ML20148E832

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Exam Rept 50-445/OL-87-02 During Wk of 871214.Exam Results: Three of Five Senior Reactor Operator Candidates Passed Exams.Written Exam Waived for One Candidate
ML20148E832
Person / Time
Site: Comanche Peak Luminant icon.png
Issue date: 01/13/1988
From: Pellet J, Whittemore J
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20148E805 List:
References
50-445-OL-87-02, 50-445-OL-87-2, NUDOCS 8801260123
Download: ML20148E832 (44)
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Text

' APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV 0perator Licensing Exam Report:- 50-445/0L 87-02 Docket: 50-445 ,

Licensee: Texas Utilities Generating Company 400 North Olive Street Lock Box 81 Dallas, Texas 760 Facility Name: Comanche Peak Steam Electric Station Examination At:

Chief Examiner: ~

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/ O!}f J$hn E. Whittemore, Operator Date Licensing Section, Division of Reactor Safety Approved By: M 5, / /3 I

d. L. Pellet, Chief, Operator IIate Licensing Section, Division of Reactor Safety Summary:

NRC administered five (5) Senior Reactor Operator (SRO) licensee examinations at CPSES during the week of December 14, 1987. The written examination was waived for one candidcte. Three candidates passed their examinations and have been issued the appropriate license, i

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8801260123 880115 hDR ADOCK 05000445 PDR l

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DETAILS.

1. PERSONS EXAMINED

' Licensee Examinations: SR0 Pass TE7%

Fail 2-33%

2. EXAMINERS J. E. Whittemore (Chief Examiner)

T. P. Guilfoil F. W. Victor

3. EXAMINATION REPORT
a. EXAMINATION REVIEW COMMENT RESOLUTION In general, editorial comments or changes made durir.g the examination, review, or subsequent grading reviews are.not addressed by this resolution section. 'This section reflects resolution of substantive comments made by CPSES. The only coments addressed in this section

- are those which were not accepted for incorporation .into the examination and/or answer key. Those coments accepted are incorporated into the master examination key which is included in this report. Comments from CPSES may be paraphrased for brevity. The full text of the comments is attached.

5.06 The question is similar to 5.04 a.nd should be dropped from the examination. The question is also double jeopardy between Parts a. and b. Adjust answer in Part b to allow

- answer given in Part a. The question is totally confusing in that the question refers to the ratio of PU-239 and PU-240 atoms to U-235 atoms in relation to delayed neutron fraction and reactor period, however this ratio has little or nothing to do with the doppler coefficient. The ratio of PU-239 and PU-240 to U-238 does. Therefore Part c. should be dropped from the exam and points adjusted. If question is allowed, no point value is assigned for the how. Reassign point values for all 5.06 for equal weighting for a, b, and c.

2 Resp; Partially accepted. The only change made to the key was to allow answer in Part b. to reflect assumption made.in Part a.

The candidate's performance was 95 percent regardless of the change. Point value for Part c. was not allowed as this same information was solicited-in Question 5.04. The comment about similarity to Question 5.04 is not understood as the only comon thread is doppler coefficient, which candidates I are asked to evaluate from two different perspectives.-

5.13 The wording "action" may have confused some examinees and led them to think of indications. Maximum grader discretion.

Also accept c. for full credit because the action of raising RCS temperature means there is a greater delta T between the heat source and the heat sink and the thermal driving head is proportional to the delta T.

-Resp: Coment not accepted. Deliberate operator action to enhance natural circulation should be to carefully decrease heat sink temperature vice increase RCS temperature.

6.11.c.1 Change answer to'yes because answer concerns the basic flow path of the Containment Air Cooling and Recirculation System (CACRS) during a LOCA condition. Since the CACRS is tripped upon receipt of an "S" signal, this does represent a significant change in the basic flow path as air will no longer be drawn into, cooled by, and discharged into the CACRS plenum.

Resp: Not accepted. Flow path does not change in the interim between trip and restoration. Only the. flow. conditions change.

7.04.a The PDP and PCV-131 are the major components manipulated to form a bubble in the pressurizer. Change answer to accept PDP.

Resp: Not acceptsd. The question solicits valves that are manipulated. The answer was expanded to include only 2 out of 3 possible valves.

7.06 Answer d. is incorrect. Steam generator becomes "faulted" implies a major cooldown of the RCS, and depressurization of the steam lines. Both will actuate SI which is equivalent to emergency boration.

3 Resp: Not accepted. SI is not the equivalent of operator-initiated emergency boration.

8.03.a Answer should include Boration Systems Flowpath, TS 3.1.2.1-and 3.1.2.2.

Resp: Not accepted. This Technical Specification is not applicable to the stated scenario.

8.03.b The answer for this question should read required by Technical Specifications, or necessary for inventory control.

Resp: Not accepted. Inventory control is'always a basis and is acceptable for no credit. However the basis in the Technical Specifications does not say "required by Tech.

Spec's," and will not be allowed.

b. EXIT MEETING

SUMMARY

An exit meeting was held on December 17, 1987, to conclude the site visit. The following personnel were present:

NRC LICENSEE J. E. Whittemore (Chief Examiner) C. L. Turner M. A. Niemeyer-C. M. Rice The following simulator problems were noted to'the licensee:

(1) The simulator suffered a complete failure of the primary computer, but the backup system functioned satisfactorily to complete the examinations.

(2) When rod speed was failed to 72 steps per minute, automatic and manual rod control was lost.

The following apparent problems or areas of weakness were observed by by the examiners:

l (1) Several candidates were unfamiliar with, and could not find out the crux of the most recent revision (5) to the facility Technical Specifications.

(2) At least 2 candidates could not manually calculate the Estimated Critical Rod Position for a given set of conditions prior to a reactor startup, i

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c. -EXAMINATION MASTER COPIES The'SR0 master examination and answer keys follow,
d. CPSES EXAMINATION REVIEW COMMENTS

.CPSES examination review comments are attached.

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i-U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _QQdeUQUg_EgeE_1_________

REACTOR TYPE: _EW8:WEQd________________

DATE ADMINISTERED:_SZl12/15________________

EXAMINER: _WHIIIgdQBgz_Jz__________

CANDIDATE: _________________________

INSIBUCIl005_IQ_CeUQ10eIE1 Use separate- paper for the answers. Write answers ors 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 locst 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY  % OF CANDIDATE'S CATEGORY

__VeLUE_ _IQIeL ___500BE___ _VeLUE__ ______________CeIEQQBl_____________

_2520D__ _25200 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIOS, AND THERMODYNAMICS

_25z00__ _25z00 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION

_252DD__ _25200 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

.252DD__ _25 ADD ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 10Qzg0__ ___________ ________t Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.

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Candidate's Signature I

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v St__IHEQBX_QE_NUCLEoB_EQWEB_EL6NI_QEE86IIQNt_EL91QSt_6NQ PAGE 2 IBEBdQQ1NedIGS QUESTION 5.01 (1.50)

How'AND why will Axial Flux Difference change if reactor power is reduced from 100% to 50%7 Assume the reactor is operating at 100% power with all rods out'(ARO), early in cycle life at equilibrium Xenon conditions when power is reduced to 50% by borating (no rod motion). Neglect changes due to Xenon. (1.5)

ANSWER 5.01 (1.50)

Due to the greater decrease in the temperature of the coolant exiting the core' relative to the decrease of the inlet coolant ( 0. 5 ), more positive reactivity will be added in the upper core regions (0.5), resulting in a more pos it ive (less negative) AFD [0.51 (1.5)

REFERENCE CP Thermal Hydraulic Principles, Pp. 13 13-49 193009K102 ...CKA'S)

QUESTION 5.02 (2.50)

a. How would inserting a control rod bank 15 steps (from ARO) affect each of the parameters below? Describe relative values at the end of the transient if the plant is initially operating at 100% power early in cycle life and all other parameters are normal for this condition.

