ML20235E268
| ML20235E268 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/16/1987 |
| From: | Burdick T, Damon D, Isaksen P, Jaggar F, Lennartz J, Roesener W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20235E253 | List: |
| References | |
| 50-315-OL-87-01, 50-315-OL-87-1, NUDOCS 8709280039 | |
| Download: ML20235E268 (70) | |
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i U.S. NUCLEAR REGULATORY COMMISSION i
REGION III Report No. 50-315/0LS-87-01 Docket Nos. 50-315; 50-316 Licenses No. DPR-58; DPR-74 Licensee: American Electric Power Service Corporation Indiana and Michigan Electric Company 1 Riverside Plaza Columbus, OH 43216 Facility Name: D. C. Cook Nuclear Power Plant l
Examination Administered At: Bridgeman, Michigan Examination Conducted: June 22-24, 1987 Examiners: ifl0fF7 D. Q) Damon Date 9//b/97 hO
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,A.
ho Lennartz Dat'e ' -
li' P. T. Isaksen~
k ll f//fG Ddte/
1 76 Alp, F. S. Jaggar w
D6t6 1
W. S. Roesener 1/Sh7 Dhte' Approved By: lIl D I /b T.'M. Burdick, Chief Dpte' Operating Licensing Section l
Examination Summary Examination administered on June 22-24, 1987 (Report No. 50-315/0LS-87-01) to 12 Senior Reactor Operators and five Reactor Operators.
Results: Seven Senior Reactor Operators failed and two Reactor Operators failed.
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-l REPORT DETAILS l
- 1. Examiners'
- D. J. Damon J. A. Lennartz I P. T. Isaksen -
F. S. Jaggar W. S. Roesener j
- Chief Examiner- j i
- 2. Examination Review Meeting '
Refer to Attachment 1.
- 3. Exit Meeting
- a. On June 25, 1987, at the conclusion of the requalification examinations, the examiners met with members of the plant staff to discuss generic findings made during the course of the examinations. The following personnel attended the exit i meeting: j W. Smith, Jr. , Plant Manager J. Rutkowski, Assistant Plant Manager, Production A. Blind, Assistant Plant Manager, Organization and Administration K. Baker, Operations Superintendent J. Stubblefield, Training Superintendent W. Nichols, Training Manager R. Strasser, Training Instructor R. Heydenburg, QA Auditor, AEPSC l P. Isaksen, Examiner, EG&G 1 F. Jaggar, Examiner, EG&G W. Roesener, Examiner, EG&G J. Heller, Resident Inspector, NRC D. Damon, Examiner, NRC G. Nejfeit, Examiner, NRC (1) Generic Weaknesses The following generic weaknesses as found during the examination process were discussed:
(a) Candidates were generally slow to use procedures'during the' scenario walkthrough portion of the exam. This was especially evident for procedures that are considered
" common usage." Exceptions to this were the Emergency Operating Procedures and the Mandatory In-Hand Procedures.
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c (b) Basic system knowledge was weak. Mos't SR0 candidates q were unable.to explain how to manipulate a system with- '
the controls and instruments available on the control boards.
(c) Some SR0 candidates were not aware of or were unable ;
to explain the basis of the. Technical Specification LCO's.
In some instances, changes.to Technical. Specifications resulted in a difference between the units that the '
candidate was unaware of.
(2) Generic Strengths The following generic strengths were noted during the exams:
(a) Candidates were aware of industry wide'and plant LER's and the actions that were taken in the plant in. response to these LER's.
(b) Emergency Operating Procedure usage was good.
,. (3) General Observations The following general observations were m ue during the exam:
(a) Procedures for entry to and exit from the Auxiliary Building were inconsistent, sometimes changing on an hourly basis. There is a need for a consistent policy that is known to both the plant staff and the examiners.
(b) Some procedures in the Control Room were either missing pages or are in need of revision.
- b. At the exit meeting and in subsequent telephone conversations, there was a discussion between plant management and NRC personnel regarding the need for specific and detailed knowledge in the areas of refueling j l systems and procedures on the part of licensed operators not routinely l
involved with refueling operations in the. Auxiliary Building and in containment. The plant management has taken the position that detailed knowledge in these areas is not required for most licensed operators and would only be taught to SR0's specifically designated to supervise core alterations (SRO-CA). The SRO-CA would be the only individual authorized to supervise fuel movement, thus eliminating the need for examination of these knowledge areas for other licensed operators.
The NRC takes the position that knowledge of refueling systems and casualty procedures is required for all licensed operators.
NUREG-1021-ES-302 and 10 CFR 55.45.a.8 provide the bases for questioning in these areas during the operating test phase of the exam for R0's and SR0's. NUREG-1021-ES-402 and 10 CFR 55.43.b.6 provide the bases for questioning in these areas on the written test for SR0's.
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- 4. On July 24, 1987, a meeting was held between W. G. Smith,-dr. and members of the D.C. Cook staff and H. Miller and members of the NRC. staff in the l Region III office to satisfy requirements of CAL-RIII-87-012. The j licensee presented an analysis of the results of the June requalification exams and the proposed short term and long term remedial actions to be taken. It was decided by the NRC staff to examine an additional 20% of- a the licensed personnel at D.C. Cook on August 7, 1987. The results of j the August 7 exam are as follows: 1 Type Candidates Passed Failed Pass Rate l l R0 5 5 0 100%
SR0 12 4 8 33%
'1 Because of the results of the August 7 exam, an onsite inspection an.1 l evaluation of the D.C. Cook requalification program was conducted by I members of the NRC regional staff during the week of September '7,1987. :!
The results of this inspection are contained in IE Inspection Report i No. 50-315/87027.
- 5. On September 9, 1987, a meeting was held between M. P. Alexich and members i of AEP Service Corp and H. Miller and members of the NRC staff in the i Region III office. This satisfied the requirements of ' CAL-RIII-87-012 and l CAL-RIII-87-012 Amendment 1. The results of the August 7 exam were discussed,- along with findings by the staff of D.C. Cook and AEP Service Corp concerning the requalification program at D.C. Cook. Weaknesses in q the program were discussed and long term corrective actions were outlined. l At the conclusion of the meeting, the following agreements were reached i between the NRC staff and D.C. Cook staff:
- a. The NRC staff will perform a complete parallel grading of l the Week 4 and 5 facility requalification exams. l
- b. Any personnel failing the Week 4 and 5 requalification exams based on the grading noted above will be removed from shift duties,
- c. The NRC staff will continue to be involved in the remediation of the failures from the August requalification exams.
- d. The D.C. Cook staff will submit a detailed program of improvements to the requalification program to the NRC staff in writing within 60 days.
As a result of these discussions CAL-RIII-87-012 and CAL-RIII-87-012 Amendment 1 are considered closed.
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. Facility: D. C. Cook '
Examiners: Damon, Lennartz, Isaksen,-Jaggar, and Roesener .
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Dates of Evaluation: June 22-24, 1987
- Areas Evaluated: X- Written X Oral Simulator i l
Examination Results: l 1
R0 SR0 Total Evaluation . )
Pass / Fail Pass / Fail Pass / Fail (S, M, or U) I Written Examination 3/2 '4/7 7/9 U l Operating Examination Oral 5/0 12/0 17/0 S .
1 Simulator NA NA NA NA l Evaluation of facility written examination grading NA Overall Program Evaluation Satisfactory Marginal Unsatisfactory X SR0 - Major deficiencies exist in the areas of instrumentation. systems and control systems.
RO - Major deficiency exists in the area of instrumentation systems, tb.itted: Fo ded: - A oved:
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. J. Damon T. . Burdick / . Hehl l
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1 ATTACHMENT 1 .,
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The specific facility comments concerning.the requalification examinations, followed by the NRC resolution, are listed in the following-paragraphs.
Question 2.03 I
Facility Comment: Modify answer key to allow full credit for'use of Demin I Water Makeup Plant and Hotwells as water sources for Aux l Fd Pps.
- Supporting Reference /Information 1
Abnormal Operating Procedure OHP-4022.055.002, " Loss of Condensate to Aux Feed Pumps," lists several sources of water to the Aux' feed Pps. These include: .
4.2.2.1 Opposite unit's condensate storage tank.
4.2.2.2 Place in service all available trains of makeup. l 4.2.2.3 Pump hotwells to condensate storage tank.
Ref. 0HP-4022.055.007 hac been included in exam prep materials.
NRC Resolution: The references sited on the examination by the examiner and above by the facility both state that the sources of water to the auxiliary feed pumps are as stated in the exam l answer key, i.e., Affected Unit s CST, opposite Unit's CST, l and ESW. The use of all means available to keep the CST's l
full is not' required for a complete answer. The answer key has been changed to acknowled of the steps to keep the CST'ge acceptance, s full as follows: for no credit,
" Accept for no credit:
Place in service all available trains of make up.
Pump hotwells to CST.
Credit will be deducted [.25 each] if any other answers are given."
Question 2.10 Facility Comment: Modify answer key to allow full credit for either TRUE or FALSE.
Supporting Reference /Information
Attachment l' 2' If 'the candidate assumed '! manually operate" referred to' local manipulation of the valve handwheel, the answer would be FALSE. Interlocks on these valves are electrical not mechanical therefore there are no interlocks for local valve manipulation.
, If the candidates assume " manually operate" referred-to operation of the control switch, the answer would be TRUE. Page 23 of lesson plan R0-C-NS08 Rev 1 described the interlocks for IM0-330 and IM0-331 as follows.
"Valvewillopenif.ICM-305,RHRpppsuction'from containment recirc valve is open I
and "Same as IM0-330 except west train components."
R0-C-NS08, Revision 1 is included for reference'.'
NRC Resolution: It is agreed that the question may be misinterpreted,-
therefore, the question is deleted from the test.
Question 3.05 Facility Comment: Modify annter key such that inclusion of "because charging continues [0.5]" is not required to receive full credit.
1 Support Reference /Information As worded, the question assumes normal operation of the CVCS system prior to the VCT level instrument failure.
For this reason, candidates need not mention operation of a charging pump in their answer. Diversion of letdown from the VCT should suffice.
NRC Resolution: It is agreed that the continuation of char ing is assumed in the question statement, therefore, the 0.6] credit for "because chargi,ng continues" is removed from the question, the statement is placed in parentheses, and the point value for the question is reduced to 2.0 from 2.5.
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Attachment l' 3- -
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1 Question 4.11 Facility Comment: Modify answer key to accept " locally open the generator :j output breakers for the rod drive MG Sets 2N and 2S located in the 600 V Switch Gear Room."
Supporting Reference /Information.
Even though this.is not stated in OHP-4023.001.011,.
it is'an acceptable method as stated in OHP-4023.FR-5.1
" Response to'ATWAS." Page' attached.
NRC Resolution: The question asked for those specific immediate actions contained in a specific procedure. These actions are 6 what is required for.the answer, alternate actions from 1
-other procedures are not allowed for credit. The answer j key remains unchanged.
I Question 4.12 Facility Comment: Modify answer key to allow similar wording for l
full credit. For example, i
would place the unit in an unanalyzed condition i
or Unit not analyzed for this configuration.
Supporting Reference /Information This wortiing is similar to " operation per the FSAR.'"
NRC Resolution: The above statements are not significantly different from the stated answers, therefore, they will not be included in the answer key. !
Question 5.02 I Facility Comment: Modify answer key as follows and accept similar wording / calculations.
RCS press 2215 psig is 2230 psia j 0
T sat at 2230 psig is between 650 F and 655 F on Steam Table 4
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Attachment 1 4 (T sat at 2200 psia = 649.45)
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'(T 3at at2300 psia psia=655F]F) a at 2239.2 -.652 Low and T core exit -Tsat = F subcooling 650 F - 554 F = 106 F High end Tcore exit -T sat l 655 F - 544 F = 111 F d
Accept answer from 106 F to 111 F subcooling.
NRC Resolution: Answer key modified.
Question 5.04 i
l Facility Comment: Modify answer key for the second part of the question to )
include other valid conditions required for flux boiling. !
For example, Availability of a S/G as a heat sink.
Interruption of Natural Circulation.
NRC Resolution: Answer key modified. I Question 5.06 Facility Comment: Modify answer key to include other valid indications of pump cavitation and/or'similar working of those listed.
For example, fluctuation of pump discharge flow NRC Resolution: Answer key modified.
Question 5.09 Facility Comment: Modify answer key to allow full credit for an additional U1/U2 S/G difference.
For example, differences in the annealing or manufacturing 3rocess.
NRC Resolution: Jo not concur, manufacturing differences are not
' operating conditions' of Unit 2.
