable, but not required for credit.)

b. Differential boron worth will decrease with an increase in mcderator temperature [0.5] because as moderator temperature increases, moderator density decreases,[0.5] removing boron atoms from the core. Therefore, boron will absorb fewer neutrons.[0.5]

REFERENCE ROTM 3-232, 3-218 WNTO Text, I-5 c

ANSWER 1.02 (1.50)

b. (1.5)

REFERENCE WNPO Text, p.I-3.17 thru 3.19 ANSWER 1.03 (1.50)

Xenon production and burnout are flux dependent while Xenon decay is independent of flux. To keep a balance between removal and production, there must be more equilibrium Xenon to decay. [0.75]

For Samarium (which does not decay), production and removal rates are both flux dependent. At any power level the production and removal rates are the same. [0.75] (1.5)

REFERENCE i WNPO Text, p. I-5.57 thru 5.79 l l

1

4 .

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 17 IBERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- COOK 1&2 -86/05/13-JENSEN, N.

NASER CPY ANSWER 1.04 (2.00)

a. Nucleate boiling creates much more turbulent flow than that which exists during non-boiling conditions, causing better mixing of the hotter water near the clad wall into the cooler main channel of flow. [1.0)

b. In film boiling, a film of steam coats the clad surface and forms an insulating layer (which greatly reduces the heat transfer coefficient). [1.0] (2.0)

REFERENCE ROTM 2-72/2-74 W Thermal Science, p.III-65 thru -90.

ANSWER 1.05 (3.00)

a. LESS NEGATIVE [0.25] More boron to leave core area per degree temperature change. (Or equivalent answer) (1.0)

b. MORE NEGATIVE [0.25] Less boron for opposite result as above. (1.0)

c. LESS NEGATIVE [0.25] Water density changes are less as temperature is reduced. (1.0)

REFERENCE IP-3 ECI Rx Theory; Chapter 5, Pages 21 through 27 WNTO Text, I-5.2 thru 5.16 ANSWER 1.06 (1.50)

a. (2) Core axial flux profile. [0.5]

b. The curve is more linear (due to the additive effect of the rod worths at their low values.) [1.0] (1.5)

REFERENCE IP-3 ECI Rx Theory; Chapter 7, Pages 21, 22, and 27 "NTO Text, I-5.38 thru -5.43

1. PRINCIPLES OF NUCLEAR POWER PLANT OPEBATION. PAGE 18

, IEEEMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

VASI!R CPY ANSWER 1.07 (3.00)

a. Rod groups sequenced and overlapped.

b. Rod insertion limits adhered to.

c. Axial flux difference limits adhered to.

d. Rod group allignment maintained. [1.0 each; 3 required]

REFERENCE TS 3/4.1.3 & 3/4.2.1 ANSWER 1.08 (3.00) 3 3

a. Power (2) = Power (1) * (N2/N1) = 300 * (4) = 19.2 MW (0.5) 2 2 Delta P(2) = delta P(1) * (N2/N1) = 50 * (4) = 800 psid (0.5)

Flow (2) = Flow (1) * (N2/N1) = 880

• 4 = 3520 gpm (0.5)

b. Answer: #1 (0.5)

c. DECREASES (0.5)

d. Pump efficiency and flowrate are reduced and mechanical pump damage (erosion, pitting and vibration) may occur. (0.5)

REFERENCE THF-4 W Thermal Science, p.X-31 ANSWER 1.09 (1.50)

a. Decreases [0.33]. Fu 239 concentration increases (while U 235 concentration decreases) [0.66]. (1.0)

b. Larger SUR. (0.5) e

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 19 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

n REFERENCE l'i Li l V W, I HO-RTR-23 to-27 UNTO Text, p.I-3.10 ANSWER 1.10 (1.00)

a. (1.0)

REFERENCE Thermal-Hydraulic Principles and Applications to the PWR II, p 13-61 WNTO Text, p.XIII-56 ANSWER 1.11 (2.25)

a. Extraction steam pressure and flow rate INCREASE as power is increased, [0.125] causing plant efficiency to INCREASE

[0.125] because of the increase in heat transfer rate across the feed heaters [1.0]. (1.25)

b. Increased heat transferred may result in a lower condenser pres-sure, increasing efficiency. OR; Increasing heat transfer past the point of minimum condenser pressure will subcool the condensate and decrease efficiency.

(Accept either answer for full credit.) (1.0)

REFERENCE W Thermal Science, p.XII-24, -25, -26 ANSWER 1.12 (1.75) a' . 1. Rotary Vane

2. Gear Pump

3. Screw Pump [any 2, 0.5 each]

b. At the pump's design pressure limit back leakage past moving parts will occur and volume flow rate decreases. [0.75] (1.75)

REFERENCE W Thermal Science, p.X-49, -50

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 20

. ANSWERS -- COOK 1&2 . -86/05/19-JENSEN, N.

kAS"iR C ?Y ANSWER 2.01 (1.00)

Limit flow of steam from the MSR's to the LP turbine to prevent the turbine from overspeeding following a load reduction. (1.0)

REFERENCE Cook RO-C-PG2A, Pg. 10 ANSWER 2.02 (2.00)

1. Low vacuum 1

2. Thrust bearing oil pressure low o t-

~

ONS g wea r.

Dod el"g

3. Overspeed

4. Low bearing oil pressure

5. Low pump suction pressure

6. Manual (local and remote)

7. Safety injection

8. S/G hi-hi water level

9. Reactor trip [8 @ 0.25 each]

REFERENCE Cook RO-CO-PG10, Pg. 11 ad TP-8 , Dwr,yk ArEiele DCC-9-I-IV fye 2-ANSWER 2.03 (1.50)

1. Limit RCS pressure to < Rx trip setpoint for all transients including load rejection if Rod Control and Steam Dumps are in Auto.

2. Prevent challenges to the pressuriser safety valves.

3. Provide low temperature overpressure protection.

4. Allow "once-through" cooling of the core. [any 3, at 0.5 ea]

FER$E /

Cook RO-C-NSO3, Pg. 16 h5 9

&&b ag j

/g/g j m j [ '

OMP-vost.002.o07 oHP- 4051. oo3. cor l

l l

l

2. PLANT DESIGN INCLUDING SAFETY AND EE RGENCY SYSTEMS PAGE 21 ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

MAS"!R CPY ANSWER 2.04 (1.00)

1. Containment Isolation

2. Loss of all charging pumps (breakers)

3. Low pressuriser level (17.0%)

four,

4. Closure of letdown isolation valves (QRV-111, -112) Ay[0.25each]

iEiEICEACE** '"'I'*"'A ' !"-

Cook RO-C-NSO6, Pg. 8 u .ebe,- CP-f - 5/.19-/sL .

I)ro:le ANSWER 2.05 (1.00)

To inject high pressure oil between the thrust bearing faces, forming a film and lifting the faces apart prior to starting a reactor coolant pump

[0.5], thereby reducing the starting torque / current [0.5]. (1.0)

REFERENCE Cook RO-C-NSO2P, Pg. 7 ANSWER 2.06 (1.00)

1. Leakage of contaminated water into the CCW system will be detected by the radiation monitor located at the inlet to the CCW HX [0.5].

(An alarm on this monitor will also shut CRV-412.)

2. Leakage of water into the CCW system can be detected by an INCREASE in surge tank level [0.5]. (1.0)

REFERENCE Cook RO-C-ASO1, Pg. 9

1

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 I ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

GSER C0?Y ANSWER 2.07 (1.00)

1. Safety Injection

2. Signals Blackout

3. Low header pressure (40 psig) [0.33 each]

REFERENCE Cook RO-C-ASO2, Pg. 6 ANSWER 2.08 ( .50)

Approximately 6% (accept any value from 4% to 8%.) (0.5)

REFERENCE Cook RO-C-PG2A, Pg. 6 ANSWER 2.09 (1.50)

1. RER HX

2. CCP (gear oil cooler, bearing oil cooler, seal plates)

3. SI pump (mech. seal HX, bearing oil cooler)

4. RHR pump (mech. seal HX)

5. CTS pump (mech. seal HX) [0.3 each]

REFERENCE Cook RO-C-AS01, Pg. 6 ANSWER 2.10 (1.00) 175,000 nine [0.5 ea]

REFERENCE Cook RO-C-ASil, Pg. 6

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

MAHR CCoY ANSWER 2.11 (3.00)

a. East motor-driven feedpump discharges to S/G's 2 & 3.

West motor-driven feedpump discharges to S/G's 1 & 4.

Turbine-driven feedpump discharges to ALL S/Gs. [0.5 each]

b. Unit 2 motor-driven auxiliary feedpuinps can be used, the east pump discharging to S/G's 1 & 4 and the west pump discharging to S/G's 2

& 3. (The turbine-driven aux, feedpumps cannot be crosstied.) [1.5]

(3.0)

REFERENCE Cook RO-C-ASil ANSWER 2.12 (3.00)

a. The valves are opened and geenergized (prior to exceeding 1000 psig in mode 3), _ _ 0} y jy y enn J d'

b. Volume: 929-971 cu. ft, water, bot units [U.5] ff'p% ,

b8N/er F, sero,E 7

~

Boron cone.. >/=1950 ppm U-1 [0.5], 1900-2100 ppm U-2 [0.5]

Pressure: 585-658 psig U-1 [0.5] 599-644 psig U-2 [0.5] (3.0)

REFERENCE Cook RO-C-NS12, Fg. 6 C HP- yogo, Srp, 03 o DH P- +'0? Y, lot, dp 9 ANSWER 2.13 (1.50)

1. Inject borated water into RCS when pressure < shutoff head of RER pumps. ( f /g / sy 'd)

2. Provide suction supply to SI pumps and CCP's during recirculation phase.

" )

3. Provide supplemental containment spray. .5 each]

REF NE Ccok RO-C-NS08, Pg. 2, g, <

=cc 5 i bf M/ (&k', -

5'ysm.Desa.y';m yp_,)j ,

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24

• -86/05/19-JENSEN, N.

