ML20137L764
| ML20137L764 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 01/22/1986 |
| From: | Mcmillen J, Picker B, Reidinger T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20137L757 | List: |
| References | |
| 50-305-0L-85-01, 50-305-L-85-1, NUDOCS 8601280080 | |
| Download: ML20137L764 (31) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-305/0L-85-01 Docket No. 50-305 License No. DPR-43 Licensee: Wisconsin Public Service Corporation Mr. D. C. Hintz Manager - Nuclear Power P.O. Box 19002 Green Bay, WI 54307-9002 Facility Name: Kewaunee Examination Administered At: Kewaunee Nuclear Power Plant Examination Conducted: Week of December 16, 1985 Examiners: icki e Date h
E l 0B. Picker lN b.
(Tate /
Approved By: .M 1 n perating Licensing Section
//M I ITate/
Examination Summary l
l Examination administered on week of December 16, 1985 (Report No. 50-305/0L-85-01)
,. Examinations were administered to four senior operator candidates.
Results: All candidates passed the written and the oral / operating examination.
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REPORT DETAILS
- 1. Examiners T. D. Reidinger - NRC Region III B. Picker - EG&G Idaho
- 2. Examination Review Meeting Copy of the examination and answer key was given to the facility personnel for review at the conclusion of the written examination.
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Facility personnel mailed their comments to the Chief Examiner after their review. Their comments and resolution of these comments are Attachment 1 to this report.
- 3. Exit Meeting On December 20, 1985 an exit meeting was held the following personnel were present at this meeting:
F. Stanaszak, Facility Training Supervisor R. Zube. Facility Simulator Supervisor J. Brown, Facility Training Instructor K. Kwiatkowski, Westinghouse Site Service Manager K. Evers, Assistant Manager, Plant Operations D. Braun, Operation Supervisor C. Giesler, Corporate, Vice President Power Production The facility was informed of those candidates who had clearly passed the oral / operating examination. The Chief Examiner also made known to the facility his concerns relating to a potential violation of CFR 50.54 (k)
. . . operator or senior operator licensed pursuant to Part 55 shall be present at the controls of the facility and 50.54(iii) . . . when a nuclear power unit is in operational mode other than cold shutdown . . .
each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. The facility standing orders dated June 12, 1985 define the area designated as the reactor operator watch station but it allows both operators to station themselves in the back of the control room with~no precautionary measures behind a series of seven feet to nine feet high storage cabinets thus isolating and restricting their line of sight of essentially the whole control room and associated readings and parameters. The Senior Resident Inspector will close out this item.
Another concern mentioned to the utility is the consistent violation of their Security Administrative control directive relating the Shift Supervisor's key cabinet. The directive specifies that the key locker shall be closed and locked. But on every occasion the examiner visited the plant for examinations the key locker was either found in one of the following conditions; doors left ajar with the security key inserted in the lock and no Shift Supervisor in the SS office, doors left ajar with a i i
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Shift Supervisor present, doors closed but not' locked with or without a Shift Supervisor present. . Recommendation to the utility is either to follow the directive or change the directive in the direction of Kewaunee's current operating practices. The utility indicated they will look into the matter.
Attachment:
Facility Comments - Senior Reactor Operator Examination and NRC Resolution of Facility Coments 7
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ATTACHMENT 1 QUESTION NUMBER 5.6 COMMENTS: Answer Key should be corrected to read:
(82% inserted - bank C)
(24% inserted - bank B)
Also; Reactor startup procedure precaution 2.9 reads:
" Rod insertion limits must be observed as defined in Technical Specifications 3.10.d and figure TS3.10-3."
EXAMINER'S COMMENT: The examiner will accept the revised answer. The examiner also notes that the alternative answer is in direct violation of the current Kewaunee Nuclear Plant insertion limits as established by the Kewaunee reactor data book graph RD 4.1.2 where limits are (39 - bank C), (169 - bank B.)
QUESTION NUMBER 5.10 COMMENT: Due to terminology used in question (coefficients of the power defect) - answer could be " doppler power coefficient."
EXAMINER'S COMMENT: Theoretically, the doppler coefficient is expressed as ff~(PCM/F)whereasthealternativeanswerofdoppler power coefficient would be PCM/% power a subtle' difference none the~less. The examiner will accept the alternative answer with the stated reservations.
QUESTION NUMBER 5.11 COMMENT: Answer Key, item 1-Based on how we train our people, the term " trip reactivity" may not be used. Instead, the term used for this item may be " shutdown margin."
RECOMMENDATION: Accept " shutdown margin" for item 1.
EXAMINER'S COMMENT: Shutdown margin is an acceptable equivalent alternative answer. The examiner notes this term is extracted verbatim from Technical Specifications.
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QUESTION NUMBER 5.12 COMMENTS: 1. Answer should indicate-that BOTH production and removal terms are flux dependent, therefore there is no change in equilibrium concentration due to the flux terms cancelling each other.
- 2. Possible alternative to paragraph answer is an equation.
EXAMINER'S COMMENT: 1. The examiner notes that the alternative answer provided is an equivalent acceptable response. The examiner notes also the possibilities that the answer key was misread or the question was technically misunderstood during facility review.
- 2. No comment on the equation as the facility did not provide the equation. No change to answer key.
QUESTION NUMBER 5.13 COMMENT: Answer should discuss the change in " thermal utilization factor" with an increase in boron . . . etc.