Assume the reactor does not trip and no further operator or automatic action occurs. Neglect effects of Xenon.

1. Reactor power. (0.75)
2. RCS Tave. (0.75)
b. Now and why would the above responses differ at end of life? (1.0)

51__IBEQBY_QE_N99LE68_E9 WEB _EL6NI_9EEB6_10Nt_ELu1Q$t_6NQ PAGE 3 IBEBdQQXNedICS ANSWER . 5'. 0 2 (2.50)

c. 1. (Reactor power will. initially decrease (0.35) then) return to (near)'the original value (100%) (0.4). (0.75)
2. Tave will decrease Cuntil reactor power returns to 100%) (0.75)
b. The only difference is that Tsve will decrease less (and power may decrease less) [0.5) as power defect becomes more negative with age.

[0.5) Also accept "Transient will be faster" for no credit (1.0)

REFERENCE CP Fundamentals of Nuclear Physics, Pp. 6 6-55 192004K106 192008K124 192004K107 ...(KA'S)

QUESTION 5.03 (2.00)

a. How will the reactor response differ to equal positive reactivity insertions as the reactor approaches criticality? (1.0)
b. How much reactivity must be added to a critical reactor for it to be considered prompt critical? (Actual value is NOT required.) (0,5)
c. Why are extrinsic neutron sources installed in the reactor core?(0.5)

ANSWER 5.03 (2.00)

a. The neutron population increase is greater (0.5) and the time to stabilize is longer (0.5) as Keff approaches 1. (1.0)
b. Beta (accept number). (0.5)
c. Neutron sources are required to assure sufficient neutron indication or level to control the startup. CONCEPT (0.5)

REFERENCE CP Fundamentals of Nuclear Physics, Pp. 8 8-62 192003K101 192003K111 192003K108 ...(KA'S)

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

St__IHEDBY_QE_ NUCLE 88_E9h'EB_EL8HI_QEEB8IIQNt_ELUIDSt_8NQ PAGE 4 IHEBdQQ1NedICS QUESTION 5.04 (1.00)

Which one of the following statements most correctly describes the change in the Doppler and Moderator Temperature Coefficients (FTC and MTC) as a function of core life CBOL to EOL)?

a. FTC becomes more negative and MTC becomes more negative,
b. FTC becomes more negative and MTC becomes more positive.
c. FTC becomes more positive and MTC becomes more negative.
d. FTC becomes more positive and MTC becomes more positive.

ANSWER 5.04 (1.00)

"a" (1.0)

REFERENCE EQB Generic HT & FF 192004K107 ...(KA'S)

-QUESTION- 5.05 (1.00)

Why will a decrease in boron concentration cause the Moderator Temperature Coefficient (MTC) to change. Include in your answer the physical effects of a unit temperature change and the associated effect on reactivity and MTC. (1.0)

ANSWER 5.05 (1.00)

For a given temperature change, less boron is expanded out of the core at lower concentrations.(0.50] Therefore, less positive reactivity addition in seen (0.25] and the MTC becomes more negative.[0.25]

Accept explanation of the effect on the Thermal Utilization Factor (f) for full credit. (1.0)

REFERENCE EQB GENERIC RT.

192004K107 ...(KA'S)

(***** CATEGORY 05 CONTINUE 0 ON NEXT PA0E *****)

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. St__IBEQBY_QE_NUQLE88.EQWEB_ELeNI_QEEB6IIQNt_ELU10St_8NQ PAGE IHEBdQQXNedIQ1 s-

- QUESTION 5.06 (2.50)

The, ratio of'the PU239'and PU240 atoms to U235 atoms increases over core life. -Explain how AND why this increasing ratio changes the following:

a. Delayed neutron' fraction (1.0)
b. Reactor. period (1.0)
c. Doppler Temperature Coefficient -(0. 5 ) -

- ANSWER 5.06 (2.50) a.- - Delayed neutron. fraction decreases (0.5) because the beta is'less for PU239'as compared to U235. (0.5] (1.0)

b. Shorter reactor period (0.5) because. delayed neutron fraction decreases. [0.5) Accept longer' period if assummed an increase in beta in part a above. (1.0)
c. (Doppler Coe f'f ic ie nt is more negative because) PU240 has a higher resonance cross section than U235. ( 0,5 ) .

REFERENCE CP Fundamentals of Nuclear Physics, Pp. 6-11 - 6-15 192004K107 192003K106 192003K104 . ..(KA'S)

QUESTION 5.07 (1.00)

Concerning equilibrium Samarium-149 (Sm) reactivity, which of the following

, statements is correct? 50% equilibrium Sm reactivity is:

a. one-quarter of 100% equilibrium $m reactivity.

L b.. one-half of 100% equilibrium Sm reactivity, i

c. three-quarters of 100% equilibrium Sm reactivity.

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d. equal to 100% equilibrium Sm reactivity.

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Si__IBEQBY_QE_NUCLEoB_EQWEB_EL6NI_QEEBoIIQNt_ELUIDSt_6NQ PAGE 6 IBEBdQQYNedICS-

-ANSWER 5.07 (1.00)

"d" REFERENCE EQB Generic RT 192006K115 ...(KA'S)

QUESTION 5.08 (2.50)

Answer TRUE or FALSE to each of the following statements concerning Xenon.

a. At full power, with equilibrium conditions, approximately one half of the xenon is produced by iodine decay and the other half is produced directly from fission.
b. When a reactor is shutdown, xenon burnup effectively stops while the decay.of iodine continues. Therefore, the xenon concentration starts to increase.
c. The production of xenon from iodine decay continues for up to approximately 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> after a reactor shutdown. Because all production of xenon has ceased at this time, xenon concentration reaches its minimum level in the reactor core,
d. Xenon production and removal increase linearly as power level increases; i.e., the value of 100 percent equilibrium xenon is twice the value of 50 percent equilibrium xenon.
e. The primary method of dampening Xenon oscillations is to follow secondary load changes by boration and dilution while holding control rod position constant. (2.5) l l

' ANSWER 5.08 (2.50) l o. False

b. True
c. False
d. False
e. False (0.5 ea.] (2.5) l l

St__IBEQBY_9E_NUGLE68_EQWEB_EL8NI_QEE86IIONt_ELVIQ3t_6NQ PAGE 7 18E86001860101

, REFERENCE EQB Generic RT 8 IPO 003A, Att. 3 192006K113 192006K107 192006K106 192006K104 192006K103

...(KA'S)

QUESTION 5.09 (1.00)

Provide and explain the 2 dynamic mechanisms or phenomena that cause the effect of the core delayed neutron fraction to differ from the fraction itself. (1.0)

ANSWER' 5.09 (1.00)

-1. Delayed neutrons are created at a lower energy and therefore less likely to cause fast fission. (0.5)