ANSWERS -- COOK 1&2 NASER CPY ANSWER 2.14 (2.00)

a. False

b. True

c. False

d. False [0.5 each]

REFERENCE Cook RO-C-NOS8 ANSWER 2.15 (2.00)

1. Running relay energised, OR: DG speed greater than 95% rated speed.

2. Zero voltage on respective bus.

3. RCP bus tie breaker open.

4. >80% minimal voltage on DG generator.

5. 69KV emergency feeder breaker open.

6. Control switch not in lockout. [0.4 each, five required]

7. 0"4 Fd ' re4Aer c med JA /ackagf, REFEREtEE RO-C-PG14 p32, RO-C-AS10 p29.

l ANSWER 2.16 (2.00)

1. No gravity drains.

2. Suction lines are located near the SFP surface (19 1/2 inches below ,

normal level).  ;

3. The adjustable support for the skimmers limits downward travel to 6 inches below normal level.

4. Pump discharge lines terminate >/= 6 feet ABOVE fuel assemblies and ,

include anti-siphon holes ~4 inches below normal level. l

[0.5 ea.] l l

l l

l

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

- = - cz Cook RO-C-AS05, Pg. 2 N ASTER C'PY V

l

3. INSTRUMENTS AND CONTROLS PAGE 26 ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

QSER 0:PY ANSWER 3.01 (1.50)

Interlocked so that IMO-215 and -225 must be shut before the recire sump suction valves (ICM-305 or -306) will open [0.75], to prevent draining the RWST to the recire sump [0.75]. (1.5)

REFERENCE Cook RO-C-NS15, Pg. 13 ANSWER 3.02 (2.00)

1. Interlocks all heaters off below a predetermined setpoint (17%)

[0.5] to prevent heater burnout (caused by heaters operating in a dry environment). [0.5] (1.0)

2. Turns on backup heaters if actual level varies above program by a predetermined setpoint (5%) [0.5], in anticipation of a low pressure condition caused by an insurge of relatively cold water [0.5]. (1.0)

REFERENCE Cook RO-C-NS06, Pg. 25 ANSWER 3.03 (1.50) 87% - High level alarm, trip to divert 78% - Begins auto divert 24% - Stops auto makeup 17% - Eegins auto makeup 14% "Not in Auto" alarm 7% - "VCT irew Low" alarm 1% - Auto shift to RWST Setpoints: [0.06 each]

Functions: [0.15 each] (1.5)

REFERENCE Cook RO-C-NS06, Pg. lo; o#P- f0.2 y, /0 9 .prir ff,

. <r

3. INSTRUMENTS AND CONTROLS PAGE 27 ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

MB C?Y ANSWER 3.04 (1.00)

Modes'$$)2,and3, (to provide post-accident monitoring per T/S 3.3.3.8)

Mode 5, when </= 170 F, if a PORV is required to be REFERENCE Cook TS 3.3.3.8 & 3.4.9.3 (U-1) 3.3.3.6 & 3.4.9.3 (U-2)

ANSWER 3.05 (2.50)

a. 1. 3/3 condenser vacuum > 10.6" Hg.

2. At least one circ water pump operating. N 3.

OutputvoltagefromCRIDII,{--Unit 1[0.33 only) ea] (1.0)

b. 1. Tavg > 541 F [0.5]

2. Step load decrease of at least 10% [0.5], or turbine trip [0.5].

l (Part b.2 comprises the " arming signal".) (1.5) l l REFERENCE Cook RO-C-PG12, Pg. 6-11 l

ANSWER 3.06 (1.00)

1. Steam flow: used to provide the program dP signal. (Added to the no-load setpoint.) [0.5]

l l b. Steam pressure: used in conjunction w/ feed pressure to generate the actual dP signal. [0.5] (1.0)

REFERENCE Cook RO-C-PG11, Pg. 7 h

3. INSTRUMENTS AND CONTROLS PAGE 28 l ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

MAHR C:PY ANSWER 3.07 (1.50)

a. Turbine impulse pressure (MPC-253). (0.5)

b. Program level ramps from 33% to 44% as impulse pressure varies from 0% to 20% [0.5]. From 20% to 100% impulse pressure, program level remains constant at 44% [0.5]. (1,0)

REFERENCE Cook RO-C-PG11, Pg. 2 ANSWER 3.08 (1.50)

a. Rods move IN, because Tavg is higher than Tref. (0.5)

b. rods move OUT, because Tavg becomes less than Tref. (0.5)

c. Rods move OUT, because the power mismatch circuit sees turbine power (as sensed by Pimp) increasing above Rx power. (0.5)

REFERENCE Cook RO-C-NSO4, Pg. 2 ANSWER 3.09 (1.50)

Because at higher power levels, need a smaller change in reactivity to get the same change in % power than that which is needed at lower power levels.

(1.5)

REFERENCE Cook RO-C-NSO4, Pg. 6

3. INSTRUMENTS AND CONTROLS PAGE 29 ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

VASTER CCPY ANSWER 3.10 (2.00)

1. Power range high flux - 103% [0.2]

2. Intermediate range overpower - current equiv. to 20% [0.2]

3. OP Delta-T - 3% below setpoint [0.2]

4. OT Delta-T - 3% below setpoint [0.2]

5. Urgent Failure Alarm (no setpoint required) [0.1]

1 thru 5 affect both manual AND auto rod withdrawal. [0.5]

6. Turbine power < 15% [0.2] 2.30

7. Control bank D withdrawal stop - M steps [0.2]

6 and 7 affect auto withdrawal only. [0.2] (2.0)

REFERENCE Cook RO-C-NSO4, Pg. 8 l-OHf- yop y, //0, 0yo, [s-f d'OHf- yppy jjo,99en,[$.g ANSWER 3.11 (2.00)

1. High pressure Rx trip, 2/4, >2378 psig

2. Low pressure Rx trip, 2/4, <1872 U-1, <1966 U-2 ,

3. Low pressure SI, 2/3, <1837 U-1, <1908 U-2

4. Low pressure SI Block, 2/3, <1915 U-1, <1990 U-2

)

g6 (

-l'

5. Provides input to OT Delta-T circuit C.

[Each correct component worth -G--F2tFJ (2.0) l REFERENCE Cook RO-C-NS03, Pg. 27  !

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1 1

3. INSTRUMENTS AND CONTTQLS PAGE 30 ;

ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

NASER CPY .

ANSWER 3.12 (2.00)

AUCTIONEEEED HIGH Tavg:

1. Rod control system temperature deviation circuitry.

2. Pressurizer level control system (to generate program level).

3. Steam dump control system.

4 R:d i.n erti o n 1.a . w +=ra wn .

INDIVIDUAL LOOP Tavs OT

1. M Delta-T calculator

2. OF Delta-T calculator

3. P-12 circuitry (541 F - Hi, Stm. Flow SI permissive, Stm. damp block.)

4. Feedwater isolation circuitry (554 F).

o.35 0.036 M points each function; -0,-GEr points each correct source - auctioneered or individual.) (2.0)

REFERENCE Cook RO-C-h,--Es . 17-MOV j fev, 0, fN -.

ANSWER 3.13 (3.00)

1. Low res

< Suspressurdi r>,w hh er ?k SI, o!ure - LOCA, steambreak,fandDNBprotection)P-il

2. Lower containment high pressure - LOCA, stetim or feed break inside containment.

3. High steamline Delta-P - steamline break upstream of stop volves.

4. a. Unit 1: High steamline flow coincident with low steamline pressure OR low-low Tavg. P-12 allows block of SI (but not the associated steamline isolation). - Major steamline break '

downstream of stop valves.

b. Unit 2: Low steamline pressure 4R L.sh :te '*r.: fi-a cin :i d c.. ith L., a n ... P-12 allows block (of all but high steam flow isolation). - Steam break downstream of stop valves.

[0.2 points for each component of answer.) (3.0) ,

i REFERENCE Cook RO-C-NS12, Pg. 15 i

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3. INSTRUMENTS AND CONTEOLS PAGE 31 ANSWERS -- COCK 1&2 -86/05/19-JENSEN, N.

