IR 05000271/1987010
| ML20236W158 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 11/23/1987 |
| From: | Howe A, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236W133 | List: |
| References | |
| 50-271-87-10OL, NUDOCS 8712070306 | |
| Download: ML20236W158 (113) | |
Text
{{#Wiki_filter:. . U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N0. 87-10 (OL) FACILITY DOCKET N0. 50-271 FACILITY LICENSE NO. DPR-28 LICENSEE: Vermont Yankee Nuclear Power Corporation RD 5, Box 169 l Ferry Road ' Brattleboro, Vermont 05301 FACILITY: Vermont Yankee EXAMINATION DATES: July 14-17, 1987 and August 25-27, 1987 _ ; a: CHIEF EXAMINER: A. Howe, Senior Operations Engineer Date di APPROVED BY: David J. Lange, Chief, BWR Section Date Operations Branch, Division of Reactor Safety SUMMARY: Written examinations and operating tests were administered to one (1) senior reactor operator (SRO) and six (6) reactor operator (RO) candidates.
One (1) SRO and four (4) R0s passed the written examinations. All others failed the written examir,ations. The operating tests administered July 15-17, 1987 were invalidated due to the inadequate performance of the plant reference simulator during the conduct of tne examinations. All results of those operat- ) ing examinations have been destroyed. One (1) SRO and four (4) R0s passed the ' operating examinations administered August 25-27, 1987. All others failed the operating examinations.
8712070306 871202 PDR ADOCK 05000271 V PDR
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1' DETAILS TYPE OF-EXAMINATIONS: Replacement EXAMINATION RESULTS: i
l RO l SRO- l l Pass / Fail l Pass /Faii i l l l l 1 I I l Written .I 4/2 l 1/0 l l l' I I I I . I I l Operating l 4/2 1 1/0 l l 1 I I I I I i 10verall l 2/4 l 1/0 l l l. I i 1. CHIEF EXAMINER AT SITE: A. Howe, Senior Operations Engineer 2. OTHER EXAMINERS: T. Lumb, Operations Engineer L. Kolonauski, Operations Engineer ' R. Keller, Chief, PWR Section, DRS R. Miller, Sonalysts, Inc. -(Examiner) 1 M. Sullivan, Sonalysts, Inc. (Examiner
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in Training) 3. The following is a summary of generic strengths and deficiencies noted on the operating tests administered August 25-27, 1987. This information is being provided to aid the licensee in upgrading license and requalifica-tion training programs. No licensee response is required. . STRENGTHS a. Candidate ability to perform local emergency procedures b. Candidate use of the Alarm Response Procedures i
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DEFICIENCIES a. Candidate knowledge of refuel interlocks b. Candidate knowledge of Nuclear Instrumentation (except for the Rod Worth Minimizer) _ _ _ _ _- __ _ _. ._____-_____________________n
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- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . . 3 4. The following is a summary of generic strengths and deficiencies noted from the grading of the written examinations. This information is being provided.to aid the licensee in upgrading license and requalification training programs. No licensee response is required.
STRENGTHS a. Knowledge of reactivity coefficients - Question 1.01 b. Use of the Steam Tables - Questions 1.02 & 5.04 c. Knowledge of thermal limits - Questions 1.03 & 5.03 d. Explanation of meth ds for drywell equipment drain leak rate measurement - Question 2.05 e. Ability to' reset a Control Valve Fast Closure SCRAM - Question 4.07 f. Immediate actions for OT 3122, Loss of Normal Power - Questions 4.11 & 7.06 DEFICIENCIES a. Knowledge of xenon effects at low power - Question 1.04 l b. Effects of an increase in power on core flow at low power - , Question 1.07b c. Adverse effects of operating a recirc pump with the discharge valve shut - Question 2.01a d. RHR system response to a LPCI initiation signal while in the Shutdown Cooling mode - Questions 2.09 & 6.02 e. Effect of a loss of control air on the CRD flow control valve - Questions 2.11b & 6.06b f. APRM back panel switch positions - Question 3.01
, g. ADS valve logic and power supplies to the ADS valve solenoids -
Questions 3.02a, 6.10a, 3.02c & 6.10c l h. Effects of changes in drywell parameters on reactor water level l indication - Question 3.06 .i t 1. Causes of reactor building ventilation radiation alarms - j Questions 3.09b & 6.09b
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r. j. Operator responsibilities when entering into Emergency Operating ! Procedures (Standing Order #10) - Questions 4.04 & 8.01 '
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5. Personnel Present at Exit Interview, July 17, 1987: NRC Personnel T. Lumb, Operations Engineer l R. Keller, Chief, PWR Section, DRS 1 W. Raymond, Senior Resident Inspector ! Facility Personnel W. Murphy, Vice President and Manager of Operations J. Pelletier, Plant Manager R. Spinney, Training Manager E. Lindamood, Operations Training Supervisor R. Devercelly, Operations Training Instructor R. Grippardi, Quality Assurance Supervisor 6. Summary of NRC Comments Made at Exit Interview, July 17, 1987: Security at the written examination was lax. Several unauthorized personnel entered the_ examination area. There was no effect on the examination process because the situation was corrected by posting signs and informing outside personnel that an examination was in progress. The staff requests that this matter be corrected prior to the start of future examinations.
The NRC appreciated the organized manner in which the training staff reviewed and presented comments on the examination. Facility personnel were reminded of the requirements for submitting written comments.
The generic strengths and weaknesses noted on the operating tests were presented and the performance of the simulator (see Attachment 5, section 1) was discussed. The root cause of the problems encountered with the simulator was reported to be the unreliability of the power supply during electrical storms. The NRC requested that the facility evaluate the j need for a reliable or uninterruptible power supply for sensitive ) components.
l The results of the examinations would not be discussed at the exit meeting but would be contcined in the Examination Report. Every effort would be made to send the examination results in approximately 30 working days.
7. . Summary of Facility Comments Made at Exit Interview, July 17, 1987: The facility was concerned that the performance of the simulator would invalidate the operating examinations.
l The facility was concerned about the length of the plant walk through j portion of the operating examinations.