2. Lower energy delayed neutrons are less apt to leak out of the core before thermalization. (0.5)

Allow 1/2 total credit for answer of Importance Factor CI-Bar)

REFERENCE CP Fundamentals of Nuclear Physics, Pp. 7 7-35 192003K107 ...(KA'S)

QUESTION 5.10 (2.00)

TRUE or FALSE?

a. During high power operation, the secondary
  • mass (steam) flow rate changes throughout the turbine. (0,5) l t
b. The secondary side enthalpy change (delta h) across the steam generator decreases with increasing power. (0.5)
c. During a reactor startup, reactor coolant mass flow rate in the loops with operating reactor coolant pumps (RCPs) increases as each additional RCP is started. (0.5)
d. 'When operating with three RCPs running, reactor coolant in the loop with a stopped RCP is all at Tcold (no delta T). (0.5)

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St__IHEQBY_QE_ NUCLE 88_EQWEB_EL6NI_QEEB8IIQNt_ELUIQ$t_6NQ PAGE 8-

!: IHEBdQQXNedICS f

ANSWER -5.10 -(2.00)

a. T'
b. T
c. F
d. T [0.5 ea.) (2.0)

REFERENCE

. Thermal-hydraulic Principles I & II

~193006K113 193001K104 -...(KA*S)

QUESTION 5.11 (1.50)

What steam generator pressure is required to maintain 200 deg's F subcool-ing margin in the RCS when RCS pressure is 595 psig. Show all work. (1.5)

ANSWER 5.11 (1.50)

1. Add 15 psi to 595 psig = 610 psia (0.25)
2. Using steam tables, 8 610 psia, Tsat = 488 +/- 2 deg's F (0,5)
3. Tsat in S/G = Trcs - Tsubcooling = 488 - 200 = 288 +/- 2 deg's F(0.25)
4. Using steam tables, Psat 8 Tsat = 288 deg's F = 56 +/- 4 psia (0.5)

REFERENCE Steam Tables 002000K515 ...(KA'S)

QUESTION 5.12 (1.50)

a. Describe REFLUX BOILING mode of core cooling in terms of coolant flow path. (1.0)
b. List TWO Reactor Coolant System plant conditions that must be present for REFLUX BOILING to occur. (0.5)

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A St__IBEQBX_QE_NUQLEeB_EQWEB_ELeNI_QEEB6I1QNt_EL91QSt_6NQ PAGE 9 IHEBdQQ1N8dIQ1 -

ANSWER 5.12 (1.50)

a. Reflux Boiling is when steam exits the core.and is condensed in the SG

. tubes, with the resulting condensate returning to the core via the hot leg to repeat the cycle. (1.0)

b. This type of cooling occurs with:
1. voided core or saturated RCS
2. no reactor coolant pumps running
3. -secondary heat sink 4.- interruption-of natural circulation [any 2, 0.25 ea) (0.5).

REFERENCE

EQB WEC-4 Generic ~HT & FF 000011K101 193008K124 ...(KA'S)

QUESTION 5.13 (1.00)

Which one of the following actions may help, rather than hinder, natural circulation?

a. Lowering S/G 1evel-
b. Lowering RCS pressure
c. Increasing RCS temperature
.d.= Increasing PZR level l

ANSWER 5.13 (1.00)

"d" i

! REFERENCE CP Thermal-Hydraulic Principles, Pp. 14 14-29 193008123 ...(KA'S) l e

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

-St__IBEQBY_QE_NUCLEeB_EQWEB_ELoNI_QEEBoIIQNt_ELUID$t_6NQ PAGE '10 IHEBdQQXNedICQ

-QUESTION 5.14 (1.50)

Answer each of the following as TRUE or FALSE concerning natural circulation.

c '. Feeding steam generators too fast can reduce RCS system flow.

b. Two phase flow (RCS system at saturation) can be sufficient to ensure heat removal,
c. An indication of natural circulation will be RCS T-hot near saturation temperature for steam generator pressures.

ANSWER 5.14 (1.50)-

c. TRUE
b. TRUE
c. FALSE REFERENCE CP Thermal-Hydraulic Principles, Pp. 14-26 _ 14-29 193008K123 ...(KA'S)

QUESTION 5.15 (1.00)

Which one of the following represents the maximum linear power density which would be expected in the' core during full power operations?

a. Local Power Density multiplied by Hot Channel Factor.
b. Radial Peaking Factor multiplied by the Local Peaking Factor,
c. Average Kw/ft for the core multiplied by the Hot Channel Factor,
d. Hot Channel Factor multiplied by the Maximum Local Power Density.

(1.0)

ANSWER 6.15 (1.00) n

.. e l

i

- .-- . - - - - , . , , , . ,, e , - - - , . , - , . , - '- , - , - - - , - - .,. - -

Si__INEDBl_DE_NUCLEeB_EQWEB_EL8NI_QEEBeIIONt_ELUIQ1t_88Q PAGE 11 IHEBdQDINedICS.

REFERENCE CP Thermal-Hydraulic Principles, P. 13-10 193009K107 ...CKA'S)

QUESTION 5.16 (1.50)

For a centrifugal. pump operating in a closed loop system, how will the available AND the required Net Positive Suction Head vary as the pump speed is changed with other parameters constant. Explain your answer. (1.5)

ANSWER 5.16 (1.50)

As pump speed increases.the potential for cavitation is greater so the minimum NPSH required increases. [0.51 With increasing flow, there are increasing head losses at the pump suction, (0.5] so the available NPSH decreases. [0.5] (1.5)

REFERENCE CP Thermal Hydraulic Principles, Pp. 10 10 57 191004K113 ...(KA'S) l l ,

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- 61_ EL8HI_SYSIgbS_ggSIgNt_GQNIBQLt_8ND_INSIBudgNI8IION PAGE 12 QUESTION 6.01 (2.50)

Concerning.the Reactor Coolant Pump (RCP) seal. package:

c. What are the normal pressure drops across each of the 3 different seals? (0.75) b.. Why must the operator observe and maintain the specified minimum preseure across the #1 seal? (0,5)
c. What is the immediate destination (flow path) for any leakage through each of the 3 different seals? (0.75)
d. What is the purpose of the bypass flow path around the #1 seal? C0.5)

ANSWER 6.01 (2.50)

a. #1-------- Accept 2100 - 2300 psi
  1. 2-------- Accept 15 - 40 ps i
  1. 3-------- Accept 0- 3 psi (0.25 es.) (0.75)
b. Must maintain minimum psid so seal surfaces will not contact as it is designed as a film riding seal. (0.75)
c. #1 --- VCT
  1. 2 -- RCOT
  1. 3 --- Containment Sump (0.25 es.) (0.75)
d. To ensure adequate cooling for the lower radial bearing (when #1 seal flow is insufficient). (0.5)

REFERENCE CP NSTM, Pp. II-1.25A, 26.A, 003000K103 ...(KA'S)

(***** CATEGORY 06 CONTINUE 0 ON NEXT PAGE *****)

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'6t__EL8NI_SYSIEdS_QESIGNi_QQNIBQLi_8ND_INSIBudENI6IIQN PAGE 13

. QUESTION 6.02- (2.50)