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Attachment 1 5 Question 6.05 Facility Comment: Modify answer key for Part a. to include " Reactor will i trip on lo-lo S/G 1evel or Steam Flow greater than Feed ;
Flow with Lo S/G leve. For Part b., if it is assumed that the FP does not overspeed, then SGWLC will maintain 1 programmed level with'an excessively high FRV D.P.
Therefore, no protective action would result. I i
NRC Resolution: Part a. answer key is modified to accept SF > FF reactor trip. Part b., the assumption that MFP will not trip is invalid and is not accepted.
Question 6.10 Facility Comment: Modify answer key to allow full credit for either "C-9 permissive not met" or "no circulating water pumps running."
NRC Resolution: Partially concur, answer key modified to accept C-9 or steam dump block.
Question 7.01 y Facility Comment: Modify answer key to allow similar wording for full I credit on Part 2.
For example, hydrazine or ammonia within the system l will strip mixed beds of chlorine.
NRC Resolution: Similar wording is acceptable. .
Question 7.02 Facility Comment: Change answer key for Part b. to False. The D/G is considtred operable when paralleled to the system.
NRC Resolution: Due to confusion between the procedure and technical specification clarification, this part is deleted.
Question 7.05 Facility Comment: Modify answer key to accept "30 minutes after the accident" as an additional answer.
l Attachment 1 6 l
NRC Resolution: Answer key has been modified to accept 30 minutes after the accident. ,
l Question 7.06 l Facility Comments: Modify answer key for Part a. to include other valid )
indications of pump cavitation resulting from steam binding.
For example,-fluctuation in pump discharge press / flow.
Modify answer key to allow for operator action that would result in reseating leaking check valves.
For example, momentarily feed the S/G.
NRC Resolution: Answer key modified to accept other indications of cavitation and means that will result in reseating leaking check valves.
l Question 7.11 l
Facility Comment: Modify answer key to allow full credit for a description of the symptoms of a faulted S/G.
For example S/G pressure decreasing in an uncontrolled manner.
S/G completely depressurized.
NRC Resolution: Full credit will be given if all indications per emergency procedures are covered.
Question 8.07 Facility Comment: Modifyanswerkeytoallowfor"a. Fire in SI pump motor' to be listed either as notification required or notification not required.
NRC Resolution: Deleted Part a. due to nossible ambiguity of interpretation.
Question 8.08 Facility Commert: Modify answer key to not require specific percentages of iodine absorption and gap activity for full credit.
NRC Resolution: Exact percentages are not required.
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Attachment 1 7 The following changes were made to the examinations as a result of final review and/or candidate questions during exam administration. All the.
changes to the question statements were publicly announced to all candidates
'during the exam by the exam proctor.
- Question 2.10 1 Changed to read
l "The Residual Heat Removal (RHR) system valve to the upper containment spray header for RHR (IM0-330 or 331) is interlocked such that it can be opened manually only when the RHR suction valve to the containment sump is open."
Comment:
Changed to reflect appropriate facility nomenclature. t l
,uestion Q 3.05 l The second sentence was changed to read: l "As a result, the indicated level in that leg fails high."
Comment:
Change to reflect indicated vice actual level response.
Question 3.06 ;
The words "and vented." were added to the question stem. l The first sentence of Part a was changed to read:
. . . in the first five seconds after isolating and venting the steam pressure instruments."
Comment: ,
Changed to indicate that instrument was isolated AND vented. j i
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_ Question 5.07 Changed to read:
" Indicate at which time in core life (B0C or E0C) the following accidents !
will recult in a higher steady state power. Assume no trip." l
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l Attachment 1 8 Comment:
Changed to clarify intent of question.
Answer 1.02 Added parenthesis as follows:
"Some of the Antimony absorbs neutrons (during operation) [.50] and the excited antimony nuclei emit (high energy) gammas [.50] some of i the (high energy) gammas are absorbed in the Beryllium [.50] which then emits a neutron [.50].
Comment:
To clarify what part of answer is required for full credit.
Answer 1.06 '
Added the following:
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. . . neutron precursor, (the major neutron source shortly after l reactor trip, holds the level at -1/3 DPM)"
and ;
"or .
I Decay of. Bromine 87 (the major neutron source shortly after reactor l trip, holds the level at -1/3 DPM)."
Answer 1.11 Added:
"(or combined with high pressure)"
Comment:
Thermal shock combined with either the rapid recovery of pressure or combined I with high pressure will satisfy the definition of PTS.
Added
Reference:
"WOG ERG, Executive Volume, Generic Issue PTS, Page 3" i
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Attachment 1 9 Answer 2.05 1 Added the following grading clarification:
.5 for standing signal
.5 for containment isolation Phase A Answer 2.12 Added the following grading clarification:
"If an explanation for the instrument response is given, .25 points will be given for the instrument identified and .25 points will be given for the correct explanation of its response,"
l Answer 3.06a Changed to read:
" Isolating and venting the steam pressure instrument causes the j compensated steam flow input to the level control system to fail 4 low [0.5] which results in the shutting of the Feed Water' Regulating Valve and/or the decreasing of main feed pump speed and/or the lowering of the feed flow (accept any combination of the three for credit) [0-.5]
l which results in the lowering of SG 1evel [0.,5]."
l Answer 3.07 Will accept the following additional answer:
1 l "S.R. high voltage will remain blocked." l Comment:
Failure of the lower power permissive (P-10) to operate will also affect source range high voltage.
Answer 3.09 Added the following grading clarification:
Setpoints and coincidence is not required for full credit. If given,
.1 point is lost for each incorrect setpoint and .1 point is lost for each incorrect coincidence.
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Answer 4.13 Added the following grading clarification:
Coincidence not required for full credit, but if given and incorrect, I
.1 point is lost.
Answer 5.01 Part c modified to accept:
decrease (Unit 1) or_same (Unit 2)
I Comment:
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Due to differences in the Xenon curves for the units. I i
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U. S.. NUCLEAR REGUL'A I I - \ i
- i REACTOR OPERATOR REQUALIFICATION EXAMINATION l
. FACILITY: QQQBi162_________________
l REACTOR TYPE: CBE-HEES ________________
DATE ADMINISTERED: @Zf96/ZZ_________________
i EXAMINER:~ RQESENE8r_Sz_____________
CANDIDATE _________________________
IEEIEUCII96E_I9_C6N9I96IE1 Read the attached instr uction page carefully. This examination replaces i the current cycle facility administered requalification examination.
as Retrainins req'uirements for failure of this examination are the same l for failure of a requalification examination prepared and administered by in your training staff. Points for each question are indicated parentheses after the question. The passins grade requires at least 70%
in each category and a final grade of at least 80%. Examination papers will be picked.up four (4) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__Y6LUE. _1916L ___EG9BE___ _Y6LUE__ ______________G6IEG9BX_________..___
22162 1. PRINCIPLES OF-NUCLEAR POWER
_1E199__ ___________ ________
PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW l i
- 2. PLANT DESIGN INCLUDING SAFETY
_16 E9._ _2Ez20 ___________ ._______
AND EMERGENCY SYSTEMS
________ 3. INSTRUMENTS AND CONTROLS
_162E9._ _2E220 ___________
- 4. PROCEDURES - NORMAL, ABNORMAL, >
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_191pg__ _Z9191 ___________ ________
EMERGENCY AND RADIOLOGICAL CONTROL
________% Totals
.621E9__ ___________
Final Grade All work done on this examination is my own. I have neither siven not received aid.
Candidate's Signature a 9TtaCOPS ~D
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., i c- - NRC RULES AND GUIDELINES FOR' LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the' examination means an' automatic denial of your application and could result in nore severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave.- You must avoid all contacts with anyone outside the e'xamination room to' avoid even the ' appearance or possibilityfof cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print'your name in the blarik provided on the. cover sheet of the EMEminEtion.
- 5. Fill in the date on the cover sheet.of the examination (if necessary).
l 6. Use only the paper provided for answers.
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- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write 'End of Category __' as appropriate, start each category on a new page, write only on one side of the. paper, and write 'Last Page' on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip-st least three lines between each answer.
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-11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for.the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
' 15 . Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
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- 17. You must sign the statement on the cover sheet that indicates that the j work is your own and you have not received or been given assistance in j completing the examination. This must be dor.e' after the examination has
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been completed.
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18.When you complete your-examination,.-you shall .i
'i Assemble your exasiinstion as ' follows:
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(1) Exam questions on top. .
(2) Exam' aids - figures, tables,.etc. ;j
.y (3)- Answer pases including figures which are.part oflthe answer.
- b. Turn in your. copy of the examination and all pages used to answer i the examination questions.
- c. Tui n in s11 serap paper and the balance of the paper that you did {
not use for answering the questions. l 1
- d. Leave the examination area, as defined _by.the examiner. If after leavins, you are found in this area while the examination is still-in progress, your license may be denied.or revoked.
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12__EBINGIELEE;9E_hWGLEeB_E9BEB_ELOUI_9EEBeIIQHz Pase 4 .
F
- 1dEBd991d6 HIES' HEAT TRAN@((R'AND F(ylg_E(QW I
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OUESTION- 1'.01 '(1.00) .
l State whether the following cause the doppler-only-' power' coefficient ^
'to'become MORE~ NEGATIVE or~LESS NEGATIVE.
- a. Clad' creep o
Fuel'densificat' ion b.
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l GUESTION 1.02 (2.00) ;
Describe how an Antimony Berylium source produces neutrons? An equation is NOT desired.
-l QUESTION 1.03 (1.00)
Multiple Choice During a startup it was determined that Keff was equalito 0 9 when the Source Range (SR) instrument was reading 50 cps. 'What would the SR instrument be reading if rods were withdrawn to bring Keff equal to 0 95? Assume DOL conditions.
i 75'ces s.
- b. 100 cps l
- c. 125 eps l
- d. 150 eps l
OUESTION 1.04 (1.50)
Describe how the following affect moderator ~ temperature coefficient (MTC). (MORE NEGATIVE or LESS NEGATIVE).
- a. Increasing boron concentration
- b. Decreasing moderator temperature
- c. Core aging
(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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.. ,. e Pase 5 li.:EBIBLIELES_QE_NWELEeB_EQWEB_EleHI_QEEB'eII9Hz ISEBd90lNeb1GSt UE61_IB60@EEB_609 E(MID_ELQU t ,
QUESTION l.'05 '(1.'00)
What two physical characteristics of the1 fission product Xe-135l result 1 in it being a major' concern during reactor operation?'
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- Gl'ESTION 1.06 (1.00)
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Why is-the startup rate level, after1the initial prompt-drop >Llimited u to -1/3 decade per minute after a reactor trip? .(A numerical' proof is NOT cequired.)
DUESTION If07 .(0.50)
TRUE or FALSE?
One of the functions of the rod insertion limits is to prevent the Moderator Temperature Coefficient from:becoming positive.
j QUESTION 1.08 (0.50) l TRUE or FALSE?
The closer to criticality that a reactor _is, the GREATER will b'e'the INCREASE in count rate due-to a step insertion of reactivity.
( Assume the step insertion is the same size ar.d that the reactor ;
remains soberitical.)
l l-(***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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Page 6 1.t__EBINEIELES OF-NUCLEAR POWER PLANT _QPER6TIQN1-ISEBdggydedIggt_SE61_IS6BSEEE_669_ELQIQ.E(QW QUESTION 1 09 (1 00)
Choose the answer ~that most correctly completes the sentence'.
In a CLOSED system',.two single stage centrifugal pumps-operating in parallel will have--(choose-from-below)--, as l- compared to the same system with one single stage. centrifugal ,
ponip opor at i ng and one pump ,isc;sted. I
- a. the same head and the same flow rate.
- b. a higher head and higher flow rate.
- c. a higher head and the same flow rate.
- d. the same head and a higher flow rate-QUESTION '1.10 (3.00)
Why is Steam Generator (SG) pressure lower sat 100% load than at 50%'
load? (Use the equation, 0 = U A ( T a v s . T s t ni) , for priniary to secondary heat transfer in your answer. Include a discussion of how' 1 EACH terni in the equation is af *'ected by the load change.)
1 QUESTION 1 11 (1.00)
Define Pressurized Thermal Shock.
QUESTION 1.12 (1 50) l Hot channel factors are measurable and their Technical Specification surveillance frequency requiren'ents are relatively low provided four itenis are monitored and verified to be within their limits. Provide THREE of these four items (conditions).
l (***** END Or CATEGORY. 1 *****)
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P898 7 Z:_:ELehI_9EEI98 INGLV9ING_EBEE11_B89_EUEBGEUGX EXEIEd!
QUESTION 2 01 (2.00)
List the FOUR ways that the Main Control Room Cable Vault Halon System can be actueted. I l
GUE9 TION 2.02 (0 50) g ,
TRUE or FALSE?