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8. Post Examination Discussions with Facility: Due to the performance of the plant reference simulator during the conduct of the operating examinations administered July 15-17, 1987, the results of those operating tests were invalidated. On August 3, 1987 a conference call was held between the NRC and the licensee to discuss the effects of the simulator deficiencies (see Attachment 5, section 1) and a corrective action plan. The following is a summary.of that discussion.
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The NRC was concerned about Emergency Operating Procedure (E0P) l training,.for licensed operators, conducted on the simulator. The licensee replied that E0P training was complete prior to problems with the simulator.
- The NRC was unable to make a licensing decision due to the poor simulator fidelity and performance.
- The written. examination results would stand. All candidates would be , given another operating test without prejudice at a later date. All j results of the first operating tests would be destroyed. ! 9. Personnel Present at Exit Interview, August 27, 1987: NRC Personnel A. Howe, Senior Operations Engineer D. Haverkamp, Project Engineer Facility Personnel R. J. Wanczyk, Operations Superintendent G. LeClair, Assistant Operations Supervisor R. Spinney, Training Manager A. Chesley, Simulator Supervisor E. Lindamood, Operations Training Supervisor R. Devercelly, Operations Training Instructor D. Dyer, Quality Assurance Engineer 10. Summary of NRC Comments Made at Exit Interview, August 27, 1987: There were a few problems with the simulator performance. However, the Simulator Instructor did a good job operating the simulator for the examinations.
The plant appeared to be orderly and the Control Room atmosphere was conducive to the examination process.
The generic strengths and weaknesses noted on the operating tests (see section 4 of this report) were presented, j l ;
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. 6 The Status of License Operator Requalification exams and the reorganization of Region I were discussed.
The results of the examinations would not be discussed at the exit meeting but would be contained in the Examination Report. Every effort would be made to forward the results in approximately 30 working days. - 11. Summary of Facility Comments Made at Exit Interview, Aug. 27, 1987: The facility was concerned that candidate nervousness would affect the examinations.
The facility asked if the minor problems encountered with the simulator would invalidate the examination. The Chief examiner indicated that ' Regional managtment review would be conducted to assure valid performance.
Attachments: 1. ' Written Examination and Answer Key (RC) 2. Written Examination and Answer Key (SRO) 3. Facility Comments on Written Examinations after Facility Review 4. NRC Response to Facility Comments 5. Simulation Facility Fidelity Report
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t . U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 87-10 (OL) FACILITY DOCKET NO. 50-271 FACILITY LICENSE NO. DPR -28 LICENSEE: Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road Brattleboro, Vermont 05301 FACILITY: Vermont Yankee EXAMINATION DATES: July 14-17, 1987 and August 25-27, 1987 CHIEF EXAMINER: M. b A. Howe, Senior Operdtions Engineer
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Date APPROVED BY: - W David J. Lange,VChief, BWR Section
//-2f-P P Date Operations Branch, Division of Reactor Safety SUMMARY: Written examinations and operating tests were administered to one (1) senior reactor operator (SRO) and six (6) reactor operator (RO) candidates.
One (1) SRO and four (4) R0s passed the written examinations. All others failed the written examinations. The operating tests administered July 15-17, 1987 were invalidated due to the inadequate performance of the plant reference simulator during the conduct of the examinations. All results of those operat-ing examinations have been destroyed. One (1) SRO and four (4) R0s passed the operating examinations administered August 25-27, 1987. All others failed the operating examinations.
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DETAILS TYPE OF. EXAMINATIONS: Replacement EXAMINATION'RESULTS: t I R0 l SRO l l Pass / Fail l Pass / Fail l l l l l l l l l Written i 4/2 l 1/0 l l 1 I I
.l . I I l l Operating l 4/2 l 1/0 l l 1 I I I I I I l0verall I 2/4 l 1/0 l l 1 I i 1. CHIEF EXAMINER AT SITE: A. Howe, Senior Operations Engineer 2. OTHER EXAMINERS: T. Lumb, Operations Engineer L. Kolonauski, Operations Engineer R. Keller, Chief, PWR Section, DRS R. Miller, Sonalysts, Inc. (Examiner)
M. Sullivan, Sonalysts, Inc. (Examiner in Training) 3. The following is a summary of generic strengths and deficiencies noted on the operating tests administered August 25-27, 1987. This information is being provided to aid the licensee in upgrading license and requalifica-tion training programs. No licensee response is required.
STRENGTHS a. Candidate ability to perform local emergency procedures b. Candidate use of the Alarm Response Procedures DEFICIENCIES a. Candidate knowledge of refuel interlocks b. Candidate knowledge of Nuclear Instrumentation (except for the Rod Worth Minimizer) l k Y - - - - - - - - - - - - -
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4. The following is a summary of generic strengths and deficiencies noted from the grading of the written examinations. This information is being provided to aid the licensee in upgrading license and requalification training programs.- No licensee response is required.