TRUE or FALSE:

a. The diesel generator output breaker is not affected by an Anti-Parallel interlock feature.
b. If a safety injection occurs while.a diesel generator is being load tested, the bus will deenergize and and be normally reenergized by the diesel generator.
c. Following a loss of voltage to a 6.9 KV non-safeguards bus, a fast transfer will not INITIATE if the loss of voltage was due to a lockout function.
d. On 6.9 KV safeguards buses, the SLOW TRANSFER does not function from.

alternate to normal power supply,

e. Eech feeder breaker control switch for 6.9 KV safeguards buses has an associated synchronizing switch. -(2.5)

ANSWER 6.02 (2.50)

a. True
b. False
c. False
d. True
e. True (0.5 es.] (2.5)

REFERENCE CP NSTM VII-1.25 - VII-1.38 062000K403 062000K401 ...(KA'S)

QUESTION 6.03 (1.50)

Following loss of off site power and subsequent reactor trip, it is noticed both of the diesel generators started but 1 unit's output breaker did not shut. What are the logic conditions necessary for this breaker to close automatically. Condition values are not required. (1.5)

ki__EL8NI_SYSIEdS_QESIQNt_GQNIBQLt_8ND_INSIBudENI8IIQN PAGE 14 ANSWER 6.03 (1.50)

1. Proper' generator speed
2. Proper generator voltage 3.- Normal and alternate breakers open
4. No lockout (fault) on bus (any 3, 0.5 es.) (1.5)

REFERENCE CP NSTM XII-4.42 064000A401 062000K401 ...(KA'S)

QUESTION 6.04 (2.00)

8. What is the specific design function provided by the Emergency Core Cooling Systems (ECCS) for:
1. Loss of primary coolant
2. Loss of secondary coolant (1.0)
b. What are the 4 specific accidents that the ECCS is supposed to provide mitigation for? (1.0)

ANSWER 6.04 (2.00)

n. Primary: Provide core cooling. CONCEPT Secondary: Provide SOM (boration) CONCEPT (0.5 ea.) (1.0)
b. 1. LOCA
2. Rod eject.
3. Secondary coolant break
4. SGTR [0.25 ea.) (1.0)

REFERENCE CP NSTM II-8.2-4 006020K404 006000K404 006000A303 ...(KA'S)

QUESTION 6.05 (2,50)

Describe the response of the Component Cooling Water system to a HIGH-3 "P" signal with accompunying Phase "B" signal.

62__EL8NI_SYSIEdS_DESIGNt_CQNIBQLt_8ND_INSIB9dENI6IIQN PAGE 15 ANSWER 6.05 ( 2. 5 0 ) _

l. Non-safeguards loop isolation valves shut.
2. Train "A" and"B" cross-connect valves shut.
3. RHR H/X valves open
4. Containment Spray H/X valves open.

S.- Contmt isolation for RCP's shut. (0.5 es.) (2.5)

REFERENCE CP NSTM II-12.12 013000K107 013000A302 008000K401 ...(KA'S)

QUESTION 6.06 (2.50)

c. The operator reports there is a demand for automatic outward rod motion but the control rods are not moving. You order shifting to manual rod control where the rods move normally. What are 2 interlocks that may have prevented withdrawal in auto? (1.0)
b. For control of the rods in automatic the power mismatch signal is conditioned by a VARIABLE GAIN UNIT:
1. What is specifically accomplished by this component?
2. Why is the component necessary? C0.75)
c. For the following failures indicate if each would be annunciated as Logic or Power Cabinet, AND Urgent or Non-Urgent failure:
1. Full current is applied to a coil for excessive time.
2. A pulser fails to generate a pulse when signaled.
3. A single power supply module in a logic cabinet fails. (0.75) i

ni__EL8HI_SXSIEdS_DESIGNt_CQNIBQLt_6ND_INSIBudENI6IIQN PAGE 16 l

JANSWER 6.06 (2.50) a.- 1. Low power block (Accept C-5)

'2. Bank "D" withdrawal limit (Accept C-11) (0.5 ea.] (1.0) b.

~

1. Imposes a high. gain at low power levels (and a low gain at high power levels.) (0.35)
2. The effect of rod motion is greater at high power than at low power. (0.4)
c. . 1. Power cabinet urgent failure.
2. Logic cabinet urgent failure. '
3. Logic cabinet non-urgent failure. (0.25 es.) (0.75) .

REFERENCE ,

CP NSTM III-3.11, 14, 001050A201 001000K407. 001000G007 00100G008 ...(KA'S)

QUESTION 6.07 (2.50)

o. In addition to interlock C-9, what are 3 conditions that must be s at is f ied (values not required) to cause the steam dumps to ,

automatically.open on a load r ej ect ion? Assume no reactor trip, no operator action and steam dumps in automatic and Tave mode. (1.5)

6. With C-9 satisfied in automatic and the steam pressure mode, what are 2 conditions that must be met to automatically open steam dumps?

Assume no operator action. (1.0) i ANSWER 6.07 (2.50)

a. 1. Teve > lo-lo setpoint.
2. C-7 Caccept explanation)
3. Temperature error (Tave - Tref) (0.5 es.) (1.5)
b. 1. Tave > lo-lo setpoint.
2. Steam pressure > setpoint. (0.5 es.] (1.0)

REFERENCE CP NSTM III-7.15 041020K417 041020K105 041020A404 041020A403 041020A402 041020A101 ...(KA'S)

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

6t__EL6NI_SYSIEdS_DESIONt_GQNIB9Lt_6HD_INSIBudENI6I1ON PAGE 17 QUESTION 6.08 (2.00)

The unit- has j ust tripped. You are provided the following data that existed immediately prior to the trip.

Reactor power: 95% -

Unit auxiliary transformer output voltage: 4810 VAC Loop flows: 93% - 96%

Turbine trip system oil pressure: 43 psig

-Pressurizer Pressure: 2315 psig Steam Generator levels 45% - 47%

List the possible SPECIFIC causes of the trip based on the data provided.

Include applicable setpoints. (2.0)

ANSWER 6.08 (2.00)

1. RCP undervoltago (due to UAT low voltage) (0.75) The trip setpoint is 4830 VAC. (0.25]
2. Turbine trip above P-7. (0.75) Setpoint is 45 psig. (0.25] (2.0) l l

REFERENCE CP NSTM III-9.64 045000K411 045000A304 012000K402 012000G014 012000G012

...(KA'S)

QUESTION 6.09 (2.50)

A startup to full power is to be conducted from a present condition of 50 CPS in the source range. Describe the permissives/ interlocks associated with the nuclear instrument system that must be met during the startup.

Include in the description; SETPOINTS, COINCIDENCE, AUTO ACTION and MANUAL ACTION that is taken when the inte r lock /pe rmiss ive is met. (2.5) l l

1 l

l l

I

ht__EL6NI_SISIENS_DESIGNt_QQNIBQLt_680_INSIBudENI8IIQN -PAGE' 18 ANSWER 6.09 (2.50)

P-6 [0.25] Coincidence: 1/2 intermediate range (0.25]

Setpoint: > 1E -10 amps (0.25]

Action: Block Source range trips (0.5)

Accept: Block flux doubling-no credit P-10 (0.25] Coincidence: 2/4 Power range (0.25]

Setpoint: > 10% [0.25)

Action: 1. Block intermediate range trip (0.25)

2. Block power range low trip [0.25)

' Accept explanation of P-8 for no credit. (2.5)

REFERENCE 1 CP NSTM III-1.