If an SI occurs one s.inute after a blackout, then the Diesel Generator output breaker will trip to avoid overloading the Diesel Generator when the SI loads start sequencing on to the bus.
QUESTION 2 03 (2.00)
List the water sources to the auxiliary feedwater pumps in order of preference.
1 l
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OUESTION 2.04 (1.50)
State the alternate sources of power that can be used for the 41.
following equipment during Emergency' Remote Operations of Unit
- a. Unit 41 Residual Heat Removal pump (TWO sources)
- b. Unit #1 Pressurizer Backup Heaters (One source).
QUESTION 2 05 (1.00) t What condition would necessitate placing the containment hydrogen sample bypass switches to ' bypass' in order to begin post accident containment hydrogen monitoring system sampling of the containment l atmospher e?
l GUESTION 2.06 (2.00) l State two reasons why injection of high pressure oil by the RCP oil l j
lift system reduces the RCP starting torque / current?
I
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2__: Ele &I_DESIGH_INGLWDING SeEEI!_eHQ_ENEEGENGX Page 8 EXEIEbH QUESTION 2.07 (1.00) ",
State the~ function of the low leakage divider inside the' containment.
OUESTION 2.08 (1 50)
Match the pressure et which injection starts in Column B.to the' component of Column A.
Column A Column B
- 3. Accumulators c. 2290 psig
- d. 1560 psis
- e. 1160 psig
- f. 620 psis 3 200 psis
- h. 170 psig 00ESTION 2.09 (1.00)
Why is the positive displacement charging pump provided with a low speed stop?
i OL,ESTION 2 10 (0.00)
DELETED
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. EXEIEUS QUESTION- 2.11 (2 00)
List FOUR design features that prevent-inadvertent draining-of the Spent Fuel Pit.
1
-QUESTION 2.12 '(2.00) '
The hot les suction valves to the Residual Heat Removal (RHR) system (IMO-128 and ICM-129) are interlocked to. prevent damage to the'RHR system downstream of the RHR pumps. State both thecUnit 1 and Unit 2' setpoints st which these valves are allowed to be opened and at which these valves are automatically shut.
1 1
(***** END OF CATEGORY 2 *****)
.. .- j az__INEIEW5ENIE_eHD.G0HIB96E Pese 10 QUESTION 3.01 '( 2. 00 ) .
What.TWO rod stops are only effectiveLwhen in the automatic mode of .,
.)
rod control? Includeisetpoint of,each.
1 i
"OUESTION 3.02 (2.50) 1 1
Assume. steady' state operation at 100% power when tie. Master j Pressure Controller setpoint for-the pressurizer is inadvertently- l changed.to 2385 psis. Assume a step change in setpoint and.essume that j I
pressurizer pressure control is in' automatic.
- a. What automatic action, besides the actuation of alarms / annunciators, will occur.immediately?- ,
(1.00) l I
- b. Desc'ibe r the pressurizer pressure transient that will occur if no operator action is taken. Include in your answer any other automatic actions, besides alarm / annunciator actuations, that take place. (1.50)
OUESTION 3.03 (1.00)
Which of the following is the correct symptom of a SHORTED Cold Leg l RTD?
- a. T cold indication goes high
- b. T average indication goes high
- c. Delta T indication goes high
- d. T cold fails as is.
QUESTION 3 04 (1.00)
What TWO control roon, indications are used to differentiate between s ;
leak above and a leak below the hydraulic isolator (bellows) in.a '
pressurizer level reference les?
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Pese 11 2(__INSIBudESIH_eND_G9BISQLE 0UESTION 3 05 (2.00)
A leak develops in the reference les associated with the automatic level controller of the Volume Control Tank (VCT). As a result the indicated level in that les fails high. Describe the VCT level transient assuming that no operator action is taken and that the VCT is in automatic makeup mode. Be sure to include the reasons why level changes.
QUESTION 3,06 (3.00)
In order to ellow for the repair of a single Steam Generator'(SG) steani pressure instrument, the associated steam flow / feed flow mismatch protection bistables are tripped in accordance with an approved work instruction. The steani pressure instrument is then isolated and vented.
- a. If the steam flow instrument associated with the isolated stean pressure instrument is being used to control SG 1evel, expirin t he change that you would expect to see in SG 1evel in the first five seconds after isolating and venting the stean pressure i n s t r u nie n t . (Assume the plant is at 100%
power, the SG 1evel control systeni is in automatic, and that no operator action is taken. Your answer should include a description pf the effect of the isolation of the SG stean, p r e s s t. r e instrunient on the SG water level control system and the resultant change in SG 1evel.)
(1.50)
- b. If a reactor trip were to occur as a result, what would be the cause of the trip? (Setpoint and coincidence not required.)
(0.50)
- c. State two different ways that the control room operator could have easily avoided this transient while still acconiplishing the repairs. (Assume that the pressure instrument MUST be isolated.)
(1.00)
OUESTION 3.07 (2.00) I With channel 41 power range instrunient out of service and properly bypassed, what four protective system features will remain disabled as power drops to 5% if the channel 42 P-10 bistable fails to actuate?
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Ei_sINEIBudENIE_eHD_GQNIBQLE Paga 12 i
OUESTION 3 08 .(1 00)
L Which ONE of the.following status conditions indicates'en INOPERATIVE Radiation Monitoring Channel? .
- )
- a. Normal
- b. High Radiation Level
- c. Trend Alarm d.' Standby GUESTION 3.09 (2.00)
Safety ~ Injection (SI)-in both plants may be initiated automatically by one of four different signals. Three of the signals,' Low Pressurizer Pressure, High Containment Pressure, and High Steam Line Differential Pressure, are common to both plants.- The fourth signal differs between units. State this fourth SI signal, for DOTH units, including any required coincident signals.
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'l DUESTION 4.01' (1.00)-
Why does ECA-0.0, ' LOSS OF ALL AC POWER *, have. priority lover all Functional Restoration' Procedures? )
)
DUESTION 4.02 (3.00)-
1 i
State, .in ordert.253, the NINE action steps required to initiate BIT injection in accordance.with step 4 of FR-S.1, ' RESPONSE TO NUCLEAR POWER GENERATION /ATWS't.25 ea.3. Include the' expected !
responses for the first and last actionst.25 es.J. (When' valves are to be operated, state how many.)
i GBESTION 4.03 (0.50)
Throughout the Emergency Procedures there are steps that say either,
- s. " Task must be completed prior to proceeding," I or
- b. ' Task need NOT be completed prior to proceeding.' l 1
If the task did not specify either of the above statements, which- j one (a or b) applies? l OUESTION 4.04 (0.50)
TRUE or FALSE?
Plant personnel shall inform the Control Room Coordinator's Office imn.ediately whenever ANY plant e qui pnient is determined to be inoperable.
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OUESTION 4.05 (0.50) ,
Multiple Choice
'I Which of the statements below (a, b or c) most correctly describes the use of a pressurizer level. instrument _under the following conditions:
- Containment pressure'of 2.2 psi, and l
A LOCA is in progress and action is being performed IAH the E0Ps, and j pressurizer level instrument has a beige sticker with a circle E-with slash on a white background?
- a. Under no conditions should the operator use the l
. . pressurizer level instrument. i
- b. The pressurizer level instrument is fully qualified and !
can be used without question. j
- c. The pressurizer level instrument can be used but any information should be backed up with other qualified instrumentation. H QUESTION 4.06 (0 50)
TRUE or FALSE?
A Reactor Operator (RO) is allowed, on his own initiative, to take reasonable action that departs from a Technical Specification in an emergency when the action is imniediately needed to protect public healt,h and safety.
+
4 OUESTION 4 07 (1.00)
Stateithe lowest level of operational personnel that may approve the use oi onerators in lieu of. Clearance Permits to guard control or isolation points for performance of emergency work, when the work is of a n'ature such that a delay to obtain the clearance would seriously prolong isolation of a component or problem, protective action or termination of an undesirable event.
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.00ESTION. 4.08 -(0.50)
TRUE or' FALSE? ;
i Raciation exposure result'ing from~ medical examination and/or-treatment should be reported to the. Radiation Protection Section..
QUESTION '4.09 (1.00)
What immediate actions should be taken if.a Reactor Coolant Pump.
bearing temperature exceeds 210 F during critical operations?
DUESTION 4.10 (2.00)
List FOUR of the FIVE immediate actions of the Control Room operator on a loss of refueling water level.during. refueling operations.
QUESTION 4.11 (2.00)
What are the two methods for tripping the reactor outside of the control room according to 4023 001 011, ' REACTOR SHUTDOWN FROM HOT.
STANDBY PANEL DUE TO CONTROL ROOH INACCESSIBILITY *? Include in your, answer the loestion of the tripping device.
QUESTION 4.12 (1 00)
It is stated in the precautions section of 4021.002.003, 'RCP OPERATION', that single RCP operation is permitted in mode 3, < 541 F, as long as the reactor trip breakers are open and the'MG sets are off.
Why must the trip breakers be open and the MG sets off?
l QUESTION 4 13 (2.00)
The procedure for LOSS OF POWER TO THE CRID DISTRIBUTION CABINETS, 4022.082.001, states that a loss of a single control room instrument distrubotion (CRID) power supply will cause a. reactor trip. For each unit state the logie signal that will cause the reactor trip and the-power below which the trip will no longer occur.
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t (**xxxxxxxx END OF EXAMINATION xxxxxxxxxx) l
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1 _ EBINGIELEE_QE_HQGLEoB_EDWEB_EL6HI_QEEB611gM1 Pege 17 ISEBdQDINedIGEt_UE6I_IB6HEEEB_609_ELVID_EL9H ANSWER 1.01 (1.00)
- a. LESS NEGATIVE
- b. MORE NEGATIVE REFERENCE 1&M R0-C-1253, pp. 10 & 11 Objective-Describe how FTC changes over core life.
192004K107 ..(KA's) j ANSWER 1.02 (2.00) l Some of the. Antimony absorbs neutrons (during operation) E.50]
l and the excited antimony nuclei emit (high energy) sa nima s E.50].
l Some of the (high energy) sammas are absorbed in the Berylium E.50]
which then emits a neutron E.503.
REFERENCE I&M R0-C-NS01, p. 10 Objective-Describe the method by which installed sources produce neutrons including appropriate equations.
192001K109 ..(KA's)
ANSWER 1 03 (1 00) b REFERENCE IRM RQ-C-RXT1, pp. 14 & 15 Objective-Solve subtritical multiplication problenis given an NRC equations sheet.
192003K102 ..(KA's) l l
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. . N net: .
12_.EBINGIELEE_9E_UWGLE6B_E9BEB_EL6HI_9EEBoIIDHz Page 18 IBEBdQ9IU8dIGEt_UE01_IBBHEEEB.oH9_ELDID EL9B ANSWER 1.04 (1 50) 4 LESS NEGATIVE ,
- b. LESS NEGATIVE
- c. MORE NEGATIVE REFERENCE IRM R0-C-1254, pp. 788 Objective-Discuss the effect of changes in moderator density on the six factor formula and discuss how MTC is affected by changing boron concentration.
192004K106 ..(KA's)
ANSWER 1.05 (1.00)
- 1. H2gh (fission) yield
- 2. High absorption cross section.
REFERENCE I&M RD-C-RXT1, p. 15 Objective-Exple'in why Xe is of concern during reactor operation.
192006K102 ..(KA's)
ANSWER 1.06 (1.00)
Decay of the longest lived delayed neutron precursor, (the major neutron source shortly after reactor trip, holds the level at -1/3 dpm.)
or SUR = -26.06(.0124) = -1/3 DPMEO.53 where .0124 is the decay constant of the longest lived delayed neutron precursor.[0 53 or Decay of Bron.ine 87, (the major neutron sou*ce shortly after_ reactor trip, holds the level at - 1/3 dpm.)
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Iz__EBINGIELEE_QE_UUGLEeB_EQWEB_ELeUI_QEEEBIIQNr. Page 19-TyggdQQ1N8MIGE, HEAT TRgNEEEE_eHQ_ELUIQ_ELQH i
REFERENCE -
IRM RD-C-RXT1, pp. 18-& 19 Objective-Explain why neutrons are important to reactor control.
Given an NRC' equations sheet, perform solutions using.the inhour equation. . ,
192003K107 192003K106 ..(KA's)
ANSWER 1.07 (0.50)
FALSE REFERENCE l IRM RD-D-1254, pp. 7&8 l Describe how control rods affect HTC.
192005K115 ..(KA's)
ANSWER 1.08 (0.50)
TRUE 1
i REFERENCE IRH RQ-C-RXT1, pp. 13 to 15 Objective'-Explain suberitical multiplication. Explain the relationship between 1/M and Keff.