STRENGTHS a. Knowledge of-reactivity coefficients - Question 1.01 b. Use of the Steam Tables - Questions 1.02 & 5.04 c. Knowledge of thermal limits - Questions 1.03 & 5.03 ; d. Explanation of methods for drywell equipment drain leak rate measurement - Question 2.05 e. Ability to reset a Control Valve Fast Closure SCRAM - Question 4.07
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f. Immediate actions for OT 3122, Loss of Normal Power - Questions 4.11 & 7.06 DEFICIENCIES a. Knowledge of xenon effects at low power - Question 1.04
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b. Effects of an increase in power on core flow at low power - Question 1.07b c. Adverse effects of operating a recire pump with the discharge valve shut - Question 2.01a d. RHR system response to a LPCI initiation signal while in the Shutdown Cooling mode - Questions 2.09 & 6.02 e. Effect of a loss of control air on the CRD flow control valve - Questions 2.11b & 6.06b f. APRM back panel switch positions - Question 3.01 g. ADS valve logic and power supplies to the ADS valve solenoids - j Questions 3.02a, 6.10a, 3.02c & 6.10c j i h. Effects of changes in drywell parameters on reactor water level j indication - Question 3.06 i i 1. Causes of reactor building ventilation radiation alarms - I l Questions 3.09b & 6.09b i ! j. Operator responsibilities when entering into Emergency Operating l Procedures (Standing Order #10) - Questions 4.04 & 8.01 l l l L _ ___ __________________________o
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L Sc Psrsonnel Prssent at Exit. Interview, July 17, 1987:
'.NRC Personnel T. Lumb, Operations. Engineer R. Keller, Chief, PWR Section, DRS :W. Raymond, Senior Resident Inspector
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'W;. Murphy, .Vice President and Manager of Operations J. Pelletier,. Plant Manager
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R. Spinney,-Training Manager
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E. Lindamood,. Operations Training Supervisor R. Devercelly, Operat. ions Training Instructor
~R. Grippardi, Quality Assurance Supervisor _
6. Summary of NRC Comments Made at Exit Interview, July 17, 1987: Security at the written examination was lax. .Several unauthorized personnel' entered the examination area. There was no effect on the examination _. process because the situation was corrected by posting signs and. informing outside personnel that an examination was;in progress. The staff requests that this matter be corrected prior to the < start of' future examinations.
The NRC' appreciated the organized manner in which the training staff' reviewed andLpresented comments on the examination. Facility personnel were reminded-of the requirements for submitting written comments.
The generic strengths and weaknesses noted on the operating tests were presented and the performance of the simulator -(see Attachment 5, section - 1) was discussed. The root cause of the problems encountered with the simulator was reported to.be the unreliability of the power supply during electrical storms. The NRC requested that the facility evaluate the need for a reliable or uninterruptible power supply for sensitive components.
The results of the examinations would not be discussed at the exit meeting but would be contained in the Examination Report. Every effort would be made to' send the examination results in approximately 30 working days.
7. Summary of Facility Comments Made at Exit Interview, July 17, 1987: i The facility was concerned that the performance of the simulator would invalidate the operating examinations.
.The facility was concerned about the length of the plant walk through portion of the operating examinations.
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l l 8. Post Examination Discussions with Facility: l Due to the performance of the plant refe'rence simulator during the conduct of the operating examinations administered July 15-17, 1987, the results of those operating tests were invalidated. On August 3, 1987 a conference call was held between the NRC and the licensee to discuss the effects of the simulator deficiencies (see Attachment 5, section 1) and a corrective action plan. The following is a summary of that discussion.
- ' The NRC was concerned about Emergency Operating Procedure (EOP) training, for licensed operators, conducted on the simulator. The licensee replied that E0P training was complete prior to problems with the simulator.
- The NRC was unable to make a licensing decision due to the poor simulator fidelity and performance.
- The written examination results would stand. All candidates would be given another operating test without prejudice at a later date. All results of the first operating tests would be destroyed.
9. Personnel Present at Exit Interview, August 27, 1987: NRC Personnel A. Howe, Senior Operations Engineer D. Haverkamp, Project Engineer Facility Personnel R. 'J_. Wanczyk, Operations Superintendent G. LeClair, Assistant Operations Supervisor i R. Spinney, Training Manager ( A. Chesley, Simulator Supervisor E. Lindamood, Operations Training Supervisor R. Devercelly, Operations Training Instructor 1 0. Dyer, Quality Assurance Engineer J 10. Summary of NRC Comments Made at Exit Interview, August 27, 1987: There were a few problems with the simulator performance. However, the Simulator Instructor did a good job operating the simulator for the i examinations. j The plant appeared to be orderly and the Control Room atmosphere was conducive to the examination process.
The generic strengths and weaknesses noted on the operating tests (see section 4 of this report) were presented.
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( l The Status of License Operator Requalification exams and the reorganization ]' of Region I were discussed.
~ The results of the examinations would not be discussed at the exit meeting j but would be contained in the Examination Report. Every effort would be q made to forward the results in approximately 30 working days.
11. Summary of Facility Comments Made at Exit Interview, Aug. 27, 1987: ' The facility was concerned that candidate nervousness would affect the examinations.
" The facility asked if the minor problems encountered with the simulator would invalidate the examination. The Chief examiner indicated that Regional management review would be conducted to assure valid performance.
Attachments: 1.. Written Examination and Answer Key (RO) 2. Written Examination and Answer Key (SRO) 3. Facility Comments on Written Examinations after Facility Review 4. NRC Response to Facility Comments 5. Simulation Facility Fidelity Report
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _ygRMgNT_YANKEg__________ REACTOR TYPE: _gWR-Gg4_________________ DATE ADMINISTERED: _pZfgZf14________________ EXAMINER: _LUMg2_T.________________ CANDIDATE: _ _________________ , JNpl8UCJJgNj,JQ_C@NDJD9][1 Uso separate paper for the answers. Write answers on one side only.
Stcple question sheet on top of the answer sheets. Points for each qu2stion are indicated in parentheses af ter the question. The passing I grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up si x - (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY __ye6UE_ _lgl@L ___@Cg6E___ _y@LUE__ ______________C@lE@g6Y_____________ zs.ss - 29199-- .3@fgs p.,.___________ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.59 r9,3 _29 99-- ~3&r2@ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 23.50 12.YL
.3@1~9@ .M _ _________ ________ 3. INSTRUMENTS AND CONTROLS tr.39 p. ,
29299-- .2@fpgr r* ___________ ________ 4 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 96.9 Jg_ ________% Totals Final ___________ Grade All work done on this examination is my own. I have neither given nor received aid.
t ----------------------------------- l Candidate's Signatur e Quf& 3. t o c~ de W . 5 v7
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. NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS f. .During the administration of this examination the f ollowing rules apply:
1. Cheating on the examination means an automatic denial of your application ! .and could result in more severe penalties.
' 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination L room to avoid even the appearance or possibility of cheating.
i 3. Use black ink or dark pencil gnly to f acilitate legible reproductions.