EXAMINER'S COMMENT: The alternative answer is unacceptable due to the fact that it would allow two answers. One being correct if the candidates specifically state that the thermal utilization factor decreases as boron concentration increases and the negative moderator temperature coeffic'ient will decrease to a positive coefficient then it will be acceptable.
QUESTION NUMBER 5.15 COMMENT: Answer Key reference' change needed.
EXAMINER'S COMMENT: The additional reference is redundant to the listed references but will be added to answer key to provide more elucidation to the facility training staff.
QUESTION NUMBER 5.18 COMMENTS: 1. Due to Kewaunee end of cycle characteristics it is possible to answer " Divergent oscillations could result."
- 2. The answer key splits credit between two phrases that mean the same thing - should not need both phrases in answer for full credit.
Westinghouse reactor core control for large pressurized water reactors, Chapter 4, pages 4-28 and 4-29.
- 1. Accept " Divergent oscillations".
- 2. Allow full credit for either phrase of answer.
EXAMINER'S COMMENT: The examiner notes that the referenced text was not provided to the examiner. The alternative No. 1 answer is not acceptable; the alternative No. 2 answer is also not acceptable as the two' phases are entirely distinct from each other i.e., phases do not purport the same idea.
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QUESTION NUMBER 6.01 COMMENT: Fer Answer Key Request:
Low Pressure Reactor Trip OPAT Trip OTAT Trip S/G Low Low Level RCS Low Flow Power Range NI High Flux Rate Power Range NI High Range High Flux EXAMINER'S COMMENT: Will accept the requested answers, as the reference material was not available to examiner.
QUESTION NUMBER 6.07 COMMENT: Answer Key item 3 should include feed regulating bypass control valves.
EXAMINER'S COMMENT: Will include and amend answer key.
QUESTION NUMBER 6.09 COMMENT: Answer Key typo:
R-18 should be R-19 EXAMINER'S COMMENT: Will correct typographical error.
QUESTION NUMBER 6.12 COMMENT: Question is open to interpretation as to what is
" operational reasons."
EXAMINER'S COMMENT: Will allow flexibility on interpretation, as it is an operational precaution in the operating procedures.
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QUESTION NUMBER 7.8
- COMMENT
- - Answer Key part (b):
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. Should accept heating up water in Reactor or in hot ~~~
S/G tubes.
EXAMINER'S COMMENT: Will accept additional answer.
QUESTION NUMBER 7.9 i
COMMENTS: 1. Question is open to interpretation as to what'does.
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. . . addressed specifically by Technical i Specifications" mean.
- 2. Depending on how Comment No. 1 is interpreted
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"i) t DNB parameters" should be "YES".
) EXAMINER'S COMMENT: All candidates were addressed with clarifications during the examination. All candidates indicated they did fully understand the question. No other answers will be acceptable as the question is quite specific addressing
, the Technical Specifications.
QUESTION NUMBER 7.10 COMMENT: In accordance with ES-0.3, the Quick Reference Foldout page for E-0 is to be followed. This page tells the Operator to Actuate SI of in' Adverse Containment Conditions and Pressurizer Level cannot be maintained greater than 20%. Therefore, Part b) of the-question could be answered with this in mind.
EXAMINER'S COMMENT: The examiner will keep it in mind.
- - QUESTION NUMBER 7.11 i
- COMMENTS: The intent of this question is not affected by the following:
~1 . There is no orange path for Integrity at Kewaunee.
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- 2. There is no red path for Inventory at Kewaunee.
(Proctor. changed during' exam.)
, EXAMINER'S COMMENT: The supporting reference was not provided to the examiner prior to the examination. The answer key will be amended to reflect the comments.
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QUESTION NUMBER 7.13 COMMENT: Question is worded such that it could be assumed.the plant has had a Reactor Trip with no S.I. Therefore, answers could be in accordance with ES-0.1, Reactor Trip Response, Step 4.2.c contingency.
EXAMINER'S COMMENT: Alternative answer is not acceptable as it is a generic application or answer to a specific solution as called
-for by the abnormal feedwater procedure. Answer key will not be amended.
QUESTION NUMBER 7.16 COMMENT: Answer Key has "272,500 gallons - 98% level". T.S.
reference (TS3.3-2) only refers to 272,500 gallons for level, not 98%.
EXAMINER'S COMMENT: The answer key will not'be amended; 100% capacity of the RWST is 276,500 gallons as per system description. If the candidate wishes to by using an algebraic formula to convert 272,500 gallons to 98%, it is acceptable and correct per Technical Specification's.
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QUESTION NUMBER 8.3, 8.5, 8.14 COMMENT: Note attached sheet of recent title changes at Kewaunee.
EXAMINER'S COMMENT: Answers were amended to reflect the title changes of personnel at Kewaunee only. The questions and answers are essentially the same.
QUESTION NUMBER 8.4 COMMENT: Question wording (i.e., " Declared emergency") could solicit Emergency Plant notification.
EXAMINER'S COMMENT: There is no question concerning wording " declared emergency" in the administrative control directive. It is extracted verbatim from this A.C.D. The alternative answer would abrogate the Shift Supervisor's responsibilities as called for per the ACD. The facility's answer is not acceptable.
QUESTION NUMBER 8.10 COMMENTS: Kewaunee Abnormal Operating Procedure A-MI-87 specifies one hour. Even though question wording refers to Standing Orders, conservatism may lead to A-MI-87 time requirement as response.