015000K407 015000K406 ...(KA'S)

QUESTION 6.10 (2.00)

n. The High Pressurizer Level reactor trip is automatically blocked during shutdown. Why is it necessary to be able to block this trip when the reactor is shutdown? C0.5)
b. Describe the logic and coincidence necessary to automaticelly block the trip during shutdown. (0,5)
c. What other trips are blocked or unblocked simultaneously? (1.0)

ANSWER 6.10 (2.00) t

a. Allows testing such as control rod testing at low temperature or withdrawal of SD rods at low temp. CONCEPT (0.5)
b. 3/4 Power range channels less than 10% [0.2) AND [0.1) 2/2 Turbine
impulse power less than 10%. (0.2] (0.5) i l
c. 1. Low Pressurizer Pressure.
2. Turbine trip
3. RCP undervoltage
4. RCP underfrequency
5. Low flow (Allow 0.6 for RCS flow trips in lieu of 3, 4, and 5, otherwise, 0.2 ea.)

A

- 6t__EL6HI_SYSIENS_DESIGNt_GQNIBQLt_6ND_INSIBudENI6IIQN .PAGE 19 REFERENCE CP NSTM III-9.15, 25 012000K610' 012000K604 ...(KA'S)

QUESTION 6.11 (2.50)

Answer.the following concerning the Containment Cooling System:

n. What are 3 pieces of equipment, components, or spaces cooled by the "Neutron Detector Well Cooling System", and how are the fans affected by an "S" signal? (1.0)
b. What is the normal status'of the Control Rod Drive Mechanism Ventilation fans, and how does the status change in the event of "S"'and Blackout signals? (1.0)
c. Answer yes or no: L
1. Does the basic flow path of the Containment Air Cooling and Recirculation system change during a LOCA with Safety Systems actuated? (0.25) 2.- Does the Reactor Coolant Pipe' Penetration Cooling System require 2 of'the 4 fano to provide full cooling. capacity? C0.25)

ANSWER 6.11 (2.50)

a. Nuclear detectors, Reactor vessel supports, reactor cavity, and cavity annulus. (any 3, 0.25 ea) The fans trip and are not sequenced on an "S" signal. [0.25) (1.0)
b. Normal: 1 fan operating. (0.25)

"S" signal: Fan trips. [0.25)

8. O. signal: Fans trip and sequence on to safety buses. (0.5) (1.0)
c. 1. No
2. Yes (0.25 ea.) (0,5)

REFERENCE  ;

CP NSTM IX-3A.1 -10 022000K402 022000A301 ...(KA'S)

(***** END OF CATEGORY 06 *****)  ;

l

Zt__RB9G50MBE1_ _N0Bd6Lt_eHNQBd8kt_EdEBGENGI_6ND PAGE 20 88MAjt09!Q6L_QQNIBQL QUESTION 7.01 (2.50)

c. During refueling or fuel shuffle, procedures allow the_ operator to move the refueling machine "0FF INDEX" to insert or withdraw fuel assemblies. Why must the operator have this leeway? (0.5)
b. If a grappled fuel assembly cannot be placed in it's specified core location, what are 2 locations where the assembly may be' temporarily.

stored within the containment? (1.0j

c. Name the positions that must be filled for a minimum designated refueling team. (1.0)

ANSWER 7.01 (2.50)

a. To facilitate the insertion and removal of fuel assemblies that have become bowed.. (0,5)
b. 1. RCCA change fixture
2. Alternate core location
3. Upender Cany 2, 0.5 es.) (1.0)
c. 1. Fuel handling supervisor
2. Refueling Machine Operator
3. RCCA Change fixture / transfer sys. operator
4. Fuel bldg. bridge crane operator
5. New fuel elevator / transfer'sys. operator (0.2 ea.) (1.0)

Cntmt bridge crane operator accepted but no credit.

REFERENCE CP RFO-302, Pp.5, 11 034000G001 034000A203 ...(KA'S)

QUESTION 7.02 (2.00)

For RCS partial draindown conditions to 1/2 loop level:

c. What are 2 methods or systems of level indication? (0.5)
b. What affect may a nitrogon purge have on level indication? (0.5)
c. List 4 abnormal parameter indications the control room operator may j observe if an operating RHR pump experienced vortexing during partial

, draindown conditions. (1.0) l I

l

LZt__EBOCEDUBES_ _NQBd8Lt_8HNDBd8Lt_EdEBQENCY_8NQ PAGE. 21 88DIQLDQIG8L_CQNIBQL

~ ANSWER- 7.02 (2.00)

a. 1. Temporary poly hose / tubing (0.25)

'2.- Accept RVLIS or any other viable method used to attain-1/2 loop draindown conditions. [0.25) (0.5)

b. Pressure due to purge may cause level to indicate higher than actual level lif tubing vented to atmosphere) (0.5)
c.. 1. RHR Pump motor trip
2. Oscillating RHR flow
3. Oscillating RHR pressure
4. Increasing RHR temp.
5. Increasing RCS Temp.
6. Oscillating RCS pressure (any 4, 0.25 ea.) (1.0)

REFERENCE CP-SOP-101A, Pp11-14 & ABN-104A, P.6 002000K402 ...(KA'S) 4 QUESTION 7.03 (3.00)

Assume a normal shutdown from full power and match the events / occurrences in column "A" with the power level in column "8".where the event normally occurs. Power-levels may be used more than once or not at all. (3.0)

"A" "B"

a. Stop i Hester Drain Pumps. 1. 10%
b. Reduce condensate pumps to 1 operating. 2. 14%
c. Establish "Shutdown" electrical lineup 3. 20%
d. Place feedwater bypass control in automatic. 4. 25%
o. Transfer the steam dump system to the 5. 49%

pressure control mode

6. 75%
f. Remove MSR's from service

IZi__EBQQEQUBES_:_NQBdeLt_eBNQBdeLt_EdEBGENQX_6NQ PAGE 22

.88DIQLQQIQ8L_QQNIBQL ANSWER 7.03 (3.00) a.. 5, 4

b. 5, 4
c. 2, 1
d. 4, 3, 2
e. 2, 1
f. 3. (1-correct answer per step, 0.5 ea.) (3.0) l REFERENCE CP IPO-003A & IP0004A-194001A102 ...(KA'S)

QUESTION 7.04 (3.00)

According to the Start up/ Heat up procedures,

s. What 2 major valves are manipulated to form a bubble in the pressuriz-er AND how are they manipulated to commence bubble formation? ALS0-state applicable modes of valve control. (1.0)
b. What are 2 indications of bubble formation stated in the procedure?

(0,5)

c. Before commencing RCS heatup, SOP-8 cautions the operator that RCP's should be started at a pressure above 350 psig. What is the possible consequence of starting RCP's below this recommended pressure?

Explain. (0.5)

d. For the following events that occur during a normal startup, place them in the order they occur:
1. Commence power increase to 10 (E-8) amps.
2. Withdraw Shutdown Bank Control rods.

3 .- Commence withdrawing Bank "A" Control Rods.