~

KNER CTY ANSWER 3.14 (2.00)

MDAFP auto-start:

1. Safety Injection Signal

2. Blackout- ' .., .

3. S/G low-low level 2/3 channels on 1/4 S/G's. U1 - 17%, U2 - 21%

4. Both main feed pumps tripped (only if pump control switch in auto).

TDAFP auto-start: ,

1. Low volt' age on 2/4 RCP busses 2 S/G low-low level 2/3 channels on 2/4 S/G's. (Same setpoints as 3 above).

[0.2 point each condition, 0.2 point each conditions' logic, 0.2 point for correct setpoints.) (2.0)

REFERENCE Cook RO-C-AS11, Pg. 13-14

4. PROCEDURES - NORMAL. ABNORMAL.-EMERGENCY SMD PAGE 32 BbDIOLOGICAL CONTROL ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

N AS"B C:PY ANSWER 4.01 (2.00)

The diesel has the best chance of curviving overload conditions when paralleled to vital pump busses [1.0] because the diesel breaker and RCP bus tie breaker see the same overload conditions, and the overcurrent trip on the RCP tie is set lower than the diesel. [1.0] (2.0)

REFERENCE OHP 4021.032.001., RO-C-ASIO, pg 36 ANSWER 4.02 (2.00)

MAXUMUM:

</ 5000 gpm [0.5], design limit [0.5]

MINIMUM:

to prevent the flashing of CCW to steam (as a result of Sufficientflow)RHRflow).

cat input from [1.0] (2.0)

REFERENCE OHP4021.017.002, 4.2 and " Note" following 4.2 ANSWER 4.03 (2.00)

a. 170 F, 152 F [Two parts, @0.5 ea]

b. Provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. [1.0] (2.0)

REFERENCE 4021.002.003-4.3., TS pg B3/4 1-3 ANSWER 4.04 (1.50)

1. Rx Coolant Temperature <150 F, [0.5] or

2. RCP seal leakage rate is </= 5 gym, [0.5] and about 35 gpm of CCW w/

inlet temp. of about 80 F is flowing through thermal barrier cooling coils. [0.5] (1.5)

4. PRO.QEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

REFERENCE CHP 4021.002.003, Note following 4.4 ANSWER 4.05 (1.00)

To prevent possiole damage to the RCP #2 seals, OR: To ensure enough backpressure to maintain the #2 seals wet.; [0.5]

15 psig [0.5] (1.0)

REFERENCE OHP 4021.003.005, 4.1; RO-C-NSO2P, p5.

ANSWER 4.06 (2.00)

1. Place main turbine solenoid trip switch to " Solenoid Trip".

2. Place both EHC fluid pump control switches to "Stop".

3. If turbine will not trip, then manually run back turbine.

4. If turbine cannot be run back, then trip S/G stop valves closed.

[0.5 point each]

REFERENCE FR-S.1, step 2 ANSWER 4.07 ( .50)

If containment hydrogen concentration is ' O . 0"o' f g } (0.5)

~ y<

REFERENCE C > ya n nd <yri. s 2 * -

E-1, step 11.

PAGE 34

4. PRQEEDURES - NORMAL. ABNORMAL. EMERGENCY AND

, BADIOLOGICAL CONTROL ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

kAS"ER 0:PY ANSWER 4.08 (2.00)

1. If red path encountered, must immediately stop optimal recovery and I initiate function restoration to restore the CSF under challenge.

2. If orange path encountered, should continue to check status of all  !

CSF's. Then stop optimal recovery and initiate function restoration to highest priority orange path CSF under challenge.

3. If yellow path encountered, should continue monitoring all status trees. May restore CSF any time at operators discretion.

>R((Whileaddressingredororangepath, if higher priority challenge is i diagnosed, operator must terminate on-going response and initiate

(, function restoration for highest priority challenge.

E* 4< Green path - no further action required. [D 4'ea] (2.0)

c. s' REFERENCE Cook Emerg. Procs., WOG ERG Bkgnd. Doc ANSWER 4.09 (2.00)

1. Suberiticality (S)

2. Core Cooling (C)

3. Heat Sink (H)

4. Integrity (P)

5. Containment (Z)

6. Inventory (I)

(0.25 point each item, 0.5 point correct priority.)

REFERENCE Cook Emerg. Procs, WOG ERG Ekgnd. Doc l

ANSWER 4.10 (2.00)

1. Visual check of instrument

2. Ensure instrument in current calibration

3. Check battery condition

4. Check instrument response to known radiation source. [0.5 ea]

4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 35 RADIOLOGICAL CONTROL ANSWERS -- CCOK 1&2 -86/05/19-JENSEN, N.

REFERENCE k ASF"R CC)Y RO-C-AS22, pg 2 ANSWER 4.11 (3.00)

A. True D. True B. True E. False C. False F. True [0.5 ea]

REFERENCE PMP 6010. RAD.001 pg 80, 90, 122, 124 ANSWER 4.12 (1.50)

a. 2 minutes

b. 5 minutes

c. 15 minutes [0.5 ea]

REFERENCE OBP 4022.005.002, 2.3 ANSWER 4.13 (2.00)

1. Close all three motor-operated isolation valves (NMO-151, -152, and

-153.) [1.0]

2. After relief line temperature has cooled, reopen the isolation valves one at a time. Wait between opening each valve for leakage indication. [1.0] (2.0)

REFERENCE OHP 4022.002.009, 4.2 l

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. 4. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE 36 l RADIOLOGICAL CONTROL l i

ANSWERS -- COOK 1&2 -86/05/19-JENSEN, N.

J mar.P0:PY  :

ANSWER 4.14 (1.50)

a. 1. Malfunction of steam seals.

2. Malfunction of air ejectors.

3. Air in-leakage due to fault or improper valve lineup.

4. Inadequate cire. water flow. [0. '25 ea]

b. 21.9" HG and still decreasingj /4gf [0.5] (1.5) jj # A' ~ d >pt4&l, e OR : ha, .4 fn OHe <022.053.001

,n 4 (e a y)

TEST CROSS REFERENCE PAGE 1

• QUESTIbN VALUE REFERENCE a==

01.01 3.00 NCJ0000149 01.02 1.50 NCJ0000150 01.03 1.50 NCJ0000151 01.04 2.00 NCJ0000152 '

01.05 3.00 NCJ0000153 01.06 1.50 NCJ0000154 01.07 3.00 NCJ0000155 01.08 3.00 NCJ0000156 01.09 1.50 NCJ0000157 01.10 1.00 NCJ0000158 01.11 2.25 NCJ0000159 01.12 1.75 NCJ0000160 25.00 02.01 1.00 NCJ0000161 02.02 2.00 NCJ0000162 02.03 1.50 NCJ0000163 02.04 1.00 NCJ0000164 02.05 1.00 NCJ0000165 02.06 1.00 NCJ0000166 02.07 1.00 NCJ0000167 02.08 .50 NCJ0000168 02.09 1.50 NCJ0000169 02.10 1.00 NCJ0000170 02.11 3.00 NCJ0000171 02.12 3.00 NCJ0000172 02.13 1.50 NCJ0000173 02.14 2.00 NCJ0000174 02.15 2.00 NCJ0000175 02.16 2.00 NCJ0000176 25.00 03.01 1.50 NCJ0000177 03.02 2.00 NCJ0000178 03.03 1.50 NCJ0000179 03.04 1.00 NCJ0000180 03.05 2.50 NCJ0000181 03.06 1.00 NCJ0000182 03.07 1.50 NCJ0000183 03.08 1.50 NCJ0000184 03.09 1.50 NCJ0000185 03.10 2.00 NCJ0000186 03.11 2.00 NCJ0000187 03.12 2.00 NCJ0000188 03.13 3.00 NCJ0000189 03.14 2.00 NCJ0000190 25.00 04.01 2.00 NCJ0000191

TEST CROSS REFERENCE PAGE 2

==

' QUESTION VALUE REFERENCE

__________ l)y 04.02 2.00 NCJ0000192 l l 04.03 2.00 NCJ0000193 04.04 1.50 NCJ0000194 04.05 1.00 NCJ0000195 04.06 2.00 NCJ0000196 04.07 .50 NCJ0000197 04.08 2.00 NCJ0000198 04.09 2.00 NCJ0000199 04.10 2.00 NCJ0000200 04.11 3.00 NCJ0000201 04.12 1.50 NCJ0000202 04.13 2.00 NCJ0000203 04.14 1.50 NCJ0000204 25.00 100.00 l

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1 U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: COOK 1&2 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 86/05/19 EXAMINER: ISAKSEN. P.

APPLICANT:

INSTRUCTIONS TO APPLICANT:

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

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

% OF CATEGORY  % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 25.00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 25.00 25.00 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.00 100.00 TOTALS FINAL GRADE  %

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

n i

V AFPLICANT'S SIGNATURE

1 t

ES-201-2 i

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 candidata at a time may l 1 eave. 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. ,

l

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. i

8. Consecutively number each answer sheet, write "End of Category

• as appropriate, start each category on a new page, write only one sTde of the paper, and write "Last Page" on th7 east 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. l

11. Separate answer sheets free 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.

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

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

l

15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER'8 LANK.

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

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 examinatio'n. This must be done after the examination has been completed.

Examiner Standards 12 of 18

O ES-201-2

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.

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

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.

i l -

l Examiner Standards 13 of 18

5. THEOPY OF Nr.tOLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 2 THERMODYNAMICS s

QUESTION 5.01 (3.00)

The operator has brought the reactor critical at BOL and is increas-ing power to 10-8 amps.

a. If the critical data was recorded with Bank D at 120 steps, Tave at 547 F, and a critical boron concentration of 1000 ppm what would be the expected SUR if Bank D was raised to 152 steps? Assume Bank D differencial rod worth is 6 PCM/ step and lambda is 0.1/sec.