EXAMINER'S COMMENT: Standing Orders specifically states the four hour requirement for tripping bistables for instrumentation which will be taken out of service. Standing Orders also state it is effective immediately that date. Will accept the alternative conservative answer of one hour per the abnormal. procedure.
QUESTION NUMBER 8.11 COMMENi: Question could solicit directing operators to perform increased logging of rod position by core instrumentation as stated in the Standing Order referenced Technical Specification.
EXAMINER'S COMMENT: The logging of rod positions is only required if nuclear power is greater than 50%. If the candidates specifically state logging of rod position is required when reactor power is greater than 50% then it will be acceptable.
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QUESTION NUMBER 8.15 COMMENTS: Answer Key Part a) should read:
"The secondary coolant activity is based on a postulated release of contents of one steam generator to the atmosphere."
The " steam break accident combined with failure of a steam line isolation valve" is only given as an example in Technical Specifications.
EXAMINER'S COMMENT: Will accept the alternative question in addition to the original answer which allows an interpretation of the Technical Specification bases as stated. The original answer will also permit flexibility as opposed to the alternative answer. The candidate can express the Technical Specification by illustrating the bases.
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. U.S. NUCLEAR REGULATORY COMMISSION l SENIOR REACTOR OPERATOR LICENSE EXAMINAT' ION i FACILITY: Kewaunee
! REACTOR TYPE: ~ h_E("M,_0{s~e--~
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l DATE ADMINISTERED : _ _Dec embe r_1J_,__198_5_ __
i EXAMINER :~ _J.,,_D._Jeidinget__ _
j APPLICANT: _______,___,;_______
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! IN3TRUCTIONS TO APPLICANT
- Use separate paper for the answers.Weite answers on o'ne side on1y. Staple question sheet on top of the answer sheets. Points for each question are i
indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least-80%.
% OF .
4 CATEGORY % OF APPLICANT'S CATEGORY j VALUE TOTAL SCORE VALUE ________________ _ _ _ _ _ _ _ _ _ , _ _ , _ _ _
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-__ _ _2)_ _ _ _ _ ,,_,_ _ _ _ _ _____,.,_ 5. Theory of Nuclear. Power Plant Op er at_i on , Fl ui ds , and Thermodynarnics
_ _2_5_ _ _ _,,2_5___ __ ,_______ ,_______ 6. Plant systems Design, Control, and Instrumention
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,_ I_ _. _ _2],,,, _ _ _ _ _ _ _ _ _ _ _ _______ 7.1racedures-Normal, Abnormal, Emergency, and Radiological Control
_ _2,5, _ _, _ _ _ _2]_ ,, ___,____
_____,,, B. Administrative Procedures,
,_ Conditions, and Limitations A ' [_ _ ______ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ TOTALS
- Final G r a d e _ _ ,.,_ _ _ ,_ _ %
i j al1 wark done on this exam is my own, I have neither_given nor' received' aid i
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i _____________________________
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_ Applicant's Signature i
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Category 5-Theory of Nuclear Power Plant Operation, Fluids,And
. Thermodynamics 5.1 Indicate and explain if you agree or disagree with the following statements:
a)Within 10 minutes after a reactor trip, prompt neutrons production within the fuel is non-existent. (1.0) b) thermal neutron flux levels within the fuel pellets are higher than those in the coolant channel between fuel rods.
(1.0) 5.2 With RCS tave equal to 290 degrees F, what is the maximum allowable RCS pressure, in psig, which does not violate the 320 degrees differential temperature limitation between the RCS and the pressurizer? (1.25) 5.3 When the flow rate through a centrifugal pump increases, available NPSH (increases or decreases) and required NPSH (increases or decreases). . choose correct responses (1.0) 5.4 During 100% power operati on, the Departure from Nucleate
' Boiling Ratio (DNBR) is greater than the DNBR for 20% reactor power. true/ false (.75) 5.5 Which of the following best describes the effect on MTC if the RCS temperature is lower ed? (1.0) a)It becomes 1ess negative; baron and water molecules are swept into the core as a result of the outsurge from the pressuriner, so neutrons spend more time in the resonance region.
b)lt becomes less negati ve; the rate of change in the density of water per degree F is less at lower temperatures causing a teduction in the rate of change in the resonance escape probability.
c)It becomes more negative; thermal utilization increases and resonance escape probability decreases.
d)It becomes more negative; as temperature is lowered'the moderator begames more dense.This increases the amount of water molecLies in the core,so neutrons have a greater probability of colliding with a water molecule which increases the negative reactivity effect.
5.~ 6 How is adequate shutdown -.aargin verified during a reactor startup prior to cri ticali ty? (1.0)
U.7 Which two factors cause the target axial flux difference at EOL to differ from that at BOL? (1.0) 1
5.8 Which of the following radioactive isotope'found in the reactor coolant would NOT indicate a leak through the fuel cladding? (1.0) a)I-131 c)CO-60 b)Xe-133 d)Kr-85 5.9 Overtemperature (delta T) reactor trips will provide core protection against DNB provided that two plant parameters are met. What are those two conditions or parameters.? (1.25) 5.10~ Explain which of the reactivity coefficients of the power defect act first to affect reactivity on a sudden power change due to rod movement and why? (1.5)
.5.11 List the three Technical Specification bases for_the control rod insertion limits. (1.5) 5.12 a) Explain how the increase in reactor power from 10% to 100%
effects equilibruim Samarium.