4. Obtain baseline data for Neutron Doubling Curve Plot.
5. Adj ust RCS boron concentration.
6. Neutron counts double. (1.0)

Z1__EBQGEQUBES_:_NQBd8Lt_8BNQBd8Lt_EdEB9ENQY_8NQ PAGE 23 88DIQLQQIC8L_CQNIB9L ANSWER 7.04. (3.00)

.o.- The Charg'ing Flow Control valve (0.25] is (closed) in the manual mode. [0.25]

The RHR Letdown Control Valve (0.25] is placed . in manual and set _to full open. [0.25]

The Letdown Pressure control valve (0.25] is placed in auto. (0.25)

(Any 2/3, 0.5 es.1 (1.0)

b. 1. Increasing. letdown flow rate.

2 .- Marked decrease in pressurizer heatup rate.

3. Stable or increasing pressure.
4. Decreasing level. (any 2, 0.25 es.] (0.5)
c. Less than 350 psig may cause damage to RCP seals, (0.25] due to insufficient d/p across the seals. (0,5)
d. 2, 5, 4, 3, 6, 1 (Subtract 0.2 points for each perturbation necessary to achieve proper order.] (1.0)

REFERENCE IPO-001A, Pp. 18-20 & SOP-108A, P.5 & IPO-002A 002000G013 001000G013 ...(KA'S) l

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

l l

Z4__EBQQEQUBES_:_NDBdeLt_eBNQBd6Lt_EdEBQENQY_6NQ .PAGE 23 86DIQLQQIQaL_QQUIBQL QUESTION '7.051 (3.00).

Wh'ile f unctioning as Shif t Supervisor, a condition ' arises which requires ontry into containment while critical at 30% power.- The operator entering.

will receive an estimated whole body dose of 40 mrem. The following data le available:

Operator 1 2 3 4 LSox Male Female Male Female Age 27 24 38 20-Wk/ Exposure 35 mrem 0 mrem 280 mrem 30 mrem Qtr/ Exposure 1230 mrem 465 mrem 970 mrem .1120 mrem Life Exposure 5200 mrem 54730 mrem 9970 mrem Romarks History 4 months Form-4 Form 4 Unavailable pregnant on file on file Each operator is technically competent and physically capable of porforming the task. Emergency limits do not apply and Rad-Chem will not cpprove extensions. These are the only operators available and at least one of them MUST be chosen for the job.

Give your reasons for accepting or rej ecting EACH operator. (3.0) r I

l l

l I

Zi__EBQQEQUBES_=_NQBd6Lt_6BNQBd6Lt_EME80ENQX_6NQ PAGE' 25 B6DIQLQQ106L_CQNIBQL ANSWER 7.05 (3.00)

Operator - #1 Rej ected [.25] since he has no history -on file and will exceed 1.25_ REM /QTR whole body exposure [.50]

Operator #2 Rej ected C.25] since she has already exceeded whole body exposure limit for the term of her pregnancy [.50]

Operator #3 Rejected [.25] since he will exceed the 300 mrem. weekly limit ,

without Red-Chem approval [.50]

Operator #4 Accepted C.25] since she will not exceed 300 mrem / week admin.

limit [.25]-and will not exceed 1.25 REM /QTR whole body C.25]

(SCn-18) criteria does not apply since the operator will not exceed the

-allowable quarterly limit of 1.25 REM /QTR whole body.) C3.0)

REFERENCE OP RAD. WKR. TRNG. Pp. 9-11 10 CFR 20.101 USNRC Regulatory Guide 8.13 194001K104 194001K103 ...CKA'S)

QUESTION 7.06 (2.00)

For each of the following cases, state whether or not emergency boration is required:

e. One control rod fails to insert on a reactor trip
b. Reactor is in mode 6 with boron concentration of 1900
c. Reactor power is 100% with control bank D at 58 steps
d. RCS temperature is stable at 551 degrees F following a trip when a steam generator becomes faulted.
o. Reactor is in mode 3 with Keff of .98
f. During a startup, the reactor goes critical with the "Rod Insertion Low limit" alarm lit. (2.0) l l

{

b_

Zi__EBQQEQUBES_ _NQBd8Lt_8BNQBd8Lt_EMEBQENQ1_8NQ PAGE 26

'88010LQQ108L_GQUIBQL 1 ANSWER 7.06 (2.00)

a. NO
b. Yes
c. Yes
d. No
o. No.
f. . No (0.333 es.) ~C2.0) <

REFERENCE CP TS & ERG,s 000024K302 000024K301 000024A205 ...(KA'S)

QUESTION 7.07 (1.50)

a. . When Centry conditions) and why (purpose) is EOS-0.0, "Rediagnosis,"

used? (1.0)

b. What condition (s) must be met to enter EOS-0.1, "Reactor Trip Response," instead of E0P-0.0, "Reactor Trip or Saf ety Inj ection,"

when responding to a reactor trip? (0.5)

ANSWER 7.07 (1.50)

a. EOS-0.0.is used when, based on operator judgement (0.5), it is neces-sary to determine or confirm the most appropriate post accident

- recovery procedure. [0.5] (1.0)

b. EOS-0.1 is used when, based on operator judgement, SI is neither actuated nor required. (0.5)

! REFERENCE CP EOS-0.0, Re d i ag nos is , p. 2 -

CP EOS-0,1, Reactor Trip Response, p. 2 000007G011 ...(KA'S)

, t l

l

(***** CATEGORY 07 CONTINUE 0 ON NEXT PAGE *****) [

! Zi__2BQCEDUBES_=_NQBd8Li_8HNQBd6Li_EMEBGENQX_eNQ PAGE 27

-B&QIQLQQIC8L_CQNIBQL.

QUESTION 7.08' (1.50)

You are given the following summary ofiCritical Safety Function Status Trees while. performing-EOP-2,."Faulted Steam Generator Isolation." If ALL procedures were to-be used, what order would they.be used in?

.1. Heat Sink Red

2. Core Cooling Yellow
3. Suboriticality Orange
4. . Inventory Yellow
5. Containment Red
6. Integrity Orange (1.5)

ANSWER 7.08- (1.50) 1,-5, _

3, 6, 2, 4 (Deduct 0.3 for es. pertubation req'd to establish corr. order) (1.5)

REFERENCE ERG Executive Volume, Users Guide pp 10-15 000074G012 000074G011 000069G012 000069G011- 000054G012 000054G011 000029G012 000029G011 000009G012 000009G011

...(KA'S)

QUESTION 7.09 (1.50)

While performing E0P-3.0 "Steam Generator Tube Rupture," conditions are such that you decide to use the "adverse containment" values for L porformance of the E0P. State ALL conditions that must be satisfied bofore you can AGAIN use the "normal containment" values in the E0P.

i- State any assumptions you make.

~ Zi__EBQQEQUBES_=_NQBdekt_6BNQBd6Lt'_EdEBQENQ1.8NQ PAGE 28 88010LQQ106L_QQNIBQL-ANSWER' 7.09 (1.50)

1. Containment pressure less than 5'psig (0.5) AND, (0.5)
2. Containment radiation level remained less than 1ES R/hr. (0.5)

Note: Will accept for Item 2: Containment rad. levc1 less than 1ES R/hr.

and integrated dose less than 1E6 R.