Show all work. (2.0)

b. If rod motion is stopped after establishing a positive DPM startup rate at 10-8 amps, what would happen to power level over the next five minutes? Assume at least a 1.0 DPM SUR and assume no automatic action occurs. (1.0)

QUESTION E.02 (1.50)

Uhat is "Self Shielding" and how does it affect reactor operations?

QUESTION 5.03 (1.00)

When performing a reactor S/U to full power that commenced five hours after a trip from full power equilibrium conditions, a 0.5%/ min ramp was used. How would the resulting xenon transient vary if instead a 2%/ min ramp was used?

a. The xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be smaller.

b. The xenon dip for the 2%/ min ramp would occur later and the magnitude of the dip would be smaller.

c. The xenon dip for the 2%/ min ramp would occur sooner and the magnitude of the dip would be larger.

d. The xenon dip for the 2%/ min ramp would occur later and the magnitude of the dip would be larger.

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

5. THEORY OF NUCLEAR POWER PLANT CPERATION. FLUIDS. AND PAGE 3 IEERMODYUAMICE QUESTION 5.04 (1.00)

During a Xenon-free reactor startup, critical data was inadvertently taken two decades below the required Intermediate Range (IR) level (1xE-10 amps).

The critical data was taken again at the proper IR level (1xE-8 amps).

Assuming RCS temperatures and boron concentrations were the same for each set of data, which of the following statements is correct?

a. The critical rod position taken at the proper IR level is LESS THAN the critical rod position taken two decades below the proper IR level,

b. The critical rod position taken at the proper IR level is THE SAME AS the critical rod position taken two decades below the proper IR level.

c. The critical rod position taken at the proper IR level is GREATER THAN the critical rod position taken two decades below the proper IR level.

d. The critical rod position taken at the proper IR level CANNOT BE CCMPARED to the critical rod position taken two decades below the proper IR level.

QUESTION 5.05 (1.00)

In which of the following conditions is the Moderator Temperature Coefficient most negative?

a. BOL, high temperature

b. BOL, low temperature

c. EOL, low temperature

d. EOL, high temperature l

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

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

'l QUESTION 5.06 (2.50)

a. With the plant operating at 85% power and all systems in a normal / automatic configuration, the operator borates 100 PCM.

Shutdown Margin ... (1.0)

1. Increases.

2. Increases until rods move.

3. Decreases.

4. Decreases until rods move.

5. Remains unchanged, whether or not rods move.

b. State the THREE purposes for the Control Group Insertion limits according to Technical Specifications. (1.5)

QUESTION 5.07 (1.00)

List the components of the power defect in an INCREASING order of significance (reactivity value) for BOL.

QUESTION 5.08 (1.00)

Which one of the following statements concerning the power defect is correct?

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

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

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

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

QUESTION 5.09 (1.00)

During fuel loading, how does the strength of the neutron source effect the shape of the 1/M curve?

1

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

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5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIES. AND PAGE 5 i IHERMODYNAMICS I

f QUESTION 5.10 (1.00)

During a reactor startup, the first reactivity addition caused count rate to increase from 10 cps to 16 cps. The second reactivity addition caused count rate to increase from 16 cps to 32 cps. Which of the following atatements describing the relationship between the reactivity values of the first and second reactivity additions is correct?

I

a. The first reactivty addition was larger.

b. The second reactivity addition was larger.

c. The first and second reactivity additions were equal.

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

QUESTION 5.11 (1.00)

TEUE or FALSE?

o. As Keff approaches unity, a smaller change in neutron level will result for identical changes in Keff.

b. As Keff approaches unity, a longer period of time is required to reach the equilibrium neutron level for identical changes in Keff.

QUESTION 5.12 (1.00)

Initially, one centrifugal charging pump is in operation when a second centrifugal charging pump in parallel with the first pump is also put into operation. Describe the effect on system volumetric flow rate and system head loss?

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

1

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 6 THERMODYNAMICS QUESTION 5.13 (1.50)

Unit 1 is at 50% power with control rods in MANUAL when the turbine is ramped up to 60%. Indicate whether the parameters below will increase, decrease or remain the same during both the initial response (first 30 seconds of the transient) and after turbine power has stabililzed relative to the initial conditions. (Assume the following: No changes to boron / xenon Loop transport time is 10 seconds No operator actions)

NOTE: No answer required where it is already filled in below.

Initial Response Steady State a) S/G Pressure NO ANSWER RQRD b) Reactor Power NO ANSWER RQRD c) Tcold d) Tavg QUESTION 5.14 ( .50)

TRUE or FALSE?

High energy neutron exposure increases the possibility of brittle fracture of the reactor vessel by increasing the compressive stress on the reactor vessel.

QUESTION 5.15 (1.00)

If RCP's are tripped following a LOCA, and the break has been isolated, which of the following situations would be MOST desirable?

PZR HOT LEG COLD LEG Press. Temp. Temp.

a. 600 500 480

b. 800 530 520

c. 1000 540 530

d. 1200 575 565

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

5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 7 THERMODYNAMICE QUESTION 5.16 (1.50)

Indicate whether the following are TRUE or FALSE.

I a. It will take a smaller letdown flow rate to maintain a constant actual pressuriser level, during a RCS heatup, with temperature increasing at a constant rate.

b. Increasing condensate depression (subcooling) will cause BOTH a decrease in plant efficiency AND an increase in condensate (hotwell) j pump available NPSH.

c. The difference between pump suction pressure and the saturation pressure of the fluid being pumped is referred to as net positive suction head.

QUESTION 5.17 (1.00)

In order to maintain a 200 F subcooling margin in the RCS when reducing RCS pressure to 1600 psig, steam generator pressure must be reduced to approximately what value?

QUESTION 5.18 (1.00) d Which one of the following requires the most heat energy to be removed by the main condenser?

a. One pound of steam at 0 psia.

b. One pound of steam at 300 psia.

c. Two pounds of steam at 600 psia.

d. Two pounds of steam at 1200 psia.

QUESTION 5.19 (2.00)

After a secondary calorimetric and adjustment of the power range inst-ruments, it is discovered that the Auxiliary Feedwater Pumps were operating. HOW and WHY would this occurence affect the power range adjustment?

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

5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 8 TEERMODINAMICS QUESTION 5.20 ( .50)

TRUE or FALSE?

During 100% power operation, Departure from Nucleate Boiling Ratio (DNBR) is greater than the DNER for 20% reactor power.

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

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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 9 QUESTION 6.01 (2.00)

List EIGHT conditions which will cause main feedwater pumps to trip.

(Setpoints not required.)

'l QUESTION 6.02 (1.00)

What are FOUR conditions which cause automatic closure of the CVCS orifice isolation valves QRV-160, -161, and -162?

3 QUESTION 6.03 (1.50)

-i i List the loads on each SAFEGUARDS train which are cooled by CCW.

QUESTION 6.04 (1.50)

What are THREE uses of the RHR system following a Safety Injection?

QUESTION 6.05 (2.50)

State FIVE signals required for the diesel generator output breaker to automatically close during an emergency?

4 QUESTION 6.06 (2.00)

What design features prevent inadvertent draining of the Spent Fuel Pit?

(FOUR required.)

4 QUESTION 6.07 ( .50)

Describe the interlock associated with the containment spray pump cuction valves IMO-215 and IMO-225 and BRIEFLY EXPLAIN why it is installed.

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(***** CATEGCRY 06 CONTINUED ON NEXT PAGE *****)

2

- - - . - - . - - - - - - . .c - ,e ,

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTSUMENTATION PAGE 10 QUESTION 6.08 (2.50)

a. What conditions must be met to activate the Condenser Vacuum Permissive (C-9). Include setpoints, logic and unit differences.

b. What OTHER conditions must be met to unblock the steam dump arming valves if the plant is initially operating at 100%

power? Be specific.

QUESTION 6.09 (1.00)

STATE the input signals used in the main feed pump speed control circuit, and BRIEFLY DESCRIBE the purpose served by each.

QUESTION 6.10 (1.50)

Unit 1 is operating at 45% power with all systems in automatic control.

For each condition listed below, give the initial direction of rod motion, cnd state the initial reason for this rod motion.

I

a. Loop 3 Tc fails high.

b. A main steam power operated relief valve fails open.

c. The turbine is ramped to 100% power at 5% per minute.

QUESTION 6.11 (2.00)

List all conditions which will prevent rod WITHDRAWAL. Indicate whether each is effective in the MANUAL mode, AUTOMATIC mode, or BOTH. Include setpoints as applicable.

QUESTION 6.12 (2.00)

List ALL inputs to RPS/ESF provided by pressurizer pressure instrumentation. Include applicable setpoints and coincidence, and specify unit differences.

(***** CATEGORY 06 CONTINUED ON NEXT'PAGE *****)

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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION FAGE 11 QUESTION 6.13 (3.00)

List all AUTOMATIC SI actuation parameters. Include accidents protected against by each parameter listed, and any permissives associated with each. Include unit differences. .(Setpoints not required.)

QUESTION 6.14 (1.00)

Describe the safety-related function of the Condensate Storage Tank.