(1.0) b) Assume a reactor trip from 100% power with equilibruim xenon and samarium.
Would the reactivity ef'ect f of samarium concentration changes be more negative or less negative than the reactivity effett of xenon concentration changes from immediatel y af ter the trip until:
1)5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the trip (.50) 2)50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after the trip (.50) 5.13 Explain how an excessive baron concentration greater than 2000 ppm will result in a positive Moderator Temperature Cocfficient.
(2.0) 5.14 How does an increase in each of the following parameters affect DNBR? Consider each parameter separately.
a)RC5 temperature b)RCS pressure (.50)
(.50) c)RCS flow.
d) reactor power (.50)
(.50) 5.15 Describe the change in running current in "A" reactor coolant pump when "B" reactor coolant pump is turned on.
(.50) 5.16 What is the basis for establishing limits on axial flux difference? *
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5.17 As moderator temperature increases from cold shutdown to hot shutdown conditions, describe the effects'if any on the f oll owing. (consider each separately) a) differential baron worth (.50) b) control rod worth (.50) c) fuel temperature coefficient (.50) d) source range channel indications (.50) 5.18 What would most likely happen if the reactor operator took no 1
action other than to re-establish correct baron concentration when a xenon oscillation was caused by an inadvertant baron dilution? (1.25)
END OF CATEGORY 5 3
Category 6: Plant Systems Design, Control, and Instrumentation 6.01 Last year at Kewaunee a loss of power occurred during power operations on instrument bus IV and made up several partial reactor trip logics. List five partial reactor trip logics
'which were made up by the loss of'the instrument bus?
(1.5) 6.02 Last year at 100% power, a Kewaunee I&C technician inadvertently disconnected.the control and power cable from the.RM-14.The following actions occurred. (1.0)
Choose the correct answer a) radiation monitor trip signal was generated and auto started both A&B aux. bldg special . ventilation system trains.
b) radiation monitor trip signal was generated and auto started brain B aux. bldg special ventilation system train.
c) radiation monitor did not generate its trip signal, and no action resulted, i.e. no ventilation-system changes.
d) radiation trip signal was generated with a reactor trip signal and started A&B aux. bldg special ventilation system trains.
6.03 A reactor trip just occurred in hot shut.down mode; the RO was on manual control on the "A" steam generator maintaining level at 24%. and 1&C just reported that steam flow transmitter A was reading significantly higher than the feed flow transmitter A due to being out of calibration. Explain what caused the reactor trip ? (1.5) 6.04 With red pressurizer pressure channel tripped during the last refueling shutdown, a Kewaunee I&C technician was allowed to calibrated the white pressuriner channel to 2100 psig.
a) At what setpoint/s was the reactor protection system actuated? (.50) b) Nhat reactor protection signal was actuated? (.50) c) How was this signal generated? (.50) 6.05 Last year at full power, a Kewaunee SRO noticed that the boric acid tank selector switch was positioned to tank A and was notified that the D tank was physically aligned to pr ovide suction to the safety injection pumps.
Explain the. operation of the safety injection pumps during a S1 if a l ow l ow level is received from boric-acid tank B.
(include what valves will/will not open in this condition) 6.06 (1.5)
During the last refueling shutdown a Kewaunee mechanic cleared the low vacuum turbine trip at the turbine pedestal by momentarily resetting the turbine manual trip lever. Explain why both diesels started when he hit the trip lever to finish his test. (include setpoints) (1.5)
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.6,07 Explain how the feed flow system to the steam generators is
- effected by a Safety Injection during power operations? (1.5) 6.08 With no operator action on a safety injection signal, ihn reactor to the:
coolant pump seal water return will automatically go (choose one) (1.0) a) Hold up tanks b) Reactor coul ant drain tank c) Volume control tank d) pressurizer relief tank 6.09 a) List two radiation monitors which when actuated will
' automatically terminate any gaseous release from the Auxiliary Building vent. (1.0) b) List two radiation monitors which when actuated will automat i cal l y terminate a steam generator blowdown in progress.
(1.0) 6.10 Two solenoid operated valves (OPC 20-1/2) in the emergency trip fluid circuit tripped on impulse from the aux. speed channel due to turbine overspeed.
List all the valves just tripped by this action. (1.0) 6.11 a) Where does the component cooling surge tank relief valve discherge in the event that RCP A had a thermal barrier cooling coil rupture?
( 1. 0 )
b) List two sensors / instrumentation that are used to detect this RCP thermal barrier cooling coi1^ rupture in the component cooling wster system . -( 1. 5 )
c) List oncethe automatic action /s associated with these sensors actuated. ,
(1.0) 6.12 3
List two operational reasons why the huxiliary feed water pumps will be.placed in the " pull to lock" position by the SRO during NSSS heatup. ' -
(1.0) 6.13 What are two purposes of the main steam flow nor les on the main steam line?
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6.14 a)What are the two functions (purposes) of the Letdown Pressure Flow Control Valve (LD-10). (.50) b)If left in automatic control, what posi tion should (LD-10) be found in two minutes after a safety injection initiation? (.50) c ' ' 'h at is the failed position of LD-10,i.e. power / air loss?