192008K104 ..(KA's)
ANSWER 1.09 (1.00) l b.
REFERENCE IRM RD-C-TSG1, p. 16 WEC Thermal Hydraulic Principles and Applications to the PWR, p. 10-46 Objective-Explain the following concepts associated with Pump Theory:
Pump Laws.
191004K109 ..(KA's)
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1 3
1:
ANSWER 1.10- (3.00) l-Using 0 = UA(Tavs -Tstm) fl-I The hest transferred across the primary to secondary' boundary, 0',
is greater at 100%' load .00.53 .Since UAEis constant,CO.53 the temperature ~ difference acrossLthe boundary,.Tavs - Tstm, must-increase (forJthe heat transfer to increase).CO.53 Tevs rises es a function'of the, programming of reactor average temperature,EO.253 but the rise-is not enough to transfer the full asiount of heat.CO.503 Therefore,'the Tstm has to fall (to accommodate the total. amount of heat ~ transfer).CO.53 (Since the S/G is operating in a saturated condition), the S/G pressure follows the
-decrease in the Tstm.CO.25]
REFERENCE ]
IBM _RG-C-TSG1, p. 22 !
Objective-Explain the thermodynamic principles involved'in determining ]
heat transfer between primary and secondary coolant. I 193007K106 ..(KA's) 1
-l ANSWER 1.11 (1.00)
PTS is a transient that results in a RAPID primary s9 stem cooldownEO.53 combined with a rapid recovery of primary systea i pressure (or, combined with high pressure).EO.53 )
REFERENCE IRM RQ-C-TSG1, p. 24 '
WOG ERG, Executive Volume, Generic Issue PTS, p. 3 Objective-Explain the following concerning PTS: Initiating events.
193010K106 ..(KA's) 1
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1
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l ANSWER 1.12 (1 50) I Any three of.the following: ,
)
- s. Rod groups.' sequenced and overlapped. l b '. Rod insertion limits adhered to. l 1
1
- c. Axial flux difference limits adhered to. !
1
- d. Rod bank alignment' maintained. ;
i REFERENCE i
TS 3/4.1.3'8 3/4 2 1 and R0-C-NS0'4, p. 15
']
Objective-State the reasons for the rod-insertion limits and list the design cr'itseia assumptions.
193009K107 ..(KA's)
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ANSWER 2.01 (2 00)
- a. Automatically (signal from ionization detectors)
- b. Manually in control room (press. control room pushbutton or manual electric) %),
- c. Manually at Automan (pull pin and push red button'at Automan)-
- d. Manually at Main Cylinder (remove pin'and pull lever at the Main ;
Cylinder) I REFERENCE RD-C-AS19, pp. 8.8 10 1 Objective-Describe how the Control Room Cable Vault Halon system is l manually and automatically actuated,' including any interlocks.
086000K405' ~ 086000A406 ..(KA's) 1 ANSWER 2.02 (0.50) i j
FALSE I REFERENCE R0-C-AS10, p. 29 Objective-State the conditions that will automatically trip'the D/G output breakers. 1 064000K402 ..(KA's)
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. .::, a Ek6dl-QESIGN_INGLyQING_S6EEII_6HQ.ENEEsENGX e Page 23 I21
'9131E5S-I' ANSWER 2.03 -(2.00)
.a. Uriit's condensate: storage. tank
- b. Opposite unit's condensate storage tank c.. Essential Service: Water
(.5 for each correct answer, .5 for proper order)
Accept for NO credit:
place in service all available, trains of make-up.
pump hotwells to the CST.
(Credit will be deducted C.25_es.] if any other answers are given.)
REFERENCE
- 1 RO-C-AS11, p. 13 Objective-List the sources of water available to the.AFW system in l order of preference. j 061000K107 061000K105 ..(KA's) :
ANSWER 2 04 (1 50) l
- a. Spare Containment Spray pump power of Unit 42 l Spare RHR pump power of Unit 42 1
- b. 480 volt Pressurizer Heater Control panel of. Unit 42.
REFERENCE ;
RQ-C-1256, pp. 16 & 17 Objective-State the alternate sources of power for emergency remote operations equipment. .
010000K201 005000K201 ..(KA's)
ANSWER 2 05 (1 00)
A standing CO.53 containment isolation phase A signal CO.53. i
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... 1 Z _ _ _ Ei e NJ _ D E EI G N _I N G L U DI N G . E 6 E EII _ e N D _ E d E S E E N G 1 Page 24
'EIEIEbH .
REFERENCE 1
RQ-C-1252, p:.' 10 ;
~
Objective-Explain what action can be taken to place'PACHMS in service t with e containment phase A isolation signal standing. ;
028000A101 ..(KA's)' i l
1 ANSWER 2.06 (2.00)
By forming a film on the bearing surfaces (lubricating the i bearings),[1.03 and by physically. lifting the faces apart.[1 03 j REFERENCE RD-C-NS2P, p. 13.
Objective-Sate the function of-the following,RCP components: Dil' Lift' System. .
003000S004 003000K113 ..(KA's) l ANSWER 2.07 (1.00) l To divert steam flow from pipe ruptures in lower containmentCO.53 !
through the ice condenserEO.53. l REFERENCE RO-C-NS13, p. 7 i' Objective-State the function of the low leakage divider.
025000G007 ..(KA's)
ANSWER 2.08 (1.50) 1-b 2-d 3-f REFERENCE R0-C-NS12, pp. 8, 10 & 14 Objective-List the pressures at which ECCS equipment injects into the RCS.
006020K603 006020K601 ..(KA's)-
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?;- 1 2 _.EleBI_QESIGH_INGLWQIN9_SeEEIl eHQ_EdEBGgdgy Page 25 EXEIEdB ANSWER 2.09 (1.00)
To prevent the flowrate from falling below the minimum required for RCP, seals.
REFERENCE R0-C-NS06 Objective-State the function of the following CVCS-component-coolant chersing pumps.
004020K607 ..(KA's)
ANSWER 2.10 (0.00)
DELETED .
REFERENCE DELETED ANSWER 2.13 (2.00)
Any four of the followinst No gravity drains.
Suction level) lines are located neer the SFP surface'(19.5 inches below normal The adjustable support for the skimmers limits downward travel (to 6 inches below nornal level)
Pump discharge lines terminate (6 feet). above the fuel assemblies.
Pump discharge lines include anti-siphon holes (4 inches below normal water level)
REFERENCE !
R0-C-NS05, p. 5&6 !
l
-List the design features that prevent inadvertent draining of the SFP.
033000K401 ..(KA's) 1
+
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or. ,
2__,ELBUI_DESIGU_INGLUDING_S6EEI!_ BUD _EdEEGEUGX Pasa 26 EXEIEd5 i
ANSWER 2 12 (2.00) l i
Unit 1 open at 375.psig +/- 22 psis l shut at 600 psig ~/ l Unit 2 open at 425 psig +/- 15 psig !
shut at 600 psig i
REFERENCE i R0-C-NS08, p. 19 Objective-Discuss the valve interlocks associated with the RHR systeni. l 005000K407 ..(KA's) l, 1
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- ANSWER 3.01- (2.00).
'I
- a. Control bank D exceeds withdrawal limit.(C-11) e 230 steps.
s L .b. Turbine power (C-5),.15% 7 '
L )
(Interlock worth .66, setpoint worth .34) . 1 REFERENCE .]
-l R0-C-NSO4, pp. 11 & 12 j Objective-List the rods stops.when in manual and automatic control., j 001000K401 ..(KA's) ~l
( ANSWER 3.02 (2~.50)
~
- a. All pressurizer. heaters energizeti.OJ -
I
- b. Primary pressure rises [0.53 and then stabilizes at the setpoint of. 1 the pressure operated relief valvesEO.5]. (Two of the three).
PORVs will automatically open[0.53.
(It is not necessary for credit to state how many of the PORVs operate, but .5 points will be lost if the answer specifically says-thrt all THREE reliefs open.)
RFFERENCE R0-C-NS03 pp. 23-26 ;
Objective-Given a description of plant conditions, describe the j response to instrument failures both with and without automatic control. .j 010000K607
..(KA's) i ANSWER 3.03 (1.00) c 1 REFERENCE. ,
RD-C-MC06, p. 8 Objective-Correlate common f ailure modes of sensors to control roos.
indications.
002000K606 ..(KA*s)
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'l ANSWER 3.04. (1.00) 1
.. 1 Pressurizer. level Pressurized. Pressure
-(If'an e:-:planation of instrument response is given. .25 pts. will be j given_for the' instrument and .25 pts. will be given for the-correct )
eitplanation of its response.)
REFERENCE 1 I
i I
l R0-C-MC06, pp. 9L& 10 and RG-C-1242,.p. 8 :
Objective-Correlate common failure modes of sensors to control room indications.-
011000A211- 011000A210 ..(KA's)
ANSWER .3.05 (2.00) 1 . .
a With control level indicating high the actual VCT level will !
dropCO.53 (because charging continues)'but letdown is diverted' l f roni the VCT CO.53.- The VCT will eventually tur. conipletely drainedCO.53 because the charging pump suction.will.not shift to the RWSTCO.53.
REFERENCF j l
RD-C-NS06, p.' 32 and RQ-C-1242, pp. 8&9
! Objective-Given a description of plant conditions, desc' ribe' the )
l response to instrunientation failures ~both with.and without automatic '
control.
004000K605 004000K106 004000A301 004000A207 ..(KA's) l l
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l ANSWER 3.06 (3.00).
- a. Isolating and venting the. steam pressure instrument.causes the a compensated steamsflow input to the level: control system to fail l lowCO.5] which results'in.the shutting of the Feed Water Regulating Valve'and/or the decreasing of niain feed pump. ]
speed and/or the lowering of' feed flow (accept,any combination of j the three for~ credit) -[0.5] which results in the lowering of SG l 1evelEO.5]. l I
- b. Steam flow to feed flow mismatch concurrent with a lo: SG level. j CO.53
- c. 1. Place SG control in manual i
- 2. Select the alternate steam flow channel for level control. l i
E0 50 es.]
REFERENCE .
l RD-C-PG11, pp. 13-15 and Nuclear Plant. Evaluation, Operational '
l Problems $2905, Cook, N o v e nib e r , 1985 Objective-Given a description of plant conditions, describe the response to instrumentation failures'both with and without automatic j
control.
035010H112 035010A204 035010A203 035000K401 ..(KA's) )
ANSWER 3 07 (2.00)
Any four of the follwingt l
Intermediate Range (Neutron Flux Reactor) Trip (25% Intermediate Range Trip)
Power Range Low (Neutron Flux Reactor Trip)
(25% Power Range Trip)
Low power perniissive (P-10)
Intermediate Range Rod Stop (C-1)
Source Range high voltage will teniain blocked
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9,14 ..
Page 30-Lat._I,USIBydENIS_syg_ggyIggLS.
REFERENCE I
-RO-C-NS11-SH03, p. 7 and I E E Information Notice 486-105 Objective-List'all the trips, permissivestand setpoints; associated-with NIS, including setpoints and coincidences. -
015000K301 ..(KA's) i
': ANSWER 3.08 (1 00) l REFERENCE RD-C-AS21C, pp. 8-10 Objective-Given'the status of any RMS. channel, determine the status of-
.that channel.
073000A402 ..(KA's)
ANSWER 3.09 (2.00)-
Unit 1 High Steam Line FlowC.503 coincident with either Low Steam LinE-Pressuret.503 or LO-LO Tavst.503.
Unit 2 Low Steam Line Pressuret.503 Setpoint and coincidence is not required for full credit. If given,
.1 point is lost for each-incorrect setpoint and .1 point is lost for each incorrect coincidence.
REFERENCE Unit 1 and Unit 2 Technical Specifications, Sec. 3/4.3 2 Objective-List the conditions which will cause a safety' injection signal l- 013000K101 ..(KA's) l
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S2__ES99EQUBES_:_H9Bb6L1_6EH9BdBL1_EdgSEENgy ,
8N9_BeDIQL991 gel Q9EIB96 3
AMSWER 4 01 (1.00)-
Because FRPs assume at least one-train of safeguards is avail'able.
-t REFERENCE <
RG-C-E012, p. 19 Objective-State why ECA-0.0, ' LOSS.0F. ALL AC POWER *, has-priority over eny FRP.