4 Print your name in the- bl ank provided on the cover sheet of the ex ami nat i on .
5.. Fill in the date on the cover sheet of the examination (i f necessary).
'6. Use only the paper provided for answers.
7. Print your name in the upper right-hand corner of the first page of each nection of the answer sheet. ! B .- Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side ; of the paper, and write "Last Page" on the last answer sheet. ' 9, Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table. ; 12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the 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. Par ti al credit may be given. Therefore, ANSWER ALL PARTS OF THE { QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. ]
~Il 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
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E10. When'you complete your. examination, you'shalls c. Assemble your' examination as follows:
(1) Exam questions on top.
(2) _ Exam aids - figures, tables, etc.
(3) Answer pages includingEfigures which are part of the answer, b.- Turn in your copy of the examination and-all pages used to answer the examination questions.
c. ' Turn in all scrap paper.and the balance of the. paper that you did
.not use for' answering the questions.
.d. Leave the examination area, as defined by the examiner. If'after leaving, you are found in this area while the examination is still in progress,' your license may be denied.or. revoked.
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l- - ! , l l 1:__E5]NCIEbgS_gf_N9C6598_EgyEB_EL9NI_9EEB9]JgN 1 PAGE 2 l . THERMODYNAMICS2 _ HEAT _ TRANSFER _ANp_FLylp_FL99 !
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QUESTION 1.01 (2.50) For each of the f ollowing events, state which COEFFICIENT of rcactivity would act FIRST to change reactivity, c. Control rod drop at power (0.5) b. SRV opening at power (0,5) c. Loss of shutdown cooling when shutdown (0.5) d. One recirc pump trips while at 50% power (0.5) O. Loss of one feedwater heater at 100% power (extraction steam isolated) (0.5) DUESTION 1.02 (2.00) During your Shift, a r el i ef valve inadvertently opens. The reactor is at 100% power and 1000 psig. Use a Mollier Diagram or the Steam Tables to answer the following: a. STATE the tailpipe temperature, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurization. (0.5) b. If the Suppression Pool Pressure were to INCREASE, STATE whether the Tailpipe Temperature would INCREASE, DECREASE, or REMAIN THE SAME. (0.5) c. If the reactor starts to depressurize when the valve is opened, STATE whether the Tailpipe Temperature will INITIALLY INCREASE, DECREASE, or REMAIN THE SAME. (0.5) d. STATE the Reactor Pressure at which the Tailpipe Temperature would be at its MAXIMUM value (during the depressurization). (0.5)
(ASSUME A SATURATED SYSTEM AND INSTANTANEOUS HEAT TRANSFER)
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. QUESTION 1.03 (3.00) Mntch each of the lettered items with one of the numbered items.
(Numbered items may be used more than once or not at all as appropriate.) (3.0) 1. FRP 5. PCIOMR 2. APLHGR 6. APF 3. CPR 7. TPF 4. GEXL 8. LHGR _____a. Parameter by which plastic strain and deformation are limited to less than 1%. _____b. Ratio of bundle power required to produce onset of transition boiling somewhere in the bundle to actual bundle power.
_____c. Parameter by which peak cl ad temperature i s maintained less than 2200 degrees F during postulated design basis accident.
_____d. Contains guidelines restricting power ramp rates above the threshold power.
_____e. Parameter by which fuel failure involving Pellet Clad Interaction is prevented.
_____f. Limit changes over core life due to changes in the fuel's ability to transfer heat.
QUESTION 1.04 (1.00) The reactor trips from full power, equilibrium XENON condi ti ons. Twenty-four (24) hours later the reactor is brought critical and power level is maintained on range 5 of the IRMs for several hours. Which of the f oll owing etatements is CORRECT. (1.0) O. Rods will have to be withdrawn due to XENON build-in.
b. Rods will have to be rapidly inserted since the critical reattor will cause a high rate of XENON burnout, c. Rods will have to be inserted since XENON will closely follows its normal decay rate.
d. Rods will approximately remain as is as the XENON estab-li shes i ts equi li bri um val ue f or this power level.
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:ISE8MQQYN@MlCSt _dE@l_lg@NSEE6_@NQ_E(Ulp_E(QW . ' QUESTION 1.05 (2.50)
l The reactor is brought critical at 40% on IRM range 2 with the chortest permissible stable positive period allowed by OP 0100, REACTOR STARTUP TO CRITICALITY. Heating power is determined to be 40% on range B of IRM's. SHOW ALL WORK.
c. What is the doubling time if the period remains constant? (1.0) b. How long will it take for power to reach the' point of adding heat if the period remains constant? (1.5) QUESTION 1.06 (1.50) The radiation level in the vicinity of the MSIV'S is normally very high due.to the radioactivity caused by one particular isotope.
a) What is this isotope? (0.50) b) How is it produced? (0.50) c) Why is it not present in any significant quantities in the OffGas holdup piping? (0.50) QUESTION 1.07 (2.00) a. The reactor is operating at 100% power and flow. Explain WHAT happens to core flow (INCREASE, DECREASE, or NO CHANGE) and WHY it happens, with a reduction in power by control rod insertion.
Assume recirculation pump speed remains constant. (1.0) b. At low power c on di t i on s , an increase in reactor power by control rod withdrawal will (INCREASE, DECREASE, or NOT CHANGE) flow through the core. Choose the correct answer and BRIEFLY explain your choice. Assume recirculation pumps are running. (1,0) M peer e fN9RO > clots %rel. , h p N da_
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Lt. ' PRINCIPLES OF NUCLEAR POWER PLANT OPER4TIONi PAGE. 5
. IHERDODyNADJCS1 , HEAT TRANSFER AND FLUID FLOW . >n s s v.