(.50) 6.15 a) Describe how a Blackout signal when rer ived changes a Safety Injection loading sequence when the sequence is half completed. (1.0)
! b) Describe how a Safety Injection signal when received changes a Blackout signal loading sequence when i ts sequence is half completed. (1.0) 6.16 An interlock prevents opening motor operated valve RHR-300A/D (recirc. isol. valve) unless the Safety Injection pun.ps suctions (SI-5A/B) are s; hut.
What is the reason of this interlock? (1.0) 6.17 What AFW pump proble,m would be caused by back leakage of the main feed water into the AFW pump suction during power operations? <
(1.0) i i
END OF' CATEGORY 6 i
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Category 7-Procedures-Normal, Abnormal, Emergency and Radi ol og i cal Control 7.1 What two conditions specified in E-O foldout require tripping reactor coolant pumps? (1.0) 7.2 What two criteria constitute adverse containment conditions.
(1.25) 7.3 Why does the steam generator tube rupture procedure E-3 require that the 4eed flow be maintained to the ruptured S/G until level is greater than 2% narrow range?
7.4 (2.0)
Containment has high airborne gaseous nuclides after a LOCA, the Radiation protection group gives you a supplied air type respirator and a air supplied whole body suit.
For wnat nuclides does each equipment provide minimum protection against? (List one nuclide for each equipment) 7.5 (1.0)
The Operational Support Facility is activated for what Kewaunee emergency classification /s. (1.0) 7.6 What are the critical safety function criteria for the red path summery per ialdout?
(1.U) 7.7 A caution statement in SGTR with loss of reactor coolant tubcooled recovery procedure ECA-C.1 states that voi di nc may occur in the RC5 during RCS depressurization a)Why does thi s voiding occur? (.75) b)What RCS?
would be the i nd i c.a t i on / s that voiding occurred in the
(.75) 7.8 Procedure for reactor coolant pump operation has a caution which reads "A pressure spike may occur, due to temperature unbalance RCS solid."
in the RCS, when starting the first RCP with the a)What causes this temperature unbalance? ( 1. 5) b)What causes this pressure spike? (1.5) 7 4
7.9 Which of the following items are addressed specifically by Technical Specifications. Answer yes if the i tem i s or no if the item is not addressed in the Technical Speci f i cati on s.
(2.0) a) containment humidity b) boron dilution c) source range NI's during power operation d) steam dump system e) condensate storage tanl: chemistry f) rod drop time g) plant computer h) meteorological instrumentation i)DNB parameters j) steam generator PORV'S 7.10 During the performance of ES-0.3 Natural Circulation.Cooldown with steam void in Vessel:
allf a reactor coolant pump (RCP) _can be restarted, the pressurizer level should be greater than 65%.
Explain the reason for this level? (1.00)
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c)If a RCP cannot be resterted, the pressurizer level should be 20-25%.
What is the reason for this level? (1.00) i 7.11 During an accident recovery procedure, the operator who has been monitoring the Critical Safety Functions report the following.
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a)Subtriticality-red b) Core Coolin'g-orange c) Integrity-orange d) Heat Sink-red e) Containment-yel1ow f) Inventory yellow Rank the CSF's according to their importance in numerical order from the most to least important.
7.12 In kewaunee procedure " Post Accident Systems Operation" when i
sufficient core damage exists then specific permission must be obtained containment.
to bypass filters or re-rout vents and drains back to List whose approval /s must be obtained.(be specific) (1.5) 7.13 What immediate action /s per feedwater proceduro should an operator take to maintai n feed to the steam generators when all auniliary feedwater is lost and no SI signal ~is present?
(1.25) 8 5
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7.14 Last year at 100% power,a Kewaunee equipment operator inadvertently opened "B " fire pump breaker when the "A "
' pump'was tagged out. fire What general Technical Specification /s apply to fire pumps?
(list two)
(1.0) 7.15 A caution.in Plant Operations >15% procedur e requires that rod control should be in manual if control Bank "D" i s below 215 steps and reactor power is 90%.
What is the basis for this caution? (1.0) 7.16 Last year at Kewaunee an auxiliary operator i nadvertantly drained the Refueling Water Storage Tank (RWST) below Technical Specification minimum level.
a)What is that minimum l evel ? (1.0) b)What is the basis for the level in the RWST?
(list two)
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END OF CATEGORY 7 9
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Category 8-Administrative Procedures, Conditions,and Limitations 8.1 List three Shift Supervisor duties required per the ACD: when i he reviews a Temporary change request. (1.50) j 8.2 Who is the first person that the Shift Supervisor is responsible to contact per ACD if an incident occurs on the back shift.?
(1.0) 8.3 a)Who can terminate a Temporary Operating Procedure.?
(list three) (1.50) b>What is the ma>:imum t ime that a Temporary Operating Fracedure will be valid.'? (1.0) 8.4 List all the people that the Shift Supervisor is responsible to contact in a declare emergency when there is a fire _on backshift in a controlled area. (1.5)
- 8. 5 Match the following positions to their tagout control responsi bili ties. (answer s can be used more than once)
- 1) Shift Supervisor 2)Contral rcom super vi sor
- 3) Asst. Super intendent-Operations
- 4) Plant Manager a) grants permission to clear hold cards when the person listed on reached. hold card is not on site and can't be
(.50) b) maintains a list of personnel who ere authorized to install and remove cards.
(.50) c) authorizes the removal of equipment from service and its return to service.