QUESTION 6.15 (1.00)

Per the Technical Specifications, when must reactor coolant wide range pressure instrumentation be operable 7 1

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(***** END OF CATEGORY 06 *****)

? PEOCEDURES - NORMAL. ABNORMAL. EMERGENCY AEQ PAGE 12 RADIOLOGICAL CONTROL 1 .

0"ESTION 7.01 (1.00)

Which of the following statements, concerning TS actions required for Nuclear Instrument malfunctions, is correct?

a. If a source range channel fails while a startup is in progress and reactor power is below P-6, insert all control banks to to zero steps.

b. If an intermediate range channel fails while a startup is in progress and reactor power is above P-6 but below P-10, the power increase may continue using the operable intermediate range channel,

c. Failure of one power range channel during shutdown precludes reactor startup until the failed channel is returned to operable status.

d. Failure of both source range channels while shutdown requires shutdown margin requirements to be verified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

QUESTION 7.02 (1.50)

A 25 year old radiation worker's total lifetime dose is 10 Rem and his NRC Form-4 is up to date. His personnel monitoring devices are read for the first week of the quarter with the following results: 500 mrem-whole body; 500 mrem-skin; 3000 mrem-extremities. What are his dose limits for the remainder of the quarter for: (Choose from 1-4 for answers)

a. Whole body 1. 25 Rem

b. Skin 2. 15.75 Rem

c. Extremities 3. 7 Rem i

4. 2.5 Rem

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

7. FROCEDUEES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 13 RADIOLOGICAL CONTROL QUESTION 7.03 (1.00)

Bhich of the following describes the RCP Trip Criteria (E-0, Foldout),

following a valid safety injection initiation?

i

a. Less than 25 degrees subcooling and pressuriser level < 10%.

b. Less than 30 degrees subcooling and one CSIP is running.

c. RCS pressure less than 1250 psig and one CSIP is running.

d. RCS pressure less than 1600 psig and pressurizer level < 10%.

QUESTION 7.04 (1.00)

According to Foldout for E-0, which of the following conditions meet the.SI I

Actuation Criteria? Assume normal containment parameters.

! a. RCS pressure less than 2000 psig, RCS subcooling less than 50 degrees, or pressurizer level less than 20%.

b. RCS pressure less than 1250 psig, RCS subcooling less than than 50 degrees, or pressurizer level less than 50%.

c. RC3 pressure less than 1600 psig, RCS subcooling less than 30 degrees, or pressuriser level less than 4%.

, d. RCS pressure less than 1600 psig, RCS subcooling less than 30 degrees, or pressurizer level less than 50%.

QUESTION 7.05 (1.00)

What containment indications define ADVER5E CONTAINMENT conditions?

t

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

i

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND FAGE 14 RADIOLOGICAL CONTROL er i QUESTION 7.06 (2.50)

The following concern Dropped Rod procedure 1-OHP 4022.012.004 and Malfunc-i tioning Rod Position Indication procedure 1-OHP 4022.012.007. L i.

a. List SIX symptoms which may indicate a rod dropped during normal power l' operation. '1.5)

b. How is an actual dropped rod condition differentiated from a malfunct- ,,

3 icning rod position indication? (0.5) 3

c. An immediate action of the malfunctioning rod position indication j procedure is to determine actual rod position, HOW is this determina-j tion performed? (0.5)

I i

QUESTION 7.07 (2.00)

Indicate whether or not each of the following is a duty / responsibility of the SRO-CA, according to PMI-4050, Fuel Handling.

j a. Monitor communications with all fuel handling stations including the

Control rocm.

i

b. Approve the bypassing of any interlock on fuel handling equipment.

l

c. Directly responsible to the operations superintendent for compliance with technical specifications requirements and performance of core alterations in accordance with approved procedures, e

i d. Grants permission to the manipulator crane operator before any fuel j assembly is unlatched within the reactor vessel.

l 4

QUESTION 7.08 (2.00)

-List FOUR of the conditions which indiate natural circulation flow accord-j ing to ES-0.1 Attachment B.

i j

l 4

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

i

i 7 PROCEDURES - NORMAL. ABNCRMAL. EMERGENCY AND PAGE 36 BADIOLOGICAL CONTROL ANSWERS -- COCK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 7.08 (2.00)

(four required. 0.5 each)

-RCS subcooling (based on core exit TCs) >30-F

-S/G pressures - stable or decreasing 1

-RCS hot leg temperatures - stable or decreasing

-core exit TCs - stable or decreasing

-RCS cold leg temperatures - at saturation temperature for S/G pressure REFERENCE DC Cook ES-0.1 Attachment B.

ANSWER 7.09 (3.50)

o. Rod bottom lights - lit Reactor trip (and bypass) breakers open Rod position indicators - less than 25 steps (indicate rods inserted)

Neutron flux - decreasing [0.25 each)

b. -manually insert control rods

-verify turbine trip check AFW pumps running

-initiate BIT injection

-check PZR pressure (<2335 psig) [0.5 each]

REFERENCE DC Cook FR-S.1, p 2-5.

ANSWER 7.10 (2.00)

1. Suberiticality (S)

2. Core cooling (C)

3. Heat sink (H)

4. Integrity (P)

5. Containment (Z)

6. Inventory (I) [0.25 each, 0.5 correct priority]

7. PROCEDURES - NORMAL. AENORMAL. EMERGENCY A'D d PAGE 15 RADIOLOGICAL CONTROL i -

QUESTION 7.09 (3.50) i

a. How is a reactor trip verified according to E-0, Reactor Trip or Safety Injection procedure? (four responses required) (1 0)

I b. List the FIVE immediate operator actions / steps required by FR-S.1, Response to Nuclear Power Generation /ATWS procedure, after a manual trip of the reactor, using the reactor trip switches, is unsuccessful .

(2.5)

QUESTION 7.10 (2.00)

List the critical safety function status trees in order of priority. (Use number 1 as the highest priority)

QUESTION 7.11 (3.00)

Indicate whether each of the following statements concerning use of radiat-ion work permits is TRUE or FALSE, assume reactor power is 50%.

j a., In certain instances, radiation personnel may be substituted for a RWP during work performance.

b. A RWP is required for any task involving entry into containment.

c. Except in emergencies, RWP's must be initiated by radiation control i personnel, f

d. An extended RWP may be authori=ed for work which is repetitive in nature.

e. An extended RWP may be authori=eu for work in the regenerative heat j - exchangar area.

f. Personnel may deviate from the conditions of a RWP if authori=ed.tur the Shift Supervisor.

1

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

l t

.----ev ,

, - - - - ,- ~ m a, -

-. .-.,y pm-,--,~ ,,,,,-,-p- r-, r e-

7. PROCEDURES - NORMAL. AENORMAL. EMERGENCY AND PAGE 16 RADIOLOGICAL CONTROL 4

I GUESTION 7.12 (2,50)

The following concern reactor coolant pump (RCP) operation procedure 1-OHP 4021.002.003.

a. The folicwing parameters relate to proper operation and shall be maint-ained during pump operation and exist prior to starting a RCP:

Minimum No. 1 seal delta-p __1__ psid

. Minimum No. 1 seal leakoff __2__ gpm Minimum VCT pressure __3__ psig Minimum loop pressure __4__ psig

b. What is the reason for the minimum VCT pressure limit?

c. What are TWO abnormal conditions that would require No. 1 seal bypass (1-QRV-150) to be opened without meeting the normal opening-criteria?

i QUESTION 7.13 (1.00)

What operator action is required if during a normal reactor startup, criticality occurs below the insertion limit for zero power?

QUESTION 7.14 (1.00) i Prior to any cold leg temperature going to or below 170-F one of the What is the bases /

operable CCP breakers must be verified racked out.

reason for this requirement?

4 l

i i (***** END OF CATEGORY 07 *****)

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 17 o

QUESTION 8.01 (2.00)

List FOUR of the initial actions required for a chemical spill according to PMI-2230, Reporting of Spills - Nonradioactive procedure.

QUESTION 8.02 (3.00) i The concentration of the boric acid solution in the Refueling Water Storage Tank (RWST) shall be verified once per 7 days in accordance with Technical Specification 3.5.5. The chemist sampled the RWST on the following schedule. (All samples taken at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />).

Mar 1 --- Mar 8 ---Mar 16 --- Mar 24 --- Mar 31

a. EXPLAIN why surveillance time interval requirements were or were not exceeded on Mar 16.

b. EXPLAIN why surveillance time interval requirements were or were not exceeded on Mar 24.

QUESTION 8.03 (1.50)

What are the THREE provisions that must be met before a temporary change can be made to an approved operating procedure, according to Technical Specifications?

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

i I

8. ADMINISTRATIVE PROCEDUBES. CONDIT?ONS. AND LIMITATICEE PAGE 18 3 .

e h

QUESTION 8.04 (1.50)

a. The indicated AFD shall be considered outside of its target band -

when or more operable excore channels are indicating thc  ;

AFD outside the target band. (0.5)

b. Which of the following statements concerning the AFD . requirements is correct? (1.0)

1. Above 90%, within 30 minutes of going outside the target ,

band: either restore indicated AFD to within the target 4 band or reduce power to less than 90%.

I

2. If the axial flux difference alarms are out of service, the axial flux difference shall be logged eveiy hour for I the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and half-hourly thereafter until the alarms are returned to operable.

3. Below 15% power, penalty points are accumulated at one .

half point for every minute outside the target band.