N QUESTION 1.08 (2.50) N a. What is meant by the term BETA with regard to eactor Theory? (0.5)
'b. How does an increase in BETA affect the reactor's response to
. the same reactivity addition ( i'. e. response is f aster, slower, the same)? Explain your answer. (1.0)
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c. From BOL to EOL does the core' average value for BETA INCREASE, DECREASE or REMAIN THE SAME?- Explain your answer. (1.0)
. -QUESTION 1.09 (3.00)
Following an AUTO INITIATION of HPCI at a reactor pressure of 800 PSIG, racctor pressure decreases.to 400 PSIG. HOW are the following parameters offected (INCREASES, DECREASES, REMAINS CONSTANT) by the change in. reactor
'prossure? BRIEFLY EXPLAIN your choice.
c', ASSUME the HPCI System is operating in automatic a's b e'si gn ed .
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't a. HPCI pump discherge head (assuming NPSH repaths O'
constant) . (1.5)
'b. HPCI turbine RPM. 3< (1.5)
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J1..-PRINCIPLES OF NUCLEAR POWER' PLANT OPNsAT70N t 'PAGE 6
. THERMODYNAMICS t _ HEAT TRANSFER AND FLUID FLOM , .
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tQUESTION 1.10 . . ' ' (2. 00)
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A ' reactor! heat y% +a e was performed (by hand)'during the 00-08 thift due~to'the P.ocah.s omputer being OOC. ..The gain acjustment factors w2re computed /Pbut- the ' APRM gain adjustments have r'ot been made.
y: a. .TRUE or.F4LSET If the feedwater flow rate used in the heat balance Calculation was LOWER than the actual feedwater flow rate, then the' actual power is HIGHER than the currently calculated power. (0,5)- j
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b. TRUE or' FALSE? If the-reactor recirculation pump, heat input used in the. heat balance calculation was OMITTED/ then the actual power A s LOWER than the currently calcupted power. (0.5)
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c. TRUE or F A*.SE ? If the steam flow'used'tn.therheat . balance calculation was LOWER than the actual steam M.ow, then the
. actual power is LOWER than the current J y calm' at2d ppper. . (0. 5)
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d. J. RUE ordALSE? If the F thi; return temperatur/ used In the hesit* balance calcul ati on was LQWErT than the actuel RWCU re' tarn temperatw e, then t'hegadtual power is HIGHER than
. the current 19 cal cul at ed Ipower. (0,5)
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+ QUESTION)#1.11 I (3.00) I g Figur[4 1, 2 and 3 (attached) contain charts of several key reactor perpmeters following a Feedwater Cont r ef a l er Failure to Maximum Demand.
For thf areas marked, give the cause of each parameter change as otated below. Initial reactor power is 100%. All automatic systems f unction normally and no operator action is taken.
,/ A- State WHY reactor power rises then sharply DECREASES. (0.75) +j w
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B- State WHY feedwater flow DECREASES s$arj; y. (0.75) e l
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C- State WHY core flow DECREASES. 1,
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, D- Explain the variations in steam flow. ,\ (0.75)
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2 c__P(@N1_ DESIGN,[NC(h2[NG_S@[EI1_@ND_EDERGENCy_Sy@ LEOS PAGE 7
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l QUESTION 2.01 (1.50) l Concerping the Recirculation System, STATE the adverse ef f ects, if cny, that operating under the f ollowing conditions could cause.
Consider each set of condi tions separatel y, i c. Recire Pump A weefstArted at 12:50. The discharge valve (RV-53A) was opened at 13:09. (0.5) 6. Recirc Pump B was i s.olated at 10:10. Seal purge to Recirc Pump B was i r,ol ated at 10:12. (0.5) c. Recit'c Pump B is operating at minimum speed with a suction presstare of 275 psig. (0.5)
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QUESTION 2.02 (1.50) During a reactor SCRAN, ErPLAIN how a control rod would respond if i ts HCU scram inlet valve sticks shut with the scram outlet valve opent c. If the SCRAM r.ccurs wi th reactor pressure at 1000 psig? (0.75) b. ' If the SCRAM occurs with reactor pressure at 300 paig? (0.75) GL iTION 2.03 (1.SO) Vermont Yankee is operating at 100% power when the STATOR CLG RUNBACK annunciator is received in the Control Room. j c. WHAT three (3) parameters could have caused this al arm? (Setpoints not required.) (0.6) )
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b. WHAT automatic action (s) occur when this alarm is r eceived? (0.5) c. A few minutes later, the STATOR CLG TRIP annunciator is ,
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recieved. WHAT automatic action (s) occur when thi s al arm is received? (0.4) l l
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, ._ ' QUESTION l2.04 (3.00)
Match the appropriate source of cooling to the load: (3.0) 3 LOADS SOURCES o. Containment Air Compressor 1. RBCCW b. Circulating Water Pump Seals 2. TBCCW c. RWCU Non-regenerative Heat Exchanger 3. Service Water d. Fuel Pool Cooling Heat Exchangers 4. Circulating Water I p. RBCCW Heat Exchangers f. Circulating Water Booster Pump Seals g. Turbine Lube Dil Cool er s h. Condensate Pump Motor Bearing Cooler 1.' Residual Heat Removal Pump Seal s J. Main Condenser k' . Stator Coolers-1. Reactor Feed Pump Seal Coolers QUESTION '2.05 (2.00) The drywell equipment drains are equipped with a leakage rate clarm system. Describe two (2) methods by which this system can determine that excessive leakage is occurring (include applicable instrumentation in your description). (2.0)
QUESTION 2.06 (1.50) i LIST six (6) CRP 9-5 indications that the Standby Li quid Control System is running and injecting into the reactor vessel. (1.5)
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QUESTION '2.07 (3.00)
.The RCIC' System started on an automatic initiation signal 5 minutes ago.