(.50) d) authorizes the installation and removal of hold, caution and danger cards in the plant. (.50) e)maintained.
reviews tagout to ensure reactor safety is being
(.50) 8.6 The the Engineered Safety Features periodic checklist performs following functions: ( ~ answer- true or false for each)
- 1) verifies a system it operable
- 2) permits the roatine re alignment of any valve-found out(.25) of position
- 3) permits the routine removal of a lock on a valve to verify' (.25) the posi tion of the val ve
(.25) 4)misalignment verifies that a system is not operable due to a valve
(.25) 8.7 List two criteria that require a special radiation work request to be written.
(1.5) 10
8.8 What specific requirement /s is/are assigned to the Shift
- Supervisor per the ACD or ?rocedure in regards to the discharge radiation monitor when he has received 1)a liquid waste ' discharge permit.? (1.25) 2)a gaseous waste discharge permit.? (1.25) 8.9 Li st two people that the shift Supervisor is responsible to.
contact when a security alert is received during back shift.
(1.5)
'8.10 What is'the specific time requirement per Standing Orders for tripping bistables associated with instrumentation which will be taken out of service for long periods of time per Standing-Orders?
(1.0) 8.11 List two SRO responsibilities mentioned in Standing Orders, concerning rods which are aligned,and trippable, but indicate a electrical misalignment between IRPI and Group position i nd i cati on .
(1.0) 8.12 Technical Specifications require a minimum of ____ core thermocouples per quadrant be available for readout when the reactor is operated above ______%
power channel out of service.
with one excore nuclear (fill in blanks) (1.5) 8.13 What is the basis for the flow restriction orifice in the vent path from either the reactor vessel or the pressuriner?
(1.00) 8.14 With the reactor at power, is a RHR pump with its switch in the " pull to lock" position considered operable if an operator is assigned Superintendent.
to the switch by the Operations (1.00) 8.15 a)The Technical Specification requiring the secondary coolant activity of iodine-131 activity to be less than 1.0 Ci/cc is based on a specific accident. What is that accident?
b)The Technical Specification stipulating that the total(1.25)-
specific activity of the reactor coolant not exceed A=91/E uC/cc when the reactor is critical is based on a specific accident.
What is that specific accident?
(1.25) 8.16 Under what condition /s is/are a Pressurizer PORV considered inoperable pcr kewaunee Technical Specifications / Standing Orders / Procedures? (.75) 8.17 Prior to performing a safeguards surveillance procedure,.
the redundent train must be tested first. true/ false
(.75)
END OF CATEGORY 8 11
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MASiH CJPY Category 5-lheory of Nuclear Power Plant Operation, Fluids and Thermodynamics-Answers 5.1 a) disagree: prompt neutrons production will continue as long au fission continues. Fissioning does not stop after a trip.
Ref:RX Theory p I-3.17 b) disagree: Thermal flux levels within the fuel are lower because of self shielding and because neutrons are t her a a' i z ed in the moderator,not in the fuel Ref:ReacLor Theory I-6.13.
5.2 1647 psig Ref: steam tableu 5.3 decrease, increase Ref : Thermo /Hydraul I c p.10-58 5.4 false U.5 Ref: Technical Spectiications.p2.1-2 b
Ref: Reactor Theory :- 5.10 5.6 Critical rod position is predicted to be greater than the zero rod power rod insertion 1imit (G2 steps -b a n l-: c)
(24 steps on bani: b)
Ref:T/S 3.10-5,7/S incertion graph 3.10-3, T /G 3.10--8, par a 1 Reector utartup procedure pr ecaution 2.9 5.7 more negative MTC at EOL (.50) greater fuel depletion at the bottom of the core at EOL (.50)
Ref: Reactor theory 1-6.52,T/S 3.10-4,5 5.8 c Ref:Radiction Pratettion p.0-20, Emergency Rad. Cont.
EP- A D -5. 4. 7 (c ) , Tabl e AD-2.1 CHT-D,ED-EtW-3C,p.9 5.9 a) transit is slow piping delays-core to detector or flow within correct parameters b)pressure pressuretrips is within the range between the high and low Ref: T/G p2.3-5 U.10 Doppl er (FTC) (. 75) , f uel temperature changes firut (.75)
Ref: reactor theory I-5.21 1
i I
m
- 5.11 1)a'd equate trip reactivity
,2}to.a.ssure mneting power distribution limits 3)1imit the consequences of rod ejection accident Ref:T/S 3.10-8 5.12 a)there is no effect on equilibium samarium due to power changing (.50) because both the production and removal terms for equilibruim are not flux dependent *(.50) b)1)less negative 2)more negative Ref: reactor theory 1-5.77 5.13 An increase in temperature will decrease moderator density and at the same time decrease boron density (.50).The decrease in baron density (positive reactivity) will overshadow the decrease in moderator density (negative reactivity) and a net positive reactivity insertion will result. (.50)
Ref: Rx Theory I-6.7 5.14 a) decrease b) increases c) increases 1 d)decreanos Ref:T/S 2.1-2 5.15 each rcp motor running amps will decrease (.25)-there is more system head loss (.25) so lower amps are required as there is less fl ow per pump 1
Ref:thermo/ hydraulics p10-46,10-76,p10-44 5.16 Ensure that the axial peaking factor for the Fq(Z) (.50)is net exceeded
(.25) during normal operation (.25) or xenon redistribution Ref:T/5 3.10-9,3.10-14 5.17 a) decrease b) increase c)1ess negative d) increases Ref: reactor theory pl-5.31 5.10 oscillation would be convergent (self dampeni ng) ( . 75) and the oscillations would probably stop (.55)
Ref: reactor theory 1-5.57 S
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1 Category 6 Plant Systems design etc. answers.