Will also accept a stated assumption that containment rad levels never exceeded 1ES R/hr. (1.5)

REFERENCE ERG various fold out pages ERG Executive Volume, Generic Instrumentation, pg. 21 103000G015 000038G012 000038G011 ...(KA'S)

QUESTION .7.10 (2.50)

a. After a reactor trip with stable conditions attained, what are the conditions and logic that would cause the operator to actuate safety I

inj ect ion? (0.75)

b. After a reactor trip and safety injection, what conditions and logic require the operator to stop the reactor coolant pumps? (0.75)
c. While performing E0P-1.0, "Loss of Reactor or Secondary Coolant", you enter a step where the Action / Expected Response is not satisfied (left hand column). You enter the Response Not Obtained (right hand column)' tor that step and find that you carnot complete the actions as stated. What is your next action? (1.0) t

Z1__EB99EQUBES_:_N9806Lt_oHN9Bdokt_EUEBQENQ1_6NQ PAGE 29 86019L90196L_99 NIB 9L ANSWER 7.10 (2.50)

c. Subcooling < 15 oF. [0.25] OR; (0.25] Pzr. level cannot be maintained above 20%. [0.25] (0.75)
b. At least 1 CCP or SI pump running, (0.25] AND; (0.25] RCS subcooling less than 15 oF. (0.25] (0.75)
c. (If further contingency actions are provided, perform the next contingency action.) If further contingency actions are not provided, perform the next step or substep in the left hand column. (1.0)

REFERENCE CP E0P-0.0 Pp. 17 ERG Executive Volume, Users Guide. P. 5 000011012 000009G012 000009A201 ...(KA'S)

QUESTION 7.11 (2.50)

c. Immediate action in "Reactor Trip or Safety Inj ec t ion" (EOP-0.0) requires the operator to verify feedwater isolation. State a primary and alternate method to confirm proper isolation. (0.5)
b. What is the earliest step where the operator may exit E0P-0.0 and enter EOS-0.1? (0.5)
c. When the operator is instructed to "Check for faulted S/G", what indication is observed? (0.5)
d. What are the 2 parameters and threshold values that determine if main steam lines should isolated? (0,5)
o. According to E0P-0.0 step 15,"Check if RCS is intact", What are 2 of the parameter indications the operator observes to make this determination? (0.5) t l

l

Zt__EBQQEQWBES_ _NQBd6Lt_eBNQBd8Lt_EMEBQENQ1_6NQ PAGE' 30 88DIQLQQIQ6L_CQNIBQL ANSWER 7.11 = (2,50).

a. 1. Check monitor' light boxes _ ,

-2. check individual-valve / component indication.(0.25 ea). (0.5)

b. Accept either; Step 4, OR Check if SI is actuated / required. (0.5)
c. S/G pressure (0.5)
d. Containment pressure greater than 6.0 psig (0.25) ,or Steamline pressure less than 600 psig. [0.25] C0.5)
o. 1. Cont. rad.
2. Cont. press.
3. Sump level (any 2, 0.25 ea) (0.5)

REFERENCE CP E0P-0.0, Pp. 5-10 000009G012 ...(KA'S)

(***** END OF CATEGORY 07 *****) l l

l

o

' Az__8DdINISIB6IIVE_EBQQEQUBEft_00NDIIIONSt_6HQ_LIMII6IIQNS -PAGE 31 QUESTION. 8.01 (1.50)

Technical-Specification 3.1.1.1 specifies a minimum shutdown margin that must be maintained for modes 1 through 4. This shutdown margin requirement was. determined after analyzing the most restrictive conditions.under which reactivity transients occur throughout core life. Per the Technical Specification-Bases, what.are these most restrictive conditions? Specify the time in core. life, the value of RCS Temperature, and the accident that was.anclyzed. (1.5)

ANSWER 8.01 (1.50)

1. Time in core life-EOL
2. Tavg-At no load operating temperature
3. Accident-Steam line break and resulting RCS cooldown (0.5 es.) (1.5)

REFERENCE CP TS Bases 3/4.1.1.1 4

001000G006 ...(KA'S)

QUESTION 8.02 (2.50)

True or False

o. The Technical Specifications consider Pressure Isolation Valve Leakage to be a portion of the allowable identified leakage,
b. .The pressure isolation valve leakage limit is specified at an RCS pressure of 2235. Thus, a lower value is limiting at a lower pressure in mode 3 or 4.
c. All listed (in table 3.4-1) Reactor Coolant System pressure isolation valves must undergo leak rate verification surveillance after manual or auto operation.
d. The Containment Air Cooler Condensate Flow Rate and the Containment Atmosphere Gaseous Activity monitoring systems constitute 2 of the 3 normally required leakage detection systems. (2.0)
o. Measurement of CONTROLLED LEAKAGE must be performed at a specific RCS

- pressure. (2.5)

Az__8DdINISIB6IIVE_EBQQEQUBESt_QQNDIIIQNit_8NQ_LIMII6IIQBS PAGE 32 ANSWER 8.02 (2.50)

e. True
b. False'
c. True
d. -False
e. True (0.5 ea.) (2.5)

REFERENCE CP TS 83/4.4.6, 3 .' 4 . 4 . 6 , 3/4.4.20 3 -002000G011 002020K401 ...(KA'S)

- QUESTION 8.03 (2.00)

a. During operation in mode 1 with the Positive Displacement Charging Pump undergoing repair, the "A" Centrifugal Charging Pump (CCP) fails due to a power supply breaker that burns up. State 2 Technical Specification requirements that will be affected and force entry into LCO's. (1.0)
b. What is the basis for requiring only 1 CCP to be operable with the plant in mode 47 (1.0)

ANSWER 8.03 (2.00)

c. 1. Two independent ECCS subsystems shall be operable. (0.5)
2. At least two CCP's shall be operable. (0.5)
b. To assure that a mass addition pressure transient (0.5) can be relieved by 1 PORV OR, 1RHR Suction Relief. [ Accept either valve for full credit of 0.5 pts.) (1.0)

REFERENCE CP'TS 3.1.2.4, 8.3.1.2, 3.5.2 006050K506 002020G006 006000K504 ...(KA'S) 1

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

4

s, at__eDdINISIB6IIVE_EBQCEDUBESt_CQNDIIIQNSt_6ND_LIMII8IIQNS PAGE 33 iQUESTION 8.04 (1.50)

During the backshift as Shift Supervisor with the plant at_ low power, you

-ere reviewing logs and paperwork of a special test that was performed yesterday during plant startup. During this review you determine beyond a doubt that~a Safety Limit was violated during the startup/ test, and no one else is aware of this. What are 2 actions you must carry out as soon as possible and why? (1.5)

ANSWER 8.04 (1.50)

1. Commence immediate shutdown. (Accept Trip)
2. Make notifications (NRC Ops. center, Management, ETC.) (0.5 es.)

Operation of a unit that has exceeded a safety limit must be approved by the commiss ion. CONCEPT (0.5] (1.5)

REFERENCE CP TS 6.7.1 012000G005 ...(KA'S)

QUESTION 8.05 (2.00)

Any individual that is permitted to enter a high radiation area and is exempt from issuance of an RWP must satisf y ONE of THREE technical specification requirements concerning radiation monitoring. What are these THREE requirements? (2.0)

At__eQUINISIB611YE_E80CEDUBESt_GQNDIIIONSt_6HD_LIMII6IIQNg PAGE 34 ANSWER 8.05 (2.00)

Individuals permitted to enter shall be provided with or accompanied by one or more of the following:

1. Radiation monitoring device which continuously indicates the

' radiation dose rate in the area. [0.5)

2. Radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received [0.5).

3.- An-individual qualified in rad. protection procedures,[0.51 with a dose rate monitoring device. (0.5] (2.0)

REFERENCE CP T.S; 6.12.1 194001K105 ...(KA'S)

QUESTION 8.06 (3.00)

In accordance with Operations Department Work Instruction OWI-101, Operations Department Procedure Control.