For~each of'the following conditions, state whether the RCIC system WILLLor WILL NOT continue to operate. If it WILL continue to operate, will l there be any adverse ef f ects f rom RCIC operation under the conditions? If it WILL NOT continue to operate, WHY NOT7 Consider each condition
:ssparately. Assumo no operator action.
o.- The condensate pump for the barometric condenser fails causing a high level in the barometric condenser. (1.0) b. The piping of the Stay Fill System ruptures upstream of the check valve in the line to the RCIC system. (1.0) c. The RCIC lube oil pump f ails causing . oil pressure to drop to 1 psig. (1.0) QUESTION 2.08 ~(2.50) a- STATE the normal and alternate power supplies to the Reactor Protection System (RPS). (1.0) b. WHAT' automatic actions could occur on a manual transfer of RPS bus A from its normal to alternate power supply? List three (3). (1.5) QUESTION 2.09 (3.00) The Residual Heat Removal (RHR) System loop A is operating in the Shutdown Cooling mode with full flow through the heat exchanger.
A valid LPCI initiation signal occurs. STATE the final pump and valve alignments f or the RHR and RHR Service Water systems (LOOP A ONLY) as a result of thi s event. Consider affected flowpaths only.
BE SPECIFIC. A diagram of the RHR system is attached (Figure 4) for your reference. ASSUME NO OPERATOR ACTION. (3.0) l I
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QUESTION 2.10 (3.00)
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For each of the f ollowing situations, determine whether or not the activity can occur. For those activities which cannot occur, state the condition (s) which must change to allow it : to occur. l< o. Refuel bridge is over the vessel and in motion toward the fuel pool wi th the fuel grapple loaded. All rods are inserted. Mode switch is selected from SHUTDOWN to STARTUP. Will the bridge continue to move 7 (1.0) b. Refuel platform near the core. The frame mounted hoist is loaded. One rod i s at posi tion 30. Can the load on the hoist be lowered into the vessel 7 (1.0) c. Ref ueling pl atf orm over the core. Mode switch in REFUEL.
Grapple fully lowered and unloaded. Can a control rod be wi thdrawn ? (1.0) QUESTION 2.11 (2.50) h& u 2 7 -M .un M* opwb'q a tec% pow.
a. HOW will vessel level respond (DECREASE, NO CHANGE, INCREASE) if recire flow is slowly reduced using the recirc flow master controller and control air pressure is lost to the Feedwater Control Valves? WHY7 ( 1. 5 ) . b. HOW will cooling water flow respond (DECREASE, NO CHANGE, INCREASE) if control air pressure is lost to the CRD Flow Control valve? WHY7 (1.0)
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L3a. lN@l6UMENI@_AND_ CON 160L@ PAGE 11
. 'N'. l QUESTION 3.01 ( 3. 00 )'
O. For each condition given below STATE whether the affected APRM channel.will have an INOP trip? i 1) APRM channel 'E' has 9 LPRM inputs, channel not bypassed (0.5) 2) APRM! channel 'C' -(with function switch in COUNT) meter indicates 50%, channel not bypassed- (0.5) 3) APRM channel 'D' Operate-Calibrate switch in POWER, channel not bypassed ( 0, 5 )' 4) APRM channel 'A' Operate-Calibrate switch in ZERO, channel bypassed (0.5) b. What AUTOMATIC ACTIONS occur.on an APRM INOP trip? 11.0)
:)UESTION 3.02 (2.50)
For the f ollowing situations, state whether the Automatic Depres-nurization System (ADS) relief valves will OPEN, CLOSE or REMAIN AS 15. ! Consider each set of condi ti ons separatel y. l e.< ADS initiating signal seal ed in, ADS valves open . . . . reactor water level then rises to 177 inches. (0.5) b. ADS initi ating signal sealed in, ADS valves open . . . . ADS timer reset buttons are then depressed. (0.5) ' c. ADS initiating signal seal ed in, ADS valves open . . . . then a DC power. failure occurs that affects all busses supplying ADS ; valves. (0.5) l d. ADS initiating parameters present, a loss of the containment air system supply to the drywell has occurred, 120 second timer timing out . . . . then the 120 second timer times out. (0.5) e. ADS initiating perameters present, all ECCS pumps are secured except for CS pump B which is running with a discharge pressure of 195 psig, 120 second timer timing out . . . . then the 120 second timer times out. (0.5)
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QUESTION 3.03 (2.50) In order to reset a GROUP a isolation, certain valve control switches must be repositioned. Briefly describe:
- WHAT switches must be repositioned, - to WHAT posi ti on, - the reason WHY thi s i s done, - and WHY only certain Group I valve control switches require repositioning? (2.5)
l ! QUESTION 3.04 (1.50) State whether each of the following statements concerning the Rod , Block Monitor!.ng System is TRUE or FALSE. If the statement is FALSE, l state WHY it i s f alse.
a. The RBM system is bypassed when an edge rod is selected because an edge rod cannot add sufficient reactivity to approach a thermal l i mi t . (0,5) b. The RBM count circuit determines the number of operable LPRM inputs by counting the number of assigned LPRMs with their function switch in operate. (0.5) c. The purpose of the Rod Block Monitor is to detect a control rod out-of-sequence above the range of the Rod Worth Minimizer. (0.5) QUESTION 3.05 (3.00) The reactor is operating at 100*/. power with r ec i r cul ati on flow control in master manual. Explain HOW and WHY the recirculation pumps respond to the f oll owi ng condi ti ons. Where applicable, provide specific values.
a. Master Controller output fails LOW (1.0) 6. Full open indication on recirculation pump A discharge valve is lost at the valve. (Assume bypass valves are open.) (1.0) c. The f eedwater flow signal to the recirt system f ail s to z ero. (1.0)
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' QUESTION 3.06' (1.50) 1
For. each of - the : f oll owi ng parameter changes and operational conditions, f
.otate whether :the INDICATED vessel level wi11' INCREASE, DECREASE, or i LREMAIN THE'SAME forithe specified level instrument. ACTUAL' vessel ;1ovel; REMAINS.THE SAME.
c'. The Drywell temperature increases about 150^F. How will the ROSEMONT level instrumentation respond? (0.5) b.. A reactor startup is in progress. The head vent has been closed.