000055K302 ..(MA's)
ANSWER 4.02 (3 00)
Verify-CCP(s)C.253 - RUNNINGE.253 Open CCP siu'ction from RWST valves - 2 Close CCP suction from VCT valves - 2 Close BIT recirculation flowpath valves - 3 Open BIT outlet valves - 2 Open BIT inlet valves - 2 Close charging flow to regen HX valves -2 Close RC letdown to regen HX valves - 2 Verify cold injection line (BIT) flowt.253 - ESTABLISHED.253
(.25 for each step except as noted for responses. .25 for correct order)
REFERENCE 01-OHP 4023.FR-S.1, pp. 3&4 Objective-List the five immediate actions of FR-S.1 000029G010 ..(KA's)
ANSWER 4.03 (0.50) b
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'32_.EB99E99BES_ . NORMAL, ABNORMA(t;[M[$ggdQY' .Pcge 32 l BSD B69196991Gek_G9 BIB 96
. REFERENCE R0-C-E011 SH-1, p. 3 and EOP Rules of Usage, p . '2 i No facility objective ~available.
194001A102' ..(KA'.s) 1 l'
1 ANSWER 4.04 (0.50).
FALSE ]
REFERENCE q
RD-C-1243, p. B Objective-Describe what' action must;be taken by plant personnel whenever any plant equipment is determined.or suspected to,be inoperable.
001000K00.2 .,
..(KA's)
ANSWER 4.05 (0 50)
I e
REFERENCE RQ-C-CRD1, pp. 5-7 Objective-Explain what information can be obtained fron. l instrumentation if the instrument is not environmentally, and/or seismically qualified but should be.
011000K605 011000G010 ..(KA's) 1 j ANSWER 4 06 (0.50) 1 FALSE REFERENCE RO-C-TSR1, p. 7 Objective-State the actions necessary to ...take reasonable action that departs from a license condition or Tech. Spec. in an emergency...'.
000007G010 ..(KA's)
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. l S _,ESDCEDUSES_:_H9BdeL _6ENDBdeL _EdESGEH91 Pagef33 680 S60196991966_G981896 ANSWER- 0 4.'7' (1 00) -
Shift. Supervisor ,
REFERENCE p PMI-2110, p. - 2_ .
No facility objective available.
194001K102 ..(KA's)
ANSWER 4.08 '(0.50)
.TRUE REFERENCE .. _
PHP 6010. RAD.001-Objective-State the administrative regulations of the'RP Manual-related to medical exposure..
194001K103 ..(KA's)
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l ANSWER 4 09 (1.00)
. Trip the reactor [.00] and then trip the RCPE.50]'. ~1 REFERENCE 4022.002.001, p. 5, step 4.2.4.2 i No. facility objective.available.
000015K303 ..-(KA's) d y
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'fr_.EB9CEEMBEE_ _ NORMAL, ASHgBM6(t_EMEBGEdQX' ~ Page 34 g; . *AND'RAQIQLQgIgAL_QQNTEg(
-l 1 1 ANSWER 4.10 (2.00)
Any four of the following: ;
1 i
I Notify the SS 1
Notify the SRO in charge of Core. Alterations l Sound the containment evacuation alarm f
l Have sump pumps turned off ,
1 1 Have spent. fuel pit cooling and skimmer system turned off -]
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REFERENCE ,
1 1
l 4022 002.006, p. 3, step 4.2.1 l No facili-t'y objective available. l l
033000A203 . .(KA's) )
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ANSWER 4.11 (2.00)
Manually trip the Main turbinet.50] at the front standardt.503.
Menus 11y trip the reactor trip breakerst.503 in the Control Rod Drive Eo,vipment RoomE.50]. .
REFERENCE 4023.001.011, p. 3 No facility objective available. i 000068K318 . .(KA's) i ANSWER 4.12 (1.00) i To ensure that no rod motion can occur.
1 or To ensure. safe operation per the FSAR.
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LSA._ER95EQUBEEi _dQBd6L1 AEdQSd6LI.EdESGENGL Page 35-
'AUD RAQIQLQQIGAL CONTROL s 1 l !
l !
REFERENCE- -
! RO-C-NS2A, unnumbered page. discussing mode threefoperation with'only 1 one loop operating.
'l No facility objective available. f 003000G005- . . . ( K A ' s )'
ANSWER 4.13 (2.00)
Unit 1 Trips on loss of a reactor coolant pumpt.503.:<50% C.503 i
Unit 2 Trips on loss of a reactor coolant'pumpt.50], <29% C.503 l (Coincidence not required for full'eredit but if-siven and incorrect,.
.1 point lost.) h 1
REFERENCE l q
a 1 and 2-OHP 4022.082.001, pp. 1 &2 1 Objactive-No facility objective available.
062000A204 ...(KA's)
)
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(***** END OF CATEGORY 4 *****)
(********** END OF EXAMINATION **********)
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N, q 4pM%jy#ggg o," kreR hiisiUdifohvMonRissrDN MiNIN 4$$ } hh A r: SENIOR REACTOR DPERATOR REQUALIFICATION EXAMINATION '""
m .
FACILITY: _GQQL(_1h2________________
REACTOR TYPE: .fWR WEGS__________________
- DATE ADMINISTERED: .@Zl96fg2_____ __________
EXAMINER: _glqPARD _D,.. ,___________
CANDIDATE: _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ . .
I 195..BUQIJgyp_.lg_G6NDJ 1 L 9IE s. >
This examination replaces Read the attached instruction page caref ully. requalification examination. l the current cycle facility administered Retraining requirements f or f ailure of this examination are the same as f or f ailure of a requalification examination prepared and administered by ,
each question' are indicated in your training staff. Points for 70%
parentheses after the question. The passing grade requires at least I
Ex ami na ti on papers in each category and a firaal grade of at least GO%. -
will be picked up four (4) hours after the examination starts. )
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__ EQBEL,_ _ _M6(UEi,_ _ _ _ _ _ _ _ ___ _ _ _ _ G G I E G Q S L _ _ , _ _ _ _ _ _ _ _ _
_ _Y6WE_ .1Q106 D. THEORY OF . NUCLE AR PONER F LAN1 16 00 _2Esd2 _ _ _ . _ _ _ _ _ _ . _ _ _ . . .
O PliR A T I O N , FLU 1DS, AND THERMODYNAMICS <
l
- 6. PLANT. SYSTEMS DESIGN, CON 1ROL, 19.99.._ _2hS? .. .__._ ..___ _ . . _ .
AND INSTRUMENTATION 1
- 7. PROCEDURtiS - NORMAL, ABNDRMAL, lI _19.Zp_. 2p.00 l EMERGENCY AND RADIOLOGICAL C.ONTROL
- 8. ADMIN 19TRA11VE PROCEDURES,
_.lN Eh ..2931M _ . . . _ . _ _ _ . . _ _ . . __.__
CONDITIONS, AND LIMITATIONS S9.00 % Totals t
Final Grade I have neither given All work done on thia examination is my own. l I
< nor received aid. !
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,' Candidate's Signature 1s ;
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NRC RULES'AND.' GUIDELINES FOR LICENSE EXAMINATIONS 4 .
eDuring the administration of.this examination the following rules apply:
- 3. . Cheating on the examination'means an automatic denial of your app]ication and enuld result in more severe penaltien.
2; h%stroom trips are to be limited and only one candidate at a time. may L leave. You must avoid all contacts with anyone outside the examination-room to' avoid even-the appearance or. possibility ~'of cheating.
I3 Use ' b]ack ' ink or dark pencil only to ' f aci)Itate legib)e reproductions.
- 4. Pririt your name in the blenk provided ~ on the' cover sheet of the examination,
- b. F331 in the date on the' cover sheet of the examination tif necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of . the first page of egh section of the answer sheet,
- 8. Conceautively number each answer sheet. write "End of Category _._" as appropriate, start euch category on a new page, wri te op.ly.' 9D 2Df' Eldt .:
l' of the paper, and write "Last Page" on the last answer sheet Number each answer as to category and number, f or e>.a'mpl e . 1.4. 6.3.
9.
14 Slu p at least @ ms lines between each answer.
11 Sapsrate answer sheet.s f rom pad and place finished answer sheets f ace 1 down on your desk or table.
- 12. Use-abbreviations only if they are commonly used in facility li M thu m, j 1
- 13. The pbint ivalue f or each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 34. Show all ca).culations, met. hods, or assumptions used to obtain an answer
' d matnemat ical prob 3emo whether indicated in the question or not.
o
- 15. Part i al c'.redit may be given. Therefore,. ANSWER ALL PARTS OF THE i j
OPESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If pdrts of the examination are not clear as to intent, ask questions of the 9M mlLLer only.
7
- 17. You must sign the statement on the cover sheet that indicates that the work is your odn and you have not received or been given assistance in completing the examination. This must be done after the examination has s
been comnetod.3 8
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- 18. Whyn you complete your examination, you shall: (
- a. Assemble your examination as follows:
(1) Ex a;n questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answ+r 1.h t" eXaluj n ati011 questiODF.
<- Turn i n a l .l scrap paper and the balance of the paper that you did n ,1 o r,- f r a n tw.f r.i ec th.- que ci i nn r
- d. heave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progrecs, your licerise may be denied or revoked.
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t PAGE b& . THEORLDF _HUCLEAB_ POW R__.ELANT OPEBAI198t_Ek_MDS. AND .
THEBMUDYUAMICS i
i UDESTION b.01 (1.50)
The pjant han operated at 100% power for the last ten days. If the plant were to trip from this power level, GTATE the effect that Xenon would have Limit your answer to on shutdc>wn margin f or each of the following times.
(Consider each case separately INCREASES, DECREASES, or STAYS THE SAME.
and assume all ot her t actorc remain the same.) I
- a. b hours after trip 1
1 10 h. .o r r..
site, t r:i p
- c. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after trip i
i QUESTION 5.02 ( 76) l Afterr en 81 initiation wit h RCP's secured, what is the RCS subcoo]ing nargin (as defit:ed in 000 4023.11-0) given the following information? j RCS prensore 2216 psig i b44 degrees. F !
Core ex31 thermocouple 11 SW N Abb WORK.
QUESTION 5.03 ( .76)
Li t.t THREE different condit.ione that must be present, for Pressurized Thermal Shock (PTS) to result in reactor vessel cracking. :
I QUESTION b.04 (J.bO)
Describe hEFL,UX BOILING mode of core cooling in terms of coolant flow l path. (1.0) !
List TWu fienct or Coolant Synterr plant conditions that must be present for REFLUX BOILING to occur. (O.b) ;
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'h.L.. THEORY _QE_NELEAB_EOJfER PLAEI_QEEB6IION, FhillDS. AND
- ItjE8MQDYNAdl @
r QUESTION b.0b (2.00)
.The plant is being cooled by natural circulation following e. trip from 200% power, equilibrium Xe, EOL. Describe-HOW and WHY the parameters-listed below will change from'10 minutes after trip to 90 minutes'after trip. ( Assume that S/G pressure is being maintained 'at 1005 psig)
'a. T cold li. T hot
- c. Core delta T
- d. Loop transit time QUESTION 6.06 ( l '. 00 )
List FOUb' operational sympt oms or iridications of cavitation f or e l' centrifugal purop.
QUESTION b.07 ( .80)
J ndicat e at which t ime in core Jif e (BOC or EOC) the following accidents re!.ol t in a higher . steady state power. (Assume all control systems are in automatic. No explanation required.) Assume no trip.
- a. Main steam line break
- b. Bod withdrawal accident from low in source range.
QUESTION 5.08 (1.00)
On Unit. I during cycle 9. critical boron concentrat, ion (Cb), as measured by chemica] sample, gradually deviated f rom the predicted curve.
I
- a. Was measured Cb GREATER THAN or LESS THAN predicted boron !
concentration (3000-b500 GWD/MTC)? (0.25) .
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- b. Explain the reasons for the deviation. (0.75)
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g AND~ -PAGE 4' l l1. ,L_._.THEQBLDE__NU%EAB_EPWEB_ELANT QEEBATION. FLDIDS'. -
THEhdOUXNAdlCF ,
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I QUEST]ON -
6.09. (1,20) j A comparior>n of'. Units l ' and 2 S/G's indicated accelerated Intergranular Corrosion (IGC) in Unit 2. What are THREE different operating conditions :
in Unit 2 which contributed to this? (1.2) .]
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a QUESTION 5.10 (1.60)
Ar.wonjug a Xe 12ve reactor st#rt.op, while critica] with power'] eve 3ed o n i at 10-6' amps for critical data,.Tave 'b46 degrees F and KCS' pressure 2255 i psig, rod D-4 (control bank D). drops-to the bottom. Describe reactor. (1.5) '!
power, temperature, and pressure transients caused b'y dropped rod. i 1
QUESTION b;11 (1.50) ;
(Fer linit 1, cycle 9, Xe equilibrium) State HOW the f ollowing change f ron, BOC to EOC at. 100% power: (Limit answer to MORE NEGA7]VE, LESS NEGAT]VE, or NO CHANGE)
- a. Moderator Temperature Coefficient (MTC)
- b. Fool Temp +: rat ure Coef f icj ent (FTC) e, Duppler only power defect 00 to 100% power)
- 6. Di.f f erentisi baron worth.