6.01.OP DELTA T,OT D T, LOW PRESSURE TRIP,S/G LOW-LOW LEVEL, RCS LOW FLOW, NI HI FLUX TRIP,NI HI RANGE HIGH FLUX Ref :I:ewaunee LER 84-014 6.02 b Ref:Ler 84-016, Sept 14,1984 6.03 Steam flow transmitter was reading higher than feed flow,it made it possible to make up the coincidence (.50) with the the low (1.0) l evel S/G setpoint resulting in a reactor protection.
Ref:Ler 85-010, April 4,1985 ChaplV-12.18 6.04 a) The reattor protection system was actuated at 2000 psig.(.50) b) By 2/3 coincidence-Safety Injection Logic Gignal c) At 2000 psig-SI permi ssi ve block i s now removed by whi te channel greater than 2000 psig; the red channel was tripped, so safety injection is initiated as steam pressure is less 500 psig at ref ueling. (1. 0)
Ref:Ler 85-007,Feb 25,1985, chap IV- 11.13 6.05 A low low signal in tank U will not allow the SI p, umps to automaticall y swi tch over to the RWST because the control switch is selected to tank A, valves SI-4A/4D will not open to provide suction from the RWST,SI-2A/D will not close to prevent backflow Ref:Ler G4-021, Sys Lessons Chap 35,p22 6.06 When the mechanic trips the lever, the following actions occur; diesels start signals are generated by the turbine trip s i gn al , ( 1. 23 ) diesels are started. (.25)
Refsd5-008,Feb 25,1985 6.07 GI signal generates a:
- 1) feedwater isolation valves trip
- 2) both main feed water pumps trip off
- 3) main feed & bypass control valves are tripped closed
(.50) points each Ref: Chap Sa,pB,p12 &dwg xk-100-152 6.08 d Ref dwg xk100-35 6.09 a)R-13,R-14 b)R-15,R-19 Ref: Chap 32,p25 chap 45,p10 6.10 control valves,interecptor valves (.50)each Ref: Chap 3,p.111-1.6, fig 111-1.6, fig 111-1.14 6.11 a) waste hold up tanks b)1.CCW thermal barrier flow indicator
- 2. radiation detector 3. thermal barrier temp c)1. Thermal barrier isolation valve trips shut 2.CCW surge tank vent line shuts 3. thermal barrier t enip alarm RefaChap 31,p5,CCW sys 3
. l 6.12 a)1. prevent auto pump starts cue to both main feed pump breakers being open.
l b)2. prevent needless recirulation running' time of the aux. feed l water pumps.
c)3. consume to much condensate. storage tank water Ref:N-FW-OSB, precautions 2-3 (.50 points each) 6.13 1)D/P signal for feedwater control system
- 2) Develop safety related signals
- 3) Limit the rate of steam flow in the event of pipe break.
Ref: Chap 6,p1 6.14 all) maintain back pressure on letdown orifices to prevent flashing
- 2) maintain RCS pressure when plant is solid b) shut c)failcd open Ref Chap 35,xkl00-35-w-1,E2023 6.15 a)1) Blackout signal will generate a load shed signal
- 2) All loads which were sequenced on by the SI signal are tripped off'but will load on again.
b)The SI signal predominates and a SI sequence is now generated, so the blaclout sequence loads which were on are i stripped off and the SI loading sequence begins.
6.16 Prevents pumping contaminated sump water into the RWST with the RHR pumps.
Ref: Chap 33,p9,xk100-19,29 4
6.17 Possible pump failure due to steam binding
- i. Ref:NRC report 85-81 bulletin 64-06;85-01 Oct.29,1985 Kewaunee incident 4
Category 7-Procedures-NormalgAbnormalaEmergency and
- Nadi'ol'ogi cal control 7.1 Both condition exist:
1)one S1 pump running 2)RCS pressure <1240 normal
<1480 adverse Ref:E-O foldout 7.2 contai nment pressure >4psig containment radiation >100000 rem /hr Ref:E-O foldout 7.3 The steam generator tubes in the ruptured s/g will approach the temperature of the reactor coolant when the RCS is cooled and the RCP's continue to run. So when the steam space in the ruptured S/G expands to contact these colder tubes, condensation will occur decreasing s/g pressure, increasing the primary to secondary l eakage.
Or level will decrease the radioactive release rate Check at plant for plant specific data.
Ref: Westinghouse Owner's group procedures E-3 p6,p26 7.4 krypton , i odi ne , argon ,.< enon-supp l i ed ai r type respirator tritium-air supplied whole body suit Ref: Rad. Prot. Train. Man.p1V-3 7.5 elert, site emergency, general emergency Ref:EP-05F-2 7.6 1)subcriticality-nuclear power >5%
- 2) core cooling-incore T/C>1200*F
- 3) heat s: nk-NR i n 5/G (2%,and total feedwater flow <200gpm
- 4) integrity-cold leg temp decreases >100*F in last 60 minutes and Rcs cold leg temp <270*F
- 5) containment pressure >46 psig Ref:E-1 7.7 a)the high temperature coolant in the upper head which has a slight amount of cooling flow will flash to steam creating voids in the upper head. Check at faci 1ity for plant speciiic data, b) rapidly increasing pressurizer level Ref:ECA 3.1,p.9 7.0 a)Due to the injection of seal water cooling'into the RCG via pump cooling b)This cold slug of water is pushed into the hot S/G tubes or ex vessel by the RCP's which heats up this cold water which expands to create a pressure spike.