Vessel temperature and pressure are increased from atmospheric and 190^F to 800 psig and 51B^F. How will the NARROW' RANGE GEMAC leve11 instrumentation respond? (0. 5 );
.c. The' reactor is shutdown and a cool down is in progress. Shutdown Cooling i s ini ti ated. How will the WIDE RANGE GEMAC level-instrumentation' respond?. - (0. 5 ) -QUESTION 3.07 (2.50)
With.the reactor. operating at 100 percent power under: steady state conditions in three-element control, an instrument technician mistakenly isolates-and equalizes the pressure across one of the Fcedwater-flow transmitters which inputs to.the Feedwater Control
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.D2ecribe the RESPONSE of the.Feedwater Control System (until steady.
-otate conditions are established) and WHY the response occurs. Assume rua operator intervention. .( 2.5) q l i QUESTION 3.08 (2.50)
.A Loss of Normal Power (LNP) signal is present for Essential Bus 4 and . reactor vessel water l evel is 80" with reactor pressure at 325 psig.
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' .What five (5) ACTIONS (load sequencing actions) will occur after the ,
associated diesel generator is supplying power to the bus and load i shsdding is~ complete. Include any associated time delays. (2.5) ) l
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QUESTION 3.09- (3.00) For each of the f ollowing annunciators associated with Process Rcdiation Monitoring, STATE ALL signals that will cause the alarm cnd WHAT' automatic actions will occur, if any. Include setpoints cnd any associated time delays.
a. ADG SYS OUT RAD TIMER START (1.5)
.b. RX BLD VENT CH A RAD HI and RX BLD VENT CH B RAD HI (1.5)
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. l. 5 1 QUESTION 3.10 (JAtrf po {
Varmont Yankee is operating at 100% power with the following Mechanical Hydraulic Control (MHC) . System setpoints:
'EPR setpoint - 965 psig MPR'setpoint - 975 psig Bypass Opening Jack setpoint - 0%
Spsed/ Load. Changer demand - 100% Gewhe t mad - ioco.
Briefly describe the MHC system response to each of the f ollowing i cctions. Consider each action separately. Include in your discussion I any changes in reactor power, control val ve . posi ti on or bypass valve position and WHY these changes occur. Figures 5 and 6 are attached for your reference.
~0 %.e.s %. n %M. r & y-r-v1 ypass Opening Jack demands 10% on the bypass valves. (1.5) i b. Recirculation flow is decreased, causing r eactor power to decrease to 90%. (1.5) i (***** END OF CATEGORY 03 *****)
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4 .' PROCEDURES - NORMALt_ABNQRMALt_ EMERGENCY _AND PAGE 15
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68919L9GICeL_CQN16QL . QUESTION 4.01 (2.00) For each of the functions described below, MATCH it with the appropriate tag that would be used. (2.0) FUNCTION TAGS a. Provide a reminder that special consideration 1. White tag is required before operation 2. White tag with b. Indicate the presence of an installed ground red border c. Provide visual indication that operation is 3. Yel l ow tag with not al l owed for the protection of personnel black lettering or equipment or necessary to maintain system integrity 4. Orange tag with black lettering d. Provide visual indication that operation is not allowed except by the Local Permissive 5. Red tag with Test Person (LPTP) or a person he directly black lettering asks to operate it for him
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QUESTION 4.02 (2.00) OE 3102, RPV LEVEL CONTROL, directs the operator to steam cool the reactor if all possible injection into the reactor becomes unavailable and level drops below -87. This is accomplished by opening one SRV. If during steam cooling, RPV pressure drops below 700 psig, the procedure directs the operator to emergency depressurize.
WHAT is significant about 700 psig and WHY is RPV depressurization required below 700 psig. (2.0) QUESTION 4.03 (3.00) During a tour of the plant you notice that Secondary Containment Integrity is not established.
a. What five (5) plant conditions must exist to ensure that Technical Specification requirements are met without Secondary Containment Integrity? (1.5) b. What three (3) conditions must be met to establish Secondary Containment Integrity? (1.5)
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QUESTION 4.04 (2.00) Stcnding Order #10, dated 10/15/85, gives guidelines for operator responsibilities during accidents when EOP's are used. What are the duties of the CRO and the ACRO until reassignments are made by the Shift Supervisor? (2.0) QUESTION 4.05 (2.00) a. What are the three (3) basic objectives of OP 3126, SHUTDOWN USING ALTERNATE SHUTDOWN METHODS? (1.5)
b. TRUE or FALSE? Engaging RCIC, RHR or the DG transfer switches will cause a loss of the control f unction in the Control Room but will not affect the automatic functions or system interlocks. (0.5)
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QUESTION 4.06 (3.00) a. Define high radiation area. (1.0) b. Requirements in Technical Speci fications regarding entry into high radi ation areas state "any individual or group of individuals p er mi t t ed to enter such areas shall be provided with one or more of the following .... . " Give two (2) items which fulfill the requirements (other than requiring a RWP). (2.0) QUESTION 4.07 (1.00) In accordance with OP 0101, REACTOR AND GENERATION SYSTEMS HEATUP TO LOW POWER, the Control Valve Fast Closure Scram is reset after the turbine is brought up to speed before synchronizing to the grid (before 30% power), m. WHAT actions are required to reset the scram in the Control Room? (0.5) b. Why is thi s step perf ormed after the turbine is at rated speed? (0.5)
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QUESTION 4.08 (1.50) OP 2115, PRIMARY CONTAINMENT, and MOD Directive 79-4, Rev. 2 allow bypassing the interlock, which keeps the containment vent and purge volves shut.while in the RUN mode, for several reasons.
Give three (3) of these reasons. (1.5) QUESTION 4.09 (2.00) STATE which Emergency Classification is appropriate f or each of the f ollowing definitions.