~

4. Power level shall not be increased above 15% unless the AFD is within the target band. ,

).

QUESTION 8.05 (1.00)

Which of the following is the Technical Specification minimum licensed l shift crew composition requirement with both Units operating in Mode 17 i

i a. 1 SS, 1 SRO, 3 RO

b. 1 SS, 2 SRO, 4 RO

c. 1 SS, 1 SRO, 4 RO

d. 1 SS, 2 SRO, 2 RO -

4 i

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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l

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8 ALMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 19 l i

GUESTION 6.06 (2.00)

Indicate whether the following are TRUE or FALSE concerning Clearance Permit System, PMI-2110.

a. Only one striped tag clearance may be issued on a given isolation point and both a red tag and a striped tag clearance permit may be issued for j the same component concurrently,

b. A tag that has fallen off a component can be rehung if an independent verification is performed.

c. Grcund straps are required prior to working on electrical circuits rated in excess of 600 volts and after installation a red tag must be placed on the ground strap.

d. The clearance permit forms for shared systems should be blue. -

QUESTION 8.07 (2.50) <

The following concern containment access according to PMI-4010, Plant Operations Policy. '

h. Who by job position / title must authorize access to the lower volume of  ;

the containment (excluding the piping annulus area? (0.5)

b. State the exception to the two-man safety rule for personnel entering i the containment. (0.5)

! c. While maintaining containment integrity requirements during Mode 4 operation access to the containment requires notification to the 1__ ,

The 2 shall be responsible for maintaining a log of all entries and exits. During periods of frequent entry this log msv be maintained 3 . (Note blanks may have more than one word answers) (1.5) 1 QUESTION 8.08 (2.00)

What are the operator actions required by Technical Specifications, if the ,

cafety limit for RCS pressure has been exceeded during the followings ( IncludLL i applicable time limitations.)

a. Mode 1 operation.

b. Mode 3 operation.

1 1

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(***** CATEGCRY 08 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 20 QUESTION 8.09 '(2.50)

a. List THEEE of the Technical Specification bases / reasons for the minimum temperature for criticality. (1.5)

b. With the reacter critical and Tavg < 541-F, what TWO options / actions are required by TS? (Include applicable time limits in your answer)

(1.0)

QUESTION 8.10 (2.00)

The following concern PMI-2010, Plant Manager and Department Head Instruct-ions, Procedures and Associated Indexes:

a. What does a double asterisk (**) immediately preceeding a procedure number signify? (0.5)

b. A procedure is not used in its entirety, state TWO ways the incompleted steps should be addressed / treated. Include the conditions allowing /

requiring each method. (1.5)

QUESTION 8.11 (1.00)

Uhat is the bases / reason for the Technical Specification requirement to be in Hot Standby with Tava < 500-F when the specific activity of the RCS is above the TS limit?

QUESTION 8.12 (3.00)

For each of the following, indicate the type of ROS leakage classification as defined in Technical Specifications.

a. Leakage through a valve packing that is routed to a coolant drain tank.

b. Steam generator tube leakage.

c. Slight seepage through an elbow socket weld on a RTD bypass line,

d. RCP seal leakage.

e. Valve packing leaks of unknown origin,

f. Reactor vessel flange leakage.

(***** CATEGCRY 08 CONTINUED ON NEXT PAGE *****)

8. ADMINISTRATIVE PROCEDURES. CONDfTIONS. AND LIMITATIONS PAGE 21 l

QUESTION S.13 (1.00)

The following concern PMP 2080 EPP.015. Responsibilities of the On-Site Emergency Coordinator (OSEC).

a. State the title of the person responsible for the INITIAL plant emergency response actions.

b. Who is responsible if the person in a, above is unavailable or becomes incapacitated?

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

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

EQUATICN SHEET v = s/ C/cle efficiency = ('let'acrt f = .ne cut)/(Energy in) .

7

.=q s = V3 : - 1/2 a c 2

!=x A = Ut A = A3 e'

.<E = 1/2 mv a = (Vf - 13 )/t PE = mgn

• = e /t x = &n2/t;jg = 0.693/t1/2 vf = V,

• at 1/2aff = ((t n)(*%)]

t 3 w = , ,p go 2 A= 4 ((cl /2)

• I b}3 d = 931 ma t,te

-n m = V ,,Ao Q = 1\Cas I = I n e'"*

5 = UAa r Pw = W , .h I = I,10** E Ttt. = 1.3/u sur(t) gyg, , ,o,gg3f, P = P010 ,.  ;

P = P e'j '

n SG = 5/(1 - K,g)

SUR = 25.06/T G, = S/(1 - K,gx)

SUR = ZSs/t* + (s - o)T G j(1 - K,ff 3) = G 2(I *

• eff 2)

T = ( t*/s ) + ((a - o '/ Is ] M = 1/(1 - K,g) = G;/G, T = u(s - a) x = (1 - K,g,)/(1 - K,g3)

T = (a - s)/(Ta) SCM = ( - K,g)/K,g t' = 10 secones a = (K,ff-1)/K,ff = /.X,g/K,ff *I I = 0.1 seconds

= ((L'/(T 4,g)] + (I,g/(1 + IT)]

I411*Id Ij d; 2 =2I# 2 22 P = (:av)/(3 x 1010) 2

• N R/hr = (0.3 CE)/c (meters)

R/hr = 6 CE/d2 (f,,g) .

.Hses11aneous 0:nve-siens Watse Darweteas I gal. = 3.345 lem. 1 curie = 3.7 x 1010:ss 1gja.=3.7811:ars 1kg=2.21lem}Stu/hr 1 no = 2.34 x 10 l 1 f. = 7.48 gal. I = = ?.41 x 100 5tu/hr I Censity = 62.41 erg /f 3 Gensity = 1 gm/c:9 lin = 2.54 c:n

• F = 9/5'O + 32 Heat of va ortestion = 970 Stu/lem 'C = 5/9 (%32)

. test of fusien = 144 Stu/lem 1 STU = 778 ft-lbf 1 Atm = 14.7 ssi = 29.9 in. Hg. .

1 ft. H 2O = 0.43351:f/in.

5. THEOPY OF NOCLEAR FOWER PLANT OPERATION. FLUIDS. AND PAGE 22 IEERMODYNAMICS ANSWERS -- CCCK 1&2 -86/05/19-ISAKSEN, P.

7 F .*3 Pv 'i3 N 1:h WP'Y-ANSWER 5.01 (3.00)

a. rho = (152 steps - 120 steps) X 6 pcm/ step [0.6]

= 192 pcm [0.2]

SUR = 26 X [ rho X lambda / (Beff - rho)] [0.6]

192 pcm X 0.1/sec / LS&d'pem - 192 pcm)] [0.4]

= 26 J [M V DP Td* [0.2]

v.9 6

b. Power would level off at a point where minus reactivity due to power defect or MTC and FTC would compensate for the positive reactivity added by the rod withdrawal.[1.0]

PErtnENCE BVPS Reactor Theory Manual Chapter 5, 8 DCC Westinghouse Reactor Physics, pp. I-3.18 & I-5 ANSWER 5.02 (1.50)

The large pellet diameter (relative to the resonance peak energy path) results in a self shielding effect for fuel in the pellet interior.

As fuel pellet temperature rises the off-resonance neutrons (for the most part that would previously pass entirely through the pellet) are more readily absorbed. [1.0]

Causes negative reactivity to be added as power increases. [0.5]

REFERENCE BVPS Reactor Theory Manual Chapter 6 p 33 DCC Westinghouse Reactor Physics, pp. I-5.20 ANSWER 5.03 (1.00) a REFERENCE CNTO " Reactor Core Control" Section 4 Westinghouse Simulator Trng book, "Rx Theory and Core Physics", Fig I-5-54 001/000; K5.38(3.5/4.1)

CNTO, " Reactor Core Control", pp 4-21/28 001/000; K5.38(3.5/4.1)

MIRCCJY

5. THEQRY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 23 THEhMODYUAMICS ANSWERS -- COOK 1&2 -86/05/10-ISAKSEN, P.

ANSWER 5.04 (1.00)

I b REFERENCE

' NUS, Nuclear Energy Training, Module 3, Unit 6

! Westinghouse Reactor Physics, Sect. 3, Neutron Kinetics and Sect. 5 Core Physics HER, Reactor Theory, Sessions 20 and 24 - 31 Zion, NUS book 3, sections 6.5, 6.6, 6.7, 12.4, 12.5. ,

I l ANSWER 5.05 (1.00)

.1 d

l REFERENCE VEGP, Training Text. Vol. 9, pp. 21-60 & 61

! Westinghouse Reactor Physics, Sect. I-5.9 - 15 I

BER, Reactor Theory, Session 26, p. 2 1

1 5.06 (2.50)

ANSWER

a. 1 (Increases) [1.0]

b. 1. Acceptable core power distribution.

2. Limit the potential reactivity insertion (due to an ejected rod) l 3. To ensure core suberiticality after a reactor trip (adequate SDM)

[0.5 each] (1.5)

REFERENCE S.HNPP Technical Specifications 3.1.1.1; RT-LP-3.13, p 7-9.

SONGS TS 3.5.2; Lesson B-7, p 6-9.