Ref:N-RC-36Ap2. CIRCLE W 5
4 s
1 1
I J
7.9 a)6o b)yos c)nd
- d)no e)no -
f)yes g)no h)no i)no t/s 2.1.1 j)yes Ref: T/S-Kewaunee 7.10 a)As the RCP is started, the steam bubble will collapse and the pressurizer level will decrease rapidly to fill the void in the reactor vessel.
b)If the RCP cannot be started,a rapid cooldown will make the void larger displacing water in the RCS causing an insurge i nto the pressuri zer , (optional ) 1 evel is to accomodate void growth during cooldown Ref:ES-0.3 step' 2 introduction 7.11 1)a or 1 2)d 4 3}b 2 4)c 3 5)e 5 6)f 6 Ref: Westinghouse Owners Group 7.12 Ghift Supervisor and Event Operations Director Ref:E-0-12,p1 7.13 1) Reset both Feedwater luolation Reset pushbuttons
- 2) start a condensate pump
- 3) start a main feedwater pump 4)use feedwster bypass control valve (.325 each)
Ref:A-FW-05B,p4, step 3.2.5 7.14 1)Two fire pumps shall be operab-lb at all times each with their 2)If onedischarge aligned to the" fire header -100 psig (.50) pump is inoperable status in 7 days (.25)
'( . 2 5 ) restore it to operable
- 3) operabl e f low path-circ water intake to sprinkler (.50)
- 4) insure backup fire protection-24 hrs (.50)
Ref: T/5 3.15-2,4a 7.15 If rods are in automatic, a rod drop will cause rods to drive out resulting in exceeding hot channel-factors limits;(axial power distribution differerence) limits, radial power distribution limits,anial flux Rei:N-O-03, precautions 7.16 a)272,500 gal-98% level
- b)1) prevent return to criticality
- 2) fill reactor vessel to nozzles 3)recirculation provide enough volume of water on containment flow to initiate (any two)
Ref:Kewaunee LER 84-015 6
, ,' Category 0-Administrative Procedures , Conditions and Lim:tations
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8.1" i) signs the temporary change request
- 2) places a danger tag tag to identify the temporary change by number and description.
- 3) insures an independent verification is completed if required
- 4) files original in SS temporary change log book
- 5) makes entry on the temporary change control sheet.
- 6) set up systems conditions necessary.for the ready installation of the temporary changes (ANY THREE)
Ref:ACD 1.6 8.2 Shift Technical Advisor Ref:ACD 2.16, step 5.1 8.3 a) Operations Superintendent or Asst. Manager- Plant Operations Asst Sup. OPS or Superintendent- Plant Operations Shift Supervisor b)6 weeks Ref:ACD 4.2 0.4 Emergency Director Event Ops Director Fire Marshal Fire Team Radiation Frotection Director ACD 14.2, step 5.26 0.5 1)c 2)d.e 3)b 4)a Ref:ACD 4.3 G.6 1)f,2)f,5)f,4)t Ref:N-ESF-55-CLA i
O.7
- 1) entry into any area posted as a radiation hazard area 2)maintenence on any component that has a contact radiation reading greater than 1r/hr Ref:ACD 6.3 RWP G.O a) discharge radiation monitor has been checked and tested for proper operation gtiot to the start of the liquid discharge 4
Ref:ACD6.7,p2, step 5.3,N-LWP-32-A-3, step 2.2.4 b)by meeting the precautions of a gaseous discherge procedure
-auto isolation devi ces capable of terminating the gaseous release shall be operable
-R-13/14 operable Ref N-GWP-32D, precaution 2.b G.9 plant manager uecurity director Re f ACD11.10 -
b G.10 4 hrs.or 1 hr Ref: Night order book,Feb 1,1905,p2 , A-M1-07 7
-n-- , - - - . - - - ---. , . - - . , ,
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.-8.11 a)IRPI should be declared out of service IAW T/S-3.10f b)Pfade rods in manual c)begin repairs of IRPI d)if r.epairs are not completed in 36_ hours, notif y NLSG Superintendent (optional) (any two) e)power check>50%
rod position by core instrumentation every shif t if Ref: Standing Orders, July 27,1984 8.12 4%,85%
Ref:T/S 3.11-1 8.13 Orifice will limit the flow from an inadvertant actuation of the vent system pump. to less then the flow capacity of one charging Ref:T/S 3.1-2e 8.14 false Ref T/S:det'inition-operable, Standing Orders Oct 3,1985 8.15 a) linesteam i sol atibreak on valaccident ve. ( . 45)(.80) combined with failure of a steam or a) postulated release of the s/g to the atmosphere b) double ended break of a single s/g tube Ref:T/S p3.1-9;p3.4-3,FSAR 10.3:14.1 8.16 It is only inoperable if it cannot open or close Ref:T/S definition operable, Standing Orders Oct 3,1985 8.17 false Ref: Standing Orders Sept 24, 1985 8~
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