Q. State the 5 locations where "CONTROLLED" Operations Department procedures are available. (1.0)

b. What are 2 other designations for procedures that are not designated as "CONTROLLED" procedures? Describe the use of or limitation of each ,

type. (1.0)

c. Who is responsible for making revisions or inserting changes into controlled procedures? (0.5)
d. What is the only valid reason for removing CONTROLLEO procedures from I their normal work location? (0.5)

,I 1

Ba__eDMINISIB6IIVE_EB9CEDUBES&_CQNDIIIONSt_eND_LIMII8IIQNS PAGE 35 ANSWER. 8.06 (3.00)

e. 1. RRO work area i
2. RO work area
3. SS office
4. .AO's office
5. Remote SD panel (0.2 es.) (1.0)
b. 1. WORKING COPY. (0.25] Can be used in the field to perform plant operations. (0.25)
2. INFO ONLY. (0.25] Good only for training, etc. Cannot be used to perform plant operations. (0.25] CONCEPT. (1.0)
c. Accept either; Op's clerk or SS (0.5)
d. To make copies. (0.5)

REFERENCE CP OWI-101, Pp2,3 194001A101 ...(KA'S)

QUESTION 8.07 (2.00)

In accordance with The Station Administrative Manual, list 2 administrative mothods available for management to disseminate information to the CPSES etaff. Briefly describe the specific function of each method that is I listed. (2.0) e ANSWER 8.07 (2.00)

1. SPECIAL ORDER: (0,5) A management directive with short term applicability for the purpose of providing policy on safety related matters important to plant operation. [0.5) CONCEPT

, 2. NIGHT ORDER: (0.5) A directive to personnel on back shift concerning impending work, occurrences, or conditions. [0.5) CONCEPT

3. MANAGEMENT MEMORANDA: (0.5) A directive on conduct of non-safety related business. [0.5)

! (Any 2/3, 1.0 ea.] (2.0)

L.

- Bi__6DDINISIB6IIVE_EBQCEDUBESt_CONDIIIQNSt_6NQ_LIMII6IIQNS PAGE 36-

-REFERENCE ,

CP STA-207, P.2

'194001A103 ...(KA'S) ,

-QUESTION 8.08 (3.00)

Concerning STA-606, "Work Requests and Work Orders", answer ~the following:

c. What 3 conditions may preclude the requirement to attach a WORK REQUEST TAG to the equipment being worked on? (0.75)
b. During the performance of a WORK ORDER specifying that only trouble shooting is to be done, the problem is discovered to be a simple 2 minute repair job. How should this situation be handled administratively? (0.75)
c. True or False
1. Completion of a work order signifies that the affected equipment is now operable.
2. Preventive Maintenance work packages do not normally require a prework review by QC.
3. Work may never start without an approved work order. (1.5)

ANSWER 8.08 (3.00)

c. Do not hang if:
1. On a control board.
2. Interferes with component operation.
3. Increases personnel exposure. (0.25 ea.) (0.75)
b. The work order must be revised bef ore perf orming additional corrective action. (0.75)
c. 1. False
2. True
3. False (0.5 ea.) (1.5)

REFERENCE CP STA-606 Pp. 3-12 002000G001 ...(KA'S)

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at__eDdINISIB6IIVE_EBOCEQ9BESt_CQUDIIIQUSt_eUD_LIMII6IIQUS PAGE 37 QUESTION 8.09 (3.00)

c. For what reason may a Shift Supervisor waive the requirement for independent verification of a condition required by a clearance / safety tag order? (0.5)
b. What is the procedure to be followed in the event a clearance must be released and the Individual Responsible For Clearance is not on site or available by telephone? Explain how action is documented and any followup action required. (1.0)
c. Who is responsible to sign for a review of a clearance request to ensure unit availability and sufficient boundary protection? C0.5)
d. Besides boundary valves what additional tagging requirement is there for work on a fluid system pressure vessel? (0.5)
o. Why may there be multiple tags on components that isolate a single piece of equipment undergoing repair? (0.5)

ANSWER 8.09 (3.00)

c. May be waived to avoid significant radiation exposure. (0.5)
b. The individual's supervisor may approve release, (0.5] which will be documented in Restoration Authorization. [0.25] The individual must sign for release upon returning to the site. (0.25] (1.0)
c. Assistant Shift Supervisor. (0.5)
d. A vent or drain valve should be tagged open. (Preclude pressure buildup.) (0.5)
o. A separate clearance is required for individual work groups. (0.5)

REFERENCE CP STA-605, Pp. 3-12 19400lK102 ...(KA'S)

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at__6DdINISIB6IIVE_EBQQEDVBESt_QQUDIIION2t_6ND_LIdIIoIIQU$ PAGE 38 QUESTION 8.10 (2.00)

e. Describe the complete method of providing initial and follow-up messages to the state and local officials after classifying and declaring an emergency. State the specific equipment used. (0.75)
b. When is notification / communication respons!bil.ty transferred from the control room to the Technical Support Center? C0.5)
c. Who (title, pos it ion) is r es p o ns ib le for prioritizing messages intended for off-site distribution when communication / notification responsibility is in the:
1. Control Room?
2. Technical Support Center?
3. Emergency Off-Site Facility? (0.75)

ANSWER 8.10 (2.00)

c. Messages shall be announced by voice communication on the "ringdown" circuit, (0.41 and then transmitted by telecopier. (0.35) (0.75)
b. When the Emergency Coordinator's responsibilities transfer. (0.5)
c. 1. Emergency Coordinator (accept SS)
2. TSC advisor
3. Communications Coordinator (0.25 ea.] (0.75)

REFERENCE CP EPP-203, Pp. 2, 5, 6, 8 l 194001All6 ...(KA'S) l l

l l

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l

Hi__eQUINISIB8IIVE_EBQQEQUBESt_QQNDIIIONSt_6NQ_LIMII6IIQNS PAGE 39 QUESTION 8.11 (2.50)

During an emergency as Emergency Coordinator.

a. What authority do you have for imposing protective' response actions beyond the-limits of the site boundary? Fully explain. (0.75)
b. What is basis for recommended actions that you will piavide within a given-exposure pathway plume? (0.5)
c. After' declaring a Site Emergency and before the TSC and EOF are augmented, who (title) can provide release rate /proj ected dose calculations, and how Cequipment) is this done? (0.5)
d. What 3 specific actions must be completed to formally transfer the authority of the Emergency Coordinator to the TSC Manager? (0.75)

ANSWER 8.11 (2.50)

a. None. [0.25] The local officials have final decision making authority and are responsible for implementing protective actions. (0.5) (0.75)
b. The proj ected dose (to the whole body and thyroid.) (0.5)
c. The STA (0.25] will use the Micro-computer. (0.25] (0.5)
d. 1 Brief the successor (regarding current plant status and status of response activities.)
2. Notify offsite response organizations.
3. Document the turnover (on the Response and Recovery Activity Record Sheet.) (0.25 es.) (0.75)

! REFERENCE

. CP EPP-304, Pp. 4, 9,

! CP EPP-ll2, P. 3 CP EPP-109, P. 7 194001All6 ...(KA*S) l l

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