DCC Technical Specifications 3.1.1.1; Uestinghouse Reactor Physics, Sect. I-5.9 - 15 l

ANSWER 5.07 (1.00)

Void, MTC, Doppler

5. THEORY OF NUCLEAR POWEP PLANT OPERATION. FLUIDS. AND PAGE 24 THERMODYNAMICS ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

REFERENCE Westinghouse Reactor Physics, pp. I-5.12, 25, and 27 SHNFP RT-LP-1.10, p 14,15.

SONGS Lesson B-1, p 38-51.

4 s

ANSWER 5.08 (1.00) b REFERENCE Westinghouse Reactor Physics, pp. I-5.26 & 27 SENPP RT-LP-1.10, p 13-15.

SCNGS Lesson B-1, p 38-51.

ANSWER 5.09 (1.00)

Has no effect on curve REFERENCE Westinghouse Reactor Physics, pp. I-4.19 - 24.

SHNPP RT-LP-1.7.

SONGS Lesson A-8, p 15-18.

ANSWER 5.10 (1.00) a REFERENCE HER, Reactor Theory, Sessions 41 and 42 DCC Westinghouse Reactor Physics, pp. I-4.11 - 14 ANSWER 5.11 (1.00)

a. False

b. True REFERENCE DCC Westinghouse Reactor Physics, pp. I-4.13 - 14

B. THEORY OF NUCLEAR POWER PLANT OPEEATION. FLUIDS. AED PAGE '25 THEF.MODYNAMICS ANSWERS -- CCCK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 5.12 (1.00)

Results in a higher flow rate and higher head loss.

4 REFERENCE CNTO, " Thermal / Hydraulic Principles and Applications, II", pp 10-45/48 DCC Thermodynamics Study Guide, p. 20 006/050; K5.01(2.9/3.1)

ANSWER 5.13 (1.50) a) decrease; (no answer) (+.25 ea response) b) (no ans); increase c) decrease; decrease d) decrease; decrease REFERENCE CNTO " Thermal / Hydraulic Principles II", pp 12-39-45 039/000; A2.05(3.3/3.6)

DCC Thermodynamics Study Guide, p. 40 I

ANSWER 5.14 ( .50)

False.

REFERENCE WNTC, Thermal-Hydraulic Principles and Applications, pp 13 - 18.

DCC Thermodynamics Study Guide, pp 53-54 ANSWER 5.15 (1.00) c.

REFERENCE Steam Tables i

5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 26 THERMODYNAMICS ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 5.16 (1.50)

a. FALSE

b. TRUE

c. TRUE REFERENCE General Physics, HT&FF, pp. 155, 319, and 320 and Subcooled Liquid Density Tables DCC Thermodynamics Study Guide, p. 16-17 ANSWER 5.17 (1.00) 245 +/- psig REFERENCE SENPP Thermo-LP-1.1 and steam tables KA002/000,K5.01,3.1 Cook-Westinghouse Thermal Science, Chapter 2,Pp 63-70,79. /020,K5.06,3,4 SONGS-GP-HTFF, p 83.

ANSWER 5.18 (1.00) c REFERENCE Steam Tables ANSWER 5.19 (2.00)

NI power would be adjusted to indicate a lower value than actual reactor power [1.0]. This is caused by the lowering of the average feedwater temperature vs. the indicated feedwater temperature used in the calculation, and/or AFW flow bypasses feedflow indication and lower mass i flow in the calculation would have the same result [1.0]. (2.0)

REFERENCE Steam Tables and Mollier chart

5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 27 HERMODYNAMICS ANSWERS -- COOK 1&2 -96/05/19-ISAKSEN, P.

I ANSWER 5.20 ( .50)

False REFERENCE SGNP, HTFF text p. 202 DCC Thermodynamics Study Guide, p. 34-35 Westinghouse Thermal Science, Chapter 13, pp 17-23 i

I f

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

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTPUMENTATION PAGE 28

, ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 6.01 (2.00)

Th u t bearing oil pressure low oft M ^ ^1D^ ^^ ^U ~ d ^

3. Overspeed

4. Low bearing oil pressure

5. Low pump suction pressure

6. Manual (local and remote)

7. Safety injection G. S/G hi-hi water level

9. Reactor trip [8 @ 0.25 each]

REFERENCE Cook RO-CO-FG10, Pg. 11; occ. - A-l-84 ANSWER 6.02 (1.00)

1. Contairment Isolation.

2. Loss of all charging pumps (breakers).

3. Low pressuriner level (17.0%).

VM/

4. Closure of letdown isolation valves (QR7-111, -112).

C- 4' A ^ ='d- a. (0.25each]

REFERENCE Cook RO-C-NSO6, Pg. 8; De L.a oe-t-r,39. 2.

ANSWER 6.03 (1.50)

1. RER HX

2. CCP (gear oil cooler, bearing oil cooler, seal plates).

3. SI pump (mech, seal HX, bearing oil cooler).

4. RER pump (mech. seal HX).

5. CTS pump (mech. seal HX). [0.3 each]

REFERENCE Cook RO-C-A501, Pg. 6 l

6. PLANT SYSTEMS DESTGN. CONTROL. AND I1!5TRUMENTATION PAGE 29 ANSWERS -- COCK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 6.04 (1.50)

1. Inject borated water into RCS when pressure < shutoff head of RHR pumps.C W j y -* ' n }

2. Provide suction supply to SI pumps and CCP's during recirculation phase.

1, e 3 - -_'-d_

3. Provide supplemental containment spray. [0.5 each] ,

'l. M %e - n m '=W= 6. M 9 ,- = ? ? ' _ * - =

REFERENCE Cook RO-C-NS08, Pg. 2 ; 50 Occ.- H P s s i, p s -S.

ANSWER 6.05 (2.50)

1. Diesel generator energized >80% voltage.

2. Zero Voltage en respective bus.

3. 6.9 KV feeder breaker open.

4. RCP bus tie breaker open.

5. D/G up to speed >95% or Run relay energized.

6. Control switch not in lockout position. [5 @ 0.5 each)

EurF M 2.o-C PG N, p 32; f.o -C - 4 58 a,, e ~J.*1 ANSWER 6.06 (2.00)

1. No gravity drains.

2. Suction lines are located near the SFP surface (19 1/2 inches below normal level).

3. The adjustable support for the skimmers limits downward travel to 6 inches below normal level.

4. Pump discharge lines terminate >/= 6 feet ABOVE fuel assemblies and include anti-siphon holes ~4 inches below normal level.

[0.5 each]

REFERENCE Cook RO-C-AS05, Pg. 2

PAGE 30

3. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION ANSWERS -- COCK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 6.07 ( .50)

Interlocked so that IMO-215 and -225 must be shut before the recire i aump suction valves ICM-305 or -306 will open [0.25], To prevent I draining the RWST to the recirc sump [0.25].

REFERENCE Cook'RO-C-NS15, Pg. 13 l ANSWER 6.08 (2.50)

s. 1. 3/3 condenser vacuum > 10.6" Hg.

2. At least one circ water pump operating.

3. Output voltage from CRID II, --Unit 1 only. [0.3 each]

b. 1. Tavg > 541 F [0.5]

2. Arming signal present:

a. Step load decrease of at least 10% or

b. Turbine trip [0.5 each]

REFERENCE Cook RO-C-PG12, Pg. 6-11 ANSWER 6.09 (1.00)

1. Steam flow is used to provide the program dP signal. ( Added tx) the no-load setpoint.) (0.5)

2. Steam pressure is used in conjunction w/ feed pressure to generate the actual dP signal. (0.5)

REFERENCE Cook RO-C-PG11, Pg. 7 l

l i

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6. PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE 31

. ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 6.10 (1.50)

a. Rods move IN, because Tavg is higher than Tref. (0.5)

b. rods move OUT, because Tavg becomes less than Tref. (0.5)

c. Rods _ move OUT, because the power mismatch circuit sees turbine power (as sensed by Pimp) increasing above Rx power. (0.5)

REFERENCE Cook RO-C-NSO4, Pg. 2 ANSWER 6.11 (2.00)

1. Power range high flux - 103% [0.2]

2. Intermediate range overpower - current equiv. to 20% [0.2]

3. OP Delta-T - 3% below setpoint [0.2)

4. OT Delta-T - 3% below setpoint [0.2]

5. Urgent Failure Alarm (no setpoint required) [0.1]

1 thru 5 affect both manual AND auto rod withdrawal. [0.5]

6. Turbine power < 15%. [0.2]

Control bank D withdrawal stop - #g.so

7. "Es steps. [0.2]

6 and 7 affect auto withdrawal only. [0.2]

REFERENCE Cook RO-C-NSO4, Pg. 8 ; I - m4 P 4o 3 # 38 0. o'4 0 i

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6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTR"MENTATTCN PAGE 32 ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 6.12 (2.00)

1. High pressure Rx trip, 2/4, >2378 psig

2. Low pressure Rx trip, 2/4, <1872 U-1, <1966 U-2

3. Low pressure SI, 2/3, <1837 U-1, <1908 U-2

4. Low pressure SI Block, 2/3, <1915 U-1, <1990 U-2

5. Provides input to OT Delta-T circuit

[Each correct response worth 0.125] (2.0)

REFERENCE Cook RO-C-NS03, Pg. 27 ANSWER 6.13 (3.00)

1. Low pressuri er pressure - LOCA, p_n steam.. break,(hr.; IS: e..;;::i. 5)

2. Lower containment high pressure - LOCA, steam or feed break inside containment.

3. High steamline Delta-P - steamline break upstream of stop valves.

4. a. Unit 1: High steamline flow coincident with low steamline pressure OR low-low Tavg. P-12 allows block of SI (but not the associated steamline isolation). - Major steamline break downstream of stop valves.

b.