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FL111DS. 6HD PAGE 5
.h __TBE0BLOF..liUCLEAB_EQL48.Ek6HI_o_EERAT.19t1 THERMQDytibdlSS QUESTIOt1 b.12 (1.50)
An ECP is calculated f or a st artup f o] lowing a reactor trip f ron.100*
power equilibrium Xenon (BOC) on Unit 2. Indicate if actual critical rod position wi.1) be HIGHEh, LOWER, or the (Treat SAME from the calculated position each case individually.)
for each of the following situations.
- a. Xenon react ivity curve for trip from 80% is used to calculate conditions to startup lb hours after the trip (P-250 is it.oicrable). l (0.b)
the calculation. (0,b)
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- c. The full length rod position critical position prior to shutdown was '
l recorded as 200 steps vice actual rod position of 220 steps prior to shutdown. (O.b) l l
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. .fi&._.PLANIRETE!JE pKElGSuC_O!ilBQk. At!D_18EIBiltiEHIATION QUESTION 6.01 (.1.60)
St at+- whather Fewet.or Vesse] Level Instrumentation System (RVLIS) will indicate higher or lower than actual f or each of the following conditions during a LOCA accident with no RCF's running. (Consider each individually.) Limit your answer to RVLIS HIGHER THAN, SAME AS, or LOWER l THAN the actual level.
- a. Rod ejectivn (break in upper head) b Aeconiu j at or injection into voided downcomer (immediately after injectivi )
l e. Rev-rse flow through core
- d. Core blockage OUESTION 6.02 (1.50) 118T three protection and/or control f unctions of MFC 254 Turbine First ,
Stage Pressure Channel (PT-506). (Do not list indication.)
OUEGTION 6.03 (2.2b) l I n t.urmedi a t e rr nee nuclear instrument detector N3b has f ailed and the level trip r. elector switch for N3b is placed in BYPASS position. For each of the folJowing condit$ons, explain WHY the IR overpower reactor trip l signal will or will not trip the reactor.
- a. A t. 100% power and 1&C removes instrument and control power fuses.
- b. At 10-8 an.ps and I&C removes instrument power fuses.
e /d b% reset or power on condenser steam dumps and 1&C removes control power fuses.
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, ._ q PAGE 7. 1
.6<..JSANT SYSTEM 1_ DESIGN. QQN_IRQL_AND INSTBT)MENI6TlQN:
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' QUESTION 0.04 -(1.50) 4 j
As sunie : Uni t 2 is at 100% power, equilibrium Xe, normal' temperature and ..
pressure. How do the following instrument failures. affect the Channel II OP delta T setpoint? (Consider each' separately) himit your answer;to
~
RAISK, .h0WER, .or REMAIN.THE-SAME, j
- s. Channel Il pressurizer pressure fails,' low to 1700 psig-(NPP-152').
- b. Channel 11 Th fails low (630 degrees F) (Loop 2).
l Chenner] ' 31 Tu ebi ne Fi ret St agn Pressore (MI C 2f.4 L f ai1 s t c mero. .j e
- d. Auctioneering high Tave circuit fails.high.
- e. Channel II Tc fails high (630 degrees F) (Loop 2). ]
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QUESTION 6.0b' (2.00) J Assume Unit 1 is operating at 30%. power with one MFF running in' automatic and feed regulator valves in auto. Describe the response of.the stean. j generator levels and feed system to the following malfunctions with no l operistor action. Include any protective actions that occur.
- a. UFC-101 (Steam Generator Header Pressure FT-bO7) fails 3cw
> (600 psig).
FPC- 2bo (Feedwater pump discharge pressure FT-bO8) fails low-1 b.
(100 psig).
)
1 QUESTION 6.06 (1.26)
On a loss of instrument air, what will be the position'of the following valves? Ascume no operator action and all valves were initially open j prior to loss of alr. B
- e. OCR-300 (Letdown Containment Isolation)
- b. QRV-301 (Letdown Back Pressure) ;
- c. QRV-2b1 (Ch9rging Flow Control)
- d. QRV-200 (RCP Seal Supply) .
- e. QRV-b1 (Pressurizer Aux Spray) '
,l (0.25 pts each)
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PAGE 8
. .F . __ PL6HTRETEdd._DESLE._QQt!IBQLAtJIllNSTRUMENT AT J QB QUESTI0t4 6.07 (1.00)
Considering the turb5ne- driv.en auxi]inry f eed pump (TDAFP):
a LIFT TWO aut.omatic start signals .or TDAFP (include coincidence).
(0.5)
- b. Where must TDAFP mechanical overspeed trip be reset? (0.25) e Two minutes after starting, how can flow retention be reset? (0.25) l QUEST 10t4 6.06 (2.00)
LIST FOUR contro]lers aval]ab]e on the Control Room Hot Shutdown Panels.
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QUESTIOl1 6.09 (1.00) l Administrative cont rv:Is have been placed on Saf ety In.iection discharge crone-t.je valv-e IMO-27b and ]MO-270. What is the purpose of thes+
j controls?
QUECTlON 6.10 (1.00)
/ !
1mmediately f ollowing a reset.or trip caused by a loss of of f-site power. l' the RO is unsuccessful in trying to comnience RCS cooldown using condenser steam dumps in steam pressure control mode. Explain the reason for condenser stear dumps not opening. (RCS at normal no load Tave, condenser vacuum of 26.0' He, power available to CRID's.)
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PAGE 9 A __ERQCED1IBES - NORMbb. ABNQBMA_1,2_EMEBGENCY_6ED BAplQL_OgI96L col 3TRQ),,
QUESTION 7.01 (1.00)
Per 1-OHP-4021.017.002 (Placing in Service the Residual Heat Removal '
System), what are the reasons f or the f ollowing precautions:
- 2. I rior t o placing RHR loup in service, the CVCS mixed bed cemineri.11cer in service shvuld be bypassed and the Chemistry j Se .mi r. n nr.t i f i ed . l I
i QUESTION 7.02 ( .50) I l
l Per 2-OHP-4021.032.001 (Start 3ng, Paralleling, Loading, and Shutting Down l the Emergency Diese] Gen e ra t.o r ( s ) ) , answer the following statements true or I false.
I a. Any t.j m" the diesel enc}nes. are started from ambient conditions, they j j
should be run for at least thirty (30) minutes unloaded. i
- b. Ideleted)
- c. 140 rm11 y . an operator mur.t be present at the diesel enginels) anytime they are being run.
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OUESTION 7.03 (1.7b)
- a. Which 3 people (by title) may approve entry into an Extreme H3r Radiation (EHR) area. (0.75) i
- b. What, twr requirements must be met f or all ent ries into an EHR area? i (1.0)
QUESTION 7.04 (1.50)
PMP 6010. RAD.001 (Radiation Protection Manual) provides exposure limits to l be used during an emergency. State the exposure limits to be used during an emergency, and to what specific situations they apply.
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. L.__EB0fEDUBES '- NORMALL ABNQBMAL _EMERGENGLAND - '
hbDJQWG.J96L COMIBQL -
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WUESTJON 7.06- -(1.00) i W
I e r . 2-OHP-4 022. 009. 001 (RHR Pump Spray. Operation), what requirements .
should.'be met:before RHk pump' spray should.be-initiated during a LOCA? 1 (1.0) . i i
OUEST10N 7.06 (1.60)
L st 'l WJ cyr g tomo c4 Auxilirary . I eed ' Pump ( AFP) Steam Bindine.
a.
(0.75)
- b. State the THREE immediate' action steps of 1-OHP-4022.056.001, Steam .)
-(0. 7 5 ) ~ l Pinding in Auxiliary Feed Pumps.
QUESTION 7.07 (2.00) 01-OHl'-4023.E-3. (Steani Generator Tube Rupture), . Step 6,
~
requires:
" Maintain AFW flow to rupture'BG until level greater than 6% (22%.for*
adver.w ecentainment) . ' .
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- a. Why wbs the 6"4 (2b. adverse containment) value chosen? (0.b) .
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- b. Explain why level should be maintained above 6% (22% edversi i containment). (1.6i l
1 QUESTION 7.06 (1.50) f Given the following status of critical safety functions (CSF) from STA while in 02-OHP-4023.E-1 (LOCA): ;
- a. Orange condition in core cooling CSF FR-C.1
- b. Yellow conditic,n in cc _ lant inventory CSF FR-I . 3 -
- c. Red condition in integrity CSF FR-P.1
- d. Orange condition in suberiticality CSF FR-S.1
- e. Red condition in containment CSF Fh Z.1 i
- r. Yellow condition in heat sink CSF FR-H.4 List t he order that these would be perf ormed if all the procedures were to be used. (1.b) ;
l
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) ;
l l
PAGE 11 7 __EBQCEDURES__ _HOBMAL_.6Bl.!QRMAL_EME8 GEN {lL6BD
. 8ADIOLQGJ Gol,JQN.IRQL QUESTIOt1 7.09 (1.00) l Du ri ng 01 -OH1'-4 0 2 3. EC A-0. 0 (hoss of A)) AC Power), the STA monitvrine Critical Safety Functions (CSF) informs you of the following status:
ORANGE CONDITION SUBCRITICALITY FR.S.)
kED CONDITIOt1 C0RE COOLING FRC.1 RED CONDITION JNTEGRITY FRP.1 Explain which, if any, of t he FRP's (Functional Restoration Procedores) 1 yi o wc u I d a rin+ -f i sit e J y inp1emont
)
QUESTION 7.10 (2.00) 02-OHF-4023.E0 (Beactor Trip or Safety ]njection) immediate action, Step 3 requires: " Verify Power to AC Emergency Bates.'
What. are the Act ion / Expect ed hespone.e substeps for this step (include designators for Unit 2)? y i
Q11EST3 ON 7.31 (3.00) l'e r en ergency procedure usare, define the following terms:
l
- a. Ruptured steam generator
- b. Fault ed r,t eam generator c Uncontrolled cooldown
- d. Int act st.eam generator
(***** END OF CATEGORY 07 *****)
, o
~a J-
. PAGE- 12_
L _.APM M EIB6TlyE__ERQGED.UBES2 ,_QQNDlIlQNS.-AND_ LIMITATIONS
)
1
. s I
i 1
-QUESTIOlJ 8:.01 (1.00) t During a! Unit 1 plant'.cooldown.and depressurication, the f ollovirig data !
J was taken:
Time' Pressure (PSI) Temperature.(Degrees F) 191b 1110 320.6 ;
l1930 1120 316.3 'I l
.104! 11 f>0 311.0 2000 1260 306:0 Using Figure 3.4-3 from Technica1' Specifications (3.4.9), specify the actions,-if any, required per Technical Specifications.
QUEST 1014 S.OE (1.60)
PM1-4010 (Plant Operations Policy) states. " Steady state reactor core power levels'shall not be exceeded." State the' specific instructions for maintaininer rnaximum pnwer level per FM1-4010.
l GUEST 10N B . 0 :- (1.60)
Per 001-4034 (Conduct'of Operations: Valve Lineups), LIST FIVE different acceptable means to verify valve position.
QUESTION '8.04 (2.00) 1 I
Technical Specification 4.0.4 specifies that operational. modes will not be changed unlese surveillance requirements' associated with the mode'being .j ent ered have been performed within the stated surveillance interval, PM -4030 defines a grace period within which all surveillance tests shall be performed,
- n. St ate the grace period intervals that a surveillance may be extended.
(0.5)
- b. L]ST the THREE conditions-per PM1-4030 (Technica1' Specification Review and Surveillance) that must be met to allow the grace period to be used for changing modes. (1.5)
(***** CATEGORY 08 CONTINUED ON NEXT FAGE 4****}
PAGE 13
, L.._N!M1tRSTB6T3 VE_EI190EDURESt..CQED3 TIQ1LS t AND LIMITATIONS OUEr. TION 8.06 (3.2b)
Per 'lechnical f specification (3.b.2), for Tavg > 350 degrees F, LIST the i F1VE conponents of an independent ECCS subsystem.
l QUESTION 8.06 (2.7b) :
Per Emergency Plan Procedures (PMP-2080):
- a. W1.t. c. hall initjally oct a r. thr- Site En erg.-ney Coordinst er (SEC) until relieved? (0.25) .