Unit 2: Low steamline pressure 9P r' -t ::-lir. . __.

cc i r. : i d : '. t . _' . l .

. . - ! m , g-- P-12 allows block (of all but high steam flow isolation). - Steam break downstream of stop valves.

[0.2 points for each component of answer.) (3.0)

REFERENCE Cook RO-C-NS12, Pg. 15

6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 33

. ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 6.14 (1.00)

Store a volume of water (>/= 175,000 gal.) which will ensure hot standby conditions can be maintained for nine hours [0.5] following a total loss of offsite power with steam discharge to atmosphere [0.5].

REFERENCE Cook RO-C-ASil, Pg. 6 ANSWER 6.15 (1.00) 2, and 3, (to provide post-accident monitoring per T/S 3.3.3.8)

Modes 4.1. : 3f1,fwbin Mode 5, when </= 170 F,g if a PORV is required to be operable [M] . G asop, gy I.*

REFERENCE Cook TS 3.3.3.8 & 3.4.9.3 (U-1)

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7. PROCEERES - NORMAL, APNORMAL, EMERGENCY AND PAGE 34 RADIOLOGICAL CONTROL ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 7.01 (1.00) d es.a-EEFERENCE SENPP TS table 3.3-1.

' . 2. . t DC Cook TS Table 3. 3-1;, eWP -*f

• gag,os , W ANSWER 7.02 (1.50)

a. 4 (2.5)

b. 3 (7)

c. 2 (15.75) [0.5 each]

REFERENCE SHNFP RF-LP-1.3, p 10.

DC Cook FMP 6010 RAD.001, p 36.

ANSWER 7.03 (1.00) c REFERENCE Surry, EP-1.00, Foldout Page DC Cook 01-OHP 4023.E-0 Foldout.

ANSWER 7.04 (1.00)

C REFERENCE Surry, EP-2.00, Foldout, p. 16 DC Cook E-0 Fcidout.

ANSWER 7.05 (1.00)

S Containment pressure >l.1 psig or radiation >/OR C J._^ ^

%?- *:_: - ?J: y_ ~ ^ e t y_ oas RAD.

RWe l W. ALets co+ .Lk. W : un on esc, a **i . - & s, t t .

7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 35 RADIOLOGICAL CONTROL ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

?

REFERENCE CAF ANSWER 7.06 (2.50)

O. (six required 0.25 each) '

-IRPI rod bottom light

-IRPI alarm

-Rod sequence violation alarm

-PRNI flux deviation alarm

-rapid drop in Tavg

-rapid drop in PZR pressure (PZR heater actuation)

-Tavg-Tref deviation low alarm

-rapid drop in PZR level (letdown isolation)

-rate trip on PRNI

-computer NIS tilt alarm

-auto rod withdrawal

b. No indication of a power transient (NI power, rod motion, RCS parameter changes) [0.5]

c. By measurement of LVDT coil stack voltages on the affected rod.[0.5]

REFERENCE DC Cook 1-OHP 4022.012.004,-007.

ANSWER 7.07 (2.00)

a. Yes

b. Yes

c. No

d. Yes [0.5 each]

RE:sxENCE DC Cook FMI-4050, p 3,4.

_ - - - 9 . ,-

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t 7. PRCCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 37 l RADIOLOGICAL CONTROL i.

ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

l i

REFERENCE DC Cook Emergency Procedures and WOG ERG Background Document.

l ANSWER 7.11 (3.00)

n. T

b. T

c. F

d. T C. I

f. T [0.5 each]

REFERENCE DC Cook FMP 6010. RAD.001, p 80, 90, 122, 124. -

ANSWER 7.12 (2.50)

a. 1. 200

2. 0.2

3. 15

4. 325 [0.25 each]

b. Maintain sufficient backpressure on the No. 1 seal.(proper No. 2 seal flow) [0.5]

c. 1. High RCP bearing temperature

2. High seal leakoff temperature [0.5 each]

REFERENCE DC Cook 1-OHP 4021.002.003, p 3.

ANSWER 7.13 (1.00)

Immediately terminate the startup by commencing emergency boration (>10 gpm of 20,000 ppm boric acid >Suntil a shutdown margin of >1.6% delta k/k is obtained) [0.5) AND fully insert all control rods [0.5].

REFERENCE DC Cook **1-OHP 4021.001.002, p 10.

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7 PROCEDURES - NORMAL. AENORMAL. EMERGENCY AND PAGE 38 RADIOLOGICAL CONTROL ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 7.14 (1.00)

Insures that mass addition pressure transient can be relieved by the opera-P tion of a single 4 )ORV.

REFERENCE t .w DC Cook TS B 3/4.1.2.

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9. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39 i

. ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

l j ANSWER 8.01 (2.00)

1. Spill reported to plant authorities.

2. Isolate spill.

3. Contain and remove spill.

4. Complete Initial Oil and Chemical Spill Report and Notification.(Att. 1 REFERENCE DC Cook PMI-2230, p 3.

ANSWER 8.02 (3.00)

a. Interval requirement not exceeded [0.5]. Eight days does not exceed 1.25 times the specified interval [1.0].

b. Interval requirement exceeded [0.5]. The last 3 consecutive intervals exceed 3.25 times the specified interval [1.0].

REFERENCE SHNFP TS, p. 3/4 0-2, 5-10.

DC Cook TS, p 3/4 0-2.

ANSWER 8.03 (1.50)

1. The intent of the procedure is not altered [0.5].

2. The change is approved by two members of the plant management staff, at least one holds a SRO license [0.5](on unit affected)

3. The change is documented, reviewed by PNSRC and approved within 14 days of implementation by the Plant Manager [0.5].

REFERENCE SHNPP TS, p 6-17.

DC Cook TS, p 6-14.

ANSWER 8.04 (1.50)

a. 2 (0.5)

b. 2 (1.0) l l

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8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 40 :

, ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

REFERENCE SENPP, TS 3/4 2.1 DC Cook TS 3/4 2.1 ANSWER 8.05 (1.00)

Cp&b REFERENCE DC Cook TS , p 6-4) ow2.-4*8 8, f s .

ANSWER 8.06 (2.00)

a. F

b. F

c. T

d. F [0.5 each]

REFERENCE DC Cook PMI-2110, p 5, 9, 11, 12.

ANSWER 8.07 (2.50)

a. Plant Manager (or his designee) [0.5]

b. A supervisor may enter alone to inspect work in progress. [0.5)

c. 1- control room 2- control rocm (operator) 3- at the entry point [0.5 each]

REFERENCE DC Cook PMI-4010, p 5, 6.

ANSWER 8.08 (2.00)

a. Hot Standby and pressure in limit within 1 hr.

b. Pressure within limit in 5 min. [1.0 each]

REraxENCE DC Cook TS, 2.1.2.

6. APMlHISTRATf9E PROCEDURES. CONDITIONS. ANP LIMITATIONS PAGE 41

. ANSWERS -- COOK 1&2 -86/05/19-ISAKSEN, P.

f ANSWER 8.09 (2.50)

a. 1. MTC in analysed temperature range.

2. Protection instrumentation in normal range. s e_ ; - - - '

3. [0.5 each]

b.

"- Tavg? M P 2 L :- -

m .abovetheF-1ginterlocksetpoint.

hi Yavg within'11mit [0.35) in fifteen minutes [0.15] or be in d'dstore Hot Standby [0.35] within next fifteen minutes [0.15].

REFERENCE DC Cook TS, 3.1.1.5 and bases.

ANSWER 8.10 (2.00)

a. A copy of the procedure must be present during the performance of the procedure [0.5]

b. By an on the spot change sheet [0.5] or by use of N/A's [0.5] if allowed by the procedure (or by the PMI dealing with the procedure series) [0.25]. The on the spot change sheets must be used if the precedure becomes unworkable or no allowances are provided for N/A's

[0.25]

REFERENCE DC Cook PMI-2010, p 1, 2.

ANSWER 8.11 (1.00)

To prevent the release of activity [0.4] in event of a SGTR [0.3] since the saturation pressure of the RCS is below the lift pressure of the atmospheric steam relief valves [0.3].

REFERENCE DC Ccok TS, p B 3/4 4-5.

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8. PAGE 42 4DMINISTRATIVE PROCEDURES. CQNDITIONS. AND LIMITATIONS

. AUSWERS -- CCOK 1&2 -86/05/19-ISAKSEN, P.

ANSWER 8.12 (3.00)

n. Identified '

b. Identified

c. Pressure boundary

d. Controlled

e. Unidentified

f. Identified [0.5 each]

REFERENCE DC Cook TS, 3.4.6.2 and Section 1.

ANSWER 8.13 (1.00)

a. SS

b. .CJr? ^ ? ^ 28%0.5 each]

er Mr ::.s**

REFERENCE DC Cook FMP 2080 EPP.015, p 2; o ++I-4083,6

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