- b. List t he FOUR emergency classifications per FMP-2080, in order of l l
decreasing severity. (0,5)
- e. State the THREE actions thut may NOT be delegated by SEC. (1.5)
- d. Which two outside agencies must be notified first and in what time ,
frame (r) per PMP-2080 ETP.106 (Initial Offsite Notification)? (0.61 '
i l
I l
\
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
f ,
PAGE. 14 L ._ADMJHislRAIIVI_EROCEAUB E _CQt@lTIONS. AND_ LIM 11ATIONS q
- {
.. 1 g
4 -
)
QUESTION- 8.07.- (2.25).
ETATE f or' each of .tlw = following' whether the event' requires IMMEDIATE i (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) notification to NRC: -(Limit: answer to REQUIRED or.NOT .
REQUIRED) .
a, [deletedl
- b. -A rad 30aetive spijl causing cleanup costs of $2000 e Reactor shutdown for refueling
- d. Public Prompt Notification System (sirens). inoperable. 1 1
- e. Temporary evacuation of. building-until respiratory protection' ]
devices can be utilized. :l l
f An airborne radioactivity release that exceeds 3 times the applicable limi ts specif 3 eii in Appendix B, Table II, of 10 CFR 20, in unrestricted areas when averaged over a time period of one hour.
, -i
- g. Any deviation from Technical Specifications authorized pursuant.to- _
10 CFR 50.!,4(x). '
v a j
- h. Coolant activity of 450 uc/cc dose equivalent iodine.- >
c i Confirmed reading on core subcooling monitor' indicates O degrees F i subcoo]3ng.
.i . A tornado onsi te. ,
i l
QUEST 3ON 8.08 (1.00)
What are the bases per Technical Specifications for_. maintaining at least- 1 23 feet of water over the top of irradiated fuel assemblies seated in the 4-st. ora ge racks?
1 QUESTION 8.09 (1.00)
Per Technical Specifications, what actions are required if valves . .
- l. NMO-Ib1 (PohV block valve), NRV-152-(PORV), and NMO-153 (PORV. block valve) for Unit 2 are all. inoperable?
l.
1
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
PAGE 16 h_._ TH EQRLDE_ W ClEbB._E9EEB_J1b HT OPERA T I ON . ELU I DS . b@
Il]EBt50DLNbelc5 ANSWERS -- COOK 1&2 -87/0.6/22-SHEPARD, D.
ANSWI'R b.01 (1.60)
- n. INCREASES
- b. INCREASES
- e. DEChEASES iunit 1) or SAME (unit 2)
( 0 . 6 p t. t. each)
LE)EhENUh D. C. Cook Noelear P) ant Tech Data Book OHP-4021.001.011 192002K114 ..(KA*S)
AN.5 WEE b.02 ( .75)
RCS prersure L ; 1 b p.*.i g is 2230 psis or 6b2 +/-2 degree F
( t rom Steeni Tal le ) (0.25)
Core exit TC : 644 degrees F (0.20)
SUM 6b2 degreat F-044 degrees F : 108 +/-2 degrees F subcooled (0.25) hEFERENCE OHF 4023.E-0, Step 26 393008K11b . . ( KA' S )
AMSWER 6.03 ( .75) 1 A cooldowr, causing temperature of the resctor vessel to drop below RTNDT (Grader note: .1 if not stated that temp less than RTNDT)
- 2. Initial flaw or crack present in vessel.
- 3. Fre.ssure stress.
(0.25 pts eachi REFERENCE LP RQ C-TSG1, Objective 8 193010K10b 193010K107 . .(KA'S)
1
.tu ___THE98LOE_.HUELEAE_EREEB_EL6ELDEEMI1RIL._ ELU1DS , A89 PAGE 16 l
IHEhtVMtihMJfl5 ANSWERS'-- COOK 1&2 -87/06/22-SHEPARD, D.
i i
1 i l
ANSWER b.04 (1.60) :
Reflux Boiling is when steam exits the core and is condensed in the 1 SG tubes, wit h the resulting condensate returning to the core via the hot i J eg tc repeat the eyele. (1.0) !
This type of cooling occurs with:
1 v .i d e d . ..r- t.r .c.a t u rai r d 14CC
- 2. no reactor coolant pumps running
- 3. secondary heat sink
- 4. interruption of natural circulation (any 2@ .26 ea)
REFERENCE ,
Ll' Hn-C-MC01. Ot>jective 3 000011K101 193008K124 .. (KA'S)
ANSWER b.Ob (2.00)
- a. 'J e ren,, i n s con s t.a n t - (0.26) since it follows S/G sat urat ion pressure (0.20) l b. Th will decresce ( 0. 2b ) .
since less fission product heat is being produced (0.26) e Core delta T will decrease (0.25) since amount of decay heat it decreasing {0.26). t
- d. Loop transit time wilj increase (0.26) i since drivine head for flow is decreasing (0.25) i REFERENCE j
Thermo Text, pp 14-1b tr.> 14-29 d l 193008K122 ..(KA'S) l l
I I
1 l
i
l AND PAGE 17
^ A_ THEOFLOE.J!1C_LE6Ii_.P.QEER ELAULDEE86TlDth._EMl DL.
1HEhtM11N6ture ANSWER $ -- COOK 1&2 -87/06/22-SHEPARD, D.
ANSWEh b.06 (1.00)
- 1. Excessive noise
- 2. Excessive vibration
- 3. Fluctuations. of discharge pressure
- 4. F2 v. t ucit 1,3 . vi n> at or curr+ pt
- 6. Reduction in pump capacity
- 6. Fluctuating discharge flow
- 7. Increased pump temperature (any 4 @ .2f> es)
REFEREllCE Ll'TQ-C-TSG1, Ob.iect i ve 3 191004K101 .(hA'S) ;
J I 1 1
ANSWE), b.07 ( ,80) J i
l !
i
- a. EOC 10.40) ::l
o PAGE 18
' h ._7HE0BY_DE .H!LCLEbb..EQW_EE._EL6BI_DEEB61LON..._ FLUIDS. AND IHEhMODYH6bigS AMSWERS -- COOK 1&2
-87/06/22-SHEPARD, D. !
ANSWEh 6.08 (1.00)
- a. Measured Cb was GREATER THAN calculated Cb.
(0.25) t, . Natural boron is 19. 8% B- 10 (Exact % not required) (0.25) '
CsJ eu] at ions f or predict.ed Cb assume nominal B-10 in acid. (0.25) p 10 js gr.Jun]ly dep]..i.a1 during pow r opa rat.$ on (0.25) l 1
REFERENCE 4 1,P RQ--C-3 263, Objective 3 AEP:NRC:10012 Letter f j
000024A205 002000K507 ..(KA'S) )
ANSWEk b.09 (3 201 3 Hirher operating temperature 2 flore tripe from full power l
- 3. Higher cont aminant leve) s l l
(3 & O.4 pts each)
RE/1'hENCE L}' RQ-C-SCh1. Sho C: ective 1 '
0 3 bu l OGD0'I 0350.JKbO2 ...(KA'S)
ANSWEh 5.10 ( 1. f. i s T-mprat u re and pressure i if t ect ed by dropped rud. (0.2b pts each)
Heactor power init 181 prompt, drop (0.25) heact or power wi)) decrease ( b) and then level out at lower power level in source range as supported by subcritical multiplication. (0.26)
REFERENCE West inrhouse henctor Theory Text 19'iOODK10 3 192008K112 . (KA'S)
.i i
PAGE li-
- .__..THg0BL9F_ltucLEA8._EsgR PLAUT OPERATlG!L ELUIDS, MD IBEW@DYUMIES
-87/06/22-SHEPARD, D.
ANSWERS -- COOK 1&2 Alt 3WER b.11 (1.bO) ;
- a. MORE NEGATIVE
- b. MORE NEGATIVE
- c. LESS NEGATIVE
- d. MORE NEGATIVE t u. 376 pt s each; ;
REFERENCE !
LP RQ-C-RXT2, Objective b 192004K113 ...(KA'S) {
192004K106 i l
l ANSWEh b.12 (1.60) l
- n. HIGHER (0.b1 '
- b. HIGHER (O.b) e LOWEh (0.b) 1 hEFERENCE D C. Cool. Nuclear l'1 an t 'I + ch list s boek 2-OHl'4021.001 011 .;
0'J1000A20'l ( K A ' f'. )
i i
I l
l l
CONIgh AND_2ESIBIMENIAIIDS PAGE 20 ,
fL _ PLANLEYSIE!15_DE. EMS ,
ANSWERS,--CCOOK 1&2 -87/06/22-SHEPARD, D.
Al45WEh 6.01 (1.501
- a. HIGHEh THAN actua)
- b. LOWER THAN actual
- c. LOWER THAN actual
ANSWER 6.02 (1.60) !
l Control - stean, domp system lors of load (C-7A, C-7B) (0.5) 40,5)
Protention - F-13 torbine power input I
- Set poi nt r-terence of one Fteam flow Chhnnel f or high stear:10,5) line flow steam line isolation and safty injection (Not regt.i red to stat e which is cont rol and which le. protectic n, )
KEFEhEINE LP RQ-C-1242. Objective 1 012000K603 012000KG30 ...(KA'S)
ANSW'Eb 6.03 (2.2b)
- a. Not trip [ 0. 2!. ] since reactor trip is m6nually blocked above F-10
[ n . f.] .
- b. Not trip [0.2b] r.ince leveJ trip bypass puts control power to RPS output and inst rument power war de-energized [0. 5) .
- c. Trip [0.251 since level t rip bypass switch puts control power to RPS and control power was de-energized [0.5].
hEFERENCE DWG OP-1-98604-2 ,
OF-2-96504-2 1 LP RO-C-Nbo9 l 015000K406 ...(KA'S)
PAGE 21
. St._.1L6t!I_SYS.IEME_DESJgN . CONTROL. ANiD._lNSTRUMENTAllRS ANSWERS,-- COOK 1&2 -87/06/22-SHEPARD, D.
ANSWFF 6.04 (1 60)
- e. REMAIN THE SAME
- b. REMAIN THE SAME
- c. REMAIN THE SAME
- d. REMAIN THE SAME
- e. LOWER to.3 pts each)
REFERENCE RQ-C-1242, Objective 1 T/S Table 2.2-1 j 012000A101 ...(KA'S) l
( AMSWEh 6.Ob (2.00)
- a. Indi nat+ d de]t a P f or f eed pump increases and f eed pump speed controller decreases speed of MFP which decreases MFP discharge l pressure 10.2bl. MFP pressure decreases to .less then I actual S/G pressure, so no feed flow will go to S/G's [0.2b]. FRV's wil] open an act ual S/G 1evel decreases (0.25). R3 actor will trip q j
..n l o- ] v . teani Eerierator level or SF>FF with lo S/G 1 eve] [.25) j Indicated delta P for feed pump decreases and feed pump speed j
- b. !
controller increases sp.md of MFP which attempts to raise MFP dis. charge pressure (0.25). FRV's will close as actual S/G 1evel increases (0.261 MFP will trip on overspeed [0.2b]. MFF 1 trip wil) cause turbine trip which at 30*4 power will cause reactor i trip 10.2b]. i f
REFERENCE 1
LP RO-C-TNOb DWGF, OP-2-98bO7-2 i OP-1-98bO7-2 OP-2-98509-3 l OP 9850 9-1 Ob9000A211 . (KA'S)
~
PAGE 22.
. 62__.PLABI_.HEIEME DES 1E_.99}lIBQL, A@ 1NSTRUliENTATIQB ANSWERS -- COOK 3&2
-87/06/22-SHEPARD, D. I e
i i
AldSWER 6. 0 f' (1.25)
- a. QCR-300 fails closed
- b. ORV-301 fails open .
- c. QRV-251 fai]s open
- d. OKV-200 fails c]osed i
e OhV-51 fails closed ;
(0.2b pts each) l 1
j REFEREtJCE l
D. C. Cook P&ID's .
00006 b A20 E. 076000E302 ...(KA'S) 6.07 (1.00) l ANSWER I
- a. Undervo] ta ge on 2 /4 RCP t>uses . (0.25)
St.ean, gem. rato r lo-lo ]evel on 2/3 channels in 2/4 steam generators.
10.25)
- b. Mechanical overspeed trip device most be reset locally before turbine can be restarted from the control room. (0.25)
- c. Flow ret ention is reset when discharge valve switches are pulled to stop. (0.2b)
(grader note .06 pts for each coincidence)
Ki.:FERENCE LP h0-C- 12 41, Ob,iretives 2, b, and 7 061000K402 061000K406 061000K407 061000K411 ...(KA'S) l 1
PAGE 23
. L__J:L6HI_.EYSTEtE.DEE19ti. 09 NIB 0k_AND_JEEIBUtiENIAIIDE ANSWERS -- COOK 1&2 -87/06/22-SHEPARD D.
AM3WEk 6.08 (2.00) 3 S/G atmospheric dumps
- 2. QRV-251 (Charging flow control)
- 3. QRV-200 (RCP sea) supply)
- 4. TD AFWF gc v rrnor
.(0.b pts each)
REFERENCE LP RQ-C-1250, Ob.iective 1 000022G006 .. (KA'S) l ANSWER 6.09 (1.00) f The purpose is t.o ensure plant is not operating outside FSAR. The ECCS conf i gu ration assumed in certain accident analyses is with these valves open co that e r.e S3 pump would inject into a)) four loops. (These valves.
were ilor.ed in Unit ;- for maintenance without realicine the safety impl i .. ot i ons . )
REFERENCE LP RQ-D-12b3 LEh 33 6-086- 021 -00 SEE 2-87 Degradation of Core Cooling 00002bG007 ..(KA'S)
ANSWER 6.30 (1.00)
C-9 or st eam dump permi ssive not met [0.25); no circulating wster pumpt running I.7b)
REFERENCE LP RO-C-PG12 000001K301 ...(KA'S) a________
PAGE 24
~
- 7. PROCEDUMB - MORMAL. AleIORinL. EIGQtGENCY AND ,
RADIOLOGICAL CONTROL 1 - w87/06/22-SHEPARD, D.
ANSWERS -- COOK,1&2 3 ANSWER 7.01 (1.00)
- 1. To prevent the flashing of CCW to steam. (0.5)-
- 2. To verify that RHR operation through CVCS mixed bed demineralized will not cause a chloride excursion. .(0.5)
REFERENCE 1-OHP-4021.017.002, Precautions.4.6 and-4.11' 005000G007 ...(KA'S)
ANSWER 7.02 ( .50)
- c. False
- b. [ deleted)
- c. True (0.25 pts each)
REFERENCE 2-OHP-4021.032.001 064000A203 064000A206 ...(KA'S)
ANSWER 7.03 (1.75) :
- b. 1. Continuous coverage by a RP technician is required for all entries into an EHR area. (0.5)
- 2. RWP is required f or all entries into an EHR area'. (0.5)
REFERENCE PMP-6010. RAD.002, pg. 3 194001K103 ...(Ya'S) i 1
4
- 7. P N 9 - NORMAL. ABNORMAL. -nMHCY *n PAGE 26 l
RADIOLOGICAL CONTROL ._
1 AHSWERB -- COOK 1&2
-G7/06/22-SHEPARD, D. ,
0 ;~ _
(
i ANSWER 7.04 (1.50) 1 1
When an emergency presents serious hazards to life and health of plant I personnel or the general public or when extensive damage to equipment j which, in turn, may present a serious health hazard [0.5),; j one-in-a-lifetime dose of up to 75 Rems whole body [0.25) if imminent I danger to personnel.does not exist or there is some time for planning i
[0. 5) , a dose of 25 Rem whole body should not be exceeded [0.25). l 1
REFERENCE (
I FMP-6010. RAD.001, pg. 40-41 l 194001K103 ...(KA'S) l ANSWER 7.05 (1.00)
- 1. ECCS is operating in recirculation mode.
- 2. Containment pressure has increased to 8 psig.
i
- 3. 30 minutes after accident has occurred (any 2 @ .5 ea)
REFERENCE 2-OHP-4022.009.001 000025G010 ...(KA'S)
.7 . PRON _nURES - Emut. ABW M uL. DetR N WCY AND PAGE .26 RADIOLOGICAL OOblTROL ANSWER $ -- COOK 1&2 -87/06/22-SEEPARD, D. -
ANSWER 7.06 (1.50)
- c. 1. Temperature increase noted at AFP discharge line monitoring point.
- 2. Affected pump casing temperature high (> 185 degrees F.)
(number not required)
- 3. Erratic pump amp swings upon starting AFP caused by cavitation or other indication of cavitation.
(Any 2 @ 0.375 pts each)
- b. 1. Cycle motor-operator discharge isolation valve, monitor temperature.
, 2. Vent affected AFP.
- 3. Attempt to reseat leaking check valve or other steps that will accomplish this.
(0.25 pts each)
REFERENCE 1-OHP-4022.056.001 I
061000A204 ...(KA*S)
ANSWER 7.07 (2.00)
- a. Level was chosen to insure that water level in the ruptured SG is above the top of the U-tubes. (0.5)
- b. When the primary system is cooled in subsequent steps. SG tubes will approach the temperature of the reactor coolant, particularly if RCP's continue to run [0.5). If the steam space in the ruptured SG expands to contact these colder tubes, condensation will occur which would reduce ruptured SG pressure [0.5). This would reduce reactor coolant subcooling margin and/or increase primary-to-secondary leakage, possibly delaying S1 termination or causing S1 reinitiation (0.5].
REFERENCE 01-OHP-4023.E-3 LP RQ-C-EOP-3 Objective 3 000038A201 ...(KA'S) i
- 7. r _-=...-- . narmut. ' Anunnuar.. mannmaCY AND , - y.; 't PAGE .,37...
RADIOLOGICAL _00iiTROL . i. . ;. ,. . .t.c a7 ANSW1RS -- 000K 1&B -87/06/22-SHEPARD,D.,y'.4pi -
t s'
ii
,' ' ' +
ANSWER 7.08 (1.50)
- c. .FR-P.1 First ) RED first ,(0.2) 4 ',.
.t
- c. FR-Z.1 ) ORDER within-RED 3 (0.3) 4 s
- d. FRJ S.1 ) ORANGE:second g.to.2) ,
- a. FR-C.1- ) ORDER within ORANGE (0.3) ,
- f. FR,LH i s ) YELLOW 1ast ) (0.2)- )
b.
<FSrl.3~Last ) ORDER within YELLOW (0.3) ;
c i,
- i REFERENCE LP RO-C-E011,,0bjective 4 ,
, 'gN s*
194001G010 .. ..(KA'S) f:- !
s e i.,
l L i p , .
4 ANSWER 7.09 (1.00) 1
} , :,: '
Not implement any FRP [0.57; note in procedure says C3F status trees should be monitored for information only; FRP's should not be iv.plemented
.g l [0.5]. .
i
( , .
REFERENCE ' '
01-OHP-4023.ECA-0.0 #-
LP RO-C-E011, RO-C-E011-SH-1, pg. 12 'c 194001G010 . ..(KA'S) .g i (
1 1 -
I
, 1, a .1 t
(
1 4 1 Y
'l 4
4 h
1; l ,
4 . "i j i si e e
)
, i a '
es j 4
^' - - -
A_.. __.__._L__.____,__,,_'
~
l MEM__ERS(
B@IOLOGICAL CONTROL MN \, PAGE 28 o .
" -87/06/22-SHEPARD, D. E' ANSWElis -- 000K 102' .
l l
I>
t Al,'6W3R,,(7110, '
' (~2.00) y
'n ,V(,I < iAC emergency buses -i. AT LEAST ONE ENERGIZED: (0.5)
.. T21A
. T21B r
, or i
T21C (0.5) j
. T21D i: i s (O.5)
' b '.
AC emergency buses # ALL ENERGIZED:
T21A , l T21B
.1 and ..
. T21C
. T21D (0,5)
REFERENCE 0 2 - O H P '4 0 2 3 . E O , Step 3 000055G010 194001G011 . . . ( KA ' S ) l 1
,: ANSWER 7.11 (1.00)-
1
- a. Ruptured steam generator is a steam generator which has steam generator tube break,
- b. Faulted steam generator is a steam generator which has steam or feedwater break.
- c. Uncontrolled cooldown is cooldown not under the control of the operator and incapable of being controlled using available equipment.
[ -
- d. ' Intact: steam generator has neither tube rupture nor secondary side ll .br9ak.
- (alternate wording is acceptable) 5.
(.25 ea) .
c i REFERENCE
).I EOl' Background Writere Guide l, 0000376010 ...(KA'S)'
o +
s.
,,.. PAGE 29
/ ADMINISTRATIVE PRO-ikER. NWDITIMS. AND LYMITATIMA ANSWERS -- COOK'l&2 -87/06/22-SHEPAILD, 'D.
. x : -
x; :6 ANSWER 8.01 (1.00)
Rsstore the temperature and/or pressure to within the limit (.75) within 30 minutes. (.25) [ grader note: will accept immediately or less than 30 min]
REFERENCE Technical Specification 3.4.9 LER Unit 1 315/87005 !
010000G008 ...(KA'S)
ANSWER 8.02 (1.50)
- a. Average over any eight hour shift will not exceed the maximum 1 licensed power level.
1
- b. Excursions where power level may exceed the maximum by 2% are l permissible for periods up to 15 minutes.
- c. 102% of maximum power should not be exceeded however excursions of lower levels may be tolerated for longer periods of time ,
(1% for 30 minutes, 1/2% for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />). l l
- d. No limit on the number of excursions allowed, however, 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average ;
shall be maintained at or below maximum licensed power level.
(0.375 pts each)
REFERENCE PMI-4010, paragraph 3.9 194001G012 ...(KA'S)
I I
l o
ADMINISTRATE [VEPROCEDURES. CONDITIONS AMD LIMITATIONS , PAGE~ 30
' ANSWERS -- 000K 1&2- -87/06/22-SHEPARD, D.
l ANSWER 8.03 (1.50)
- 1. Attempt to move valve in closed direction
- 2. Stem position on a rising stem valve 3 Mechanical position indicator
- 4. Remote position indicators
- 5. System response, if other means are not available j (0.3 pts each)
REFERENCE OHI-4014, paragraph 3.4 194001K101 ...(KA'S) l l
ANSWER 8.04 (2.00)
- a. 1. Maximum allowable extension not to exceed 25% of the surveillance interval [0.25], and
- 2. Total maximura combined time f or any 3 consecutive intervals not to exceed 3.25 times the specified surveillance interval
[0.25].
- b. 1. Grace period will not expire before the surveillance can be completed in the mode being entered.
- 2. Permission to use grace period is obtained from Operations Production Supervisor or Operations Supervisor.
- 3. Use of the grace period is logged as an open item in the applicable Control Room Log until the surveillance is performed.
(0.5 pts each)
REFERENCE PMI-4030, paragraph 3.1.4 194001G005 ...(KA*S)
\
l Ne ' " %%m 8......., ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 31 ANSWERS -- COOK 1&2 -87/06/22-SHEPARD, D.
ANSWER 8.05 (1.25)
- 1. One centrifugal charging pump
- 2. One safety injection pump i
- 3. One residual heat removal heat exchanger
- 4. One residual heat removal pump i 5. A flow path capable of taking suction from RWST on a SI signal .
l and transferring suction to containment sump during recirculation l
phase of operation.
(0.25 pts each)
REFERENCE Technical Specification 3.5.2 194001G005 ...(KA'S) i l
l
- p. PAGE 32 h1NISTRATIVEPROCEDUNR. CONDITIONS. AND LIMITATIONS ANSWERS,-- COOK 1&2 -87/06/22-SHEPARD, D.
j l
ANSWER 8.06 (2.75)
- a. Shift Supervisor (0.25)
- b. General Emergency l
Site Area Emergency Alert Unusual Event (0.1 pt each for name, 0.1 pt for order)
- c. 1. Classification of the emergency 1
l 2. Directing the notification of offsite offici 1 1
- 3. Making protective action recommendations to of. site emergency management agencies.
(0.5 pts each) s
- d. Berrien County Sheriff's Department Michigan State Police Unusual Event - within 15 minutes of classification Alert, Site Area Emergency, General Emergency - immediately after classification (0.125 pts each)
REFERENCE , !
PMP-2080 EPP-101, 102, 106 194001A116 . ..(KA'S)
ANSWER 8.07 (2.25) d, g, h, i, j- required; b, c, e, f - not required (0.25 pts each)
(part a deleted]
l REFERENCE PMI-7030, Attachment 4 PMP 2080 EPP 101 10 CFR 50.72 194001G003 . ..(KA'S)
N 9
, . ~ - - _-
5-ep?" ANYMISTRATIVE PROCEDUm, nniTIf**. Aun LYMTTATI N '
- PAGE '33
.. <q y AtlSWERS -- 000K 162 -8k)D6/22-SHEPARD,D.
. , 9.y .: . , , u s .
, ;a g- ,~ ,
I l
i ANSWER B.08 (1.00) l Ensures that sufficient water depth is available to remove 99% of the cesumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. (percentages not required for full credit.) ;
REFERENCE Technical Specification Basis 3.9.11 194001G008- ...(KA'B)
'I ANSWER 8.09 (1.00)
L Within one (1) hour either:
1
- 1. Restore valves to operable status
- 2. Close and de-energize the other valve in each line and be in hot standby within next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- h. REFERENCE Technical Specification 3.4.11 194001G008 ...(KA'S) 1 l
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