'c, Events are in progress or have occurred which involve actual or potential substantial degradation of plant safety margins and could af f ect on-si te personnel, could require of f-site impact assessment, but are niat likely to require off-site public pro-tection action. (0,5) b. Events are in progress or have occurred-which involve potential degradation of plant safety margins, which are not l i kel y to affect personnel on-site or the public off-site or result in radioactive releases requiring off-site monitoring. (0,5) c. Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with the poten-tial for loss of containment i n t r y,r i t y . (0.5) d. Events are in progress or have occurred which involve an actual or likely major failure of. plant functions needed for protection of the public. (0.5) i (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) _ _ = _ - __-
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I 4. PAGE. 18-( PROCEDURES - NORMAL _tABNORMAL _tEMERGENCY _AND
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l QUESTION 4.10 (1.50) 3
.Fcr.sach of the f ollowing situations STATE which, if any, Emergency Procedure (s) you would enter. If none, state NONE.
< c. Scram Discharge Volume Level - 24 gallons, Power steady at 80% (0.3) b.-Reactor Vessel Water Level cannot be maintained above 100" (0.3) c. Drywell Pressure - 2.6 psig (0.3) d. Drywell RRU Average Return Air Temperature - 150^F, Torus Water Temperature - 10B^F (0.3) o. Torus Level - 68,750 cubic feet (0.3) QUESTION 4.11- (3.00) The Unit -has had a loss of normal power and one diesel generator has failed to start. What are all of your immediate actions per DT 3122, LOSS OF' NORMAL POWER? Be specific. (3.0) a l QUESTION 4.12 (2.00) State whether each of the following sets of plant operating conditions would be likely to cause thermal stratification (per OP 2110, REACTOR RECIRC SYSTEM). Consider each set of conditions separately.
'a. 0% power, IMMEDIATELY following a reactor scram during which both recire pumps tripped, startup of both recirc pumps is in progress. (0.5)
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b. 20% power, both recirc pumps running at mini mum speed. (0.5) c. 5% power, plant heatup in progress, one retire pump running { at minimum speed. (0.5) d. 65% power, one recirc pump running at 50% rated speed. (0.5)
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1 '. PRINCIPLES OF NUCLEAR POWER' PLANT OPERATION t PAGE '19 . E . THERMODYNAMICS t HEAT TRANSFER AND FLUID FLOW ]
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-.. ANSWERS -- VERMONT 1 YANKEE -87/07/14-LUMB, T.
ANSWER- 1.01 (2.50) c. DopplerHor fuel temperature
.b.. Void c. Moderator temperature d.' Void O.JModerator temperature-(0.5.each) (2.5).
REFERENCE VY LOT-02-207, Operati onal Characteristics (Theory), pgs 23, 34,.L 39 CRO Student Objective 4 K/A~292004 K1.14 (3.3/3.3)
. -292OO4K114 ...(KA'S)
ANSWER 1.02 (2.00) a. 295 deg F (+- 15 deg F) (0,5) b. : Increase (0.5)
.c . Increase (0.5)
d. 450 psia (+- 50 psia) (0,5) REFERENCE VY LOT-02-104, Mollier Diagram, pgs 4 & 5 CRO Student Objective 1 VY LOT-02-105, Steam Tables, pgs 6 & 10 CRO' Student Object,ive 2 K/A 293003 K1.23 (2.8/3.1) 293OO3K123 ...(KA'S)
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*' .- -q '1. ' PRINCIPLES OF NUCLEAR' POWER PLANT OPERATION t PAGE 20 * .lhERdODYNAMICS t HEAT TRANSFER AND FLUID FLOW .
ANSWERS --' VERMONT YANKEE -{ /07/14-LUMB,.T.
' ANSWER 1.'03 (3.00) c. O b.. 3 c. 2 d. 5 O. B sr 5 f. 2 (0.5 each) REFERENCE VY LOT-02-312, Thermal. Limits, pgs 19, 36, 40 L 41 CRO Student Objectives 1, 10, 11 & 12 VY LOT-02-405, PCIOMR, p gs B e S CRO Student. Objective 1 K/A 293009 K1.OB (3.0/3.4), K/A 293009 K1.06 (3.4/3.8), K/A'293OO9 K1.10.(3.3/3.7), K/A 293009 K1.18 (3.2/3.7), K/A 293OO9.K1.36 (2.8/3.4) 293OO9K106 :293OO9K108 293OO9K110 293OO9K118 293OO9K136
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ANSWER 1.04 (1.00) C.
REFERENCE VY LOT-02-018, Xenon, pgs B & 9 CRO Student Objective 6 K/A 292006 K.1.07 (3.2/3.2)
- 292OO6K107 ...(KA'S)
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IMEBDQQYN@dlCSt_dE@l_I6@NSEE6_@NQ_ELQ1Q_ELQW L- ' ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
ANSWER 1.05 (2.50) c. From OP 0100, shortest permissible stable period equals 30 sec. (0.5)
.Thus Doubling time equals 30/1.44 = 20.8 seconds. (0.5)
b. 40% range 2 is equal to 0.04% on range 8 P(0) = 0.04 P(t) = 40 Period = 30 seconds P(t) = P(O) e ^(t/ period) 40- = 0.04 e ^(t/30 sec) Ti me = ^ 6^ _ ' ::: .'r r- '
-i . OC. 5 :::-fer1 207 1 * *" " '(1.5) (NOTE: if_ incorrect period is used grading will be based on method)
REFERENCE VY LOT-02-010, Reactor Period, pgs 6 & 7 CRO Student Objective 1 K/A 292003 K1.OB (2.7/2.8) 292OO3K108 . . . (K A' S) ANSWER 1.06 (1.50) c) Nitrogen 16 ( 0. 3 ) b) produced by the activation of oxygen in the reactor water (0.5) c) N-16 decays with a half life of 7 seconds so it is decayed by the time it leaves the hotwell. (0.5) REFERENCE VY LOT-05-405, Process Radiation Monitoring, pg 12 CRO Student Objective i K/A 272000 G.04 (3.3/3.4) 272OOOGOO4 ...(:KA'S)
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ISE6MQQyN@MlCS t _dE@l_lB@NSEE6_@NQ,E(Qlp_E(Qg ' ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
ANSWER 1.07 (2.00) u. Core flow would increase (0.5) due to a decrease in two phase l flow resistance (0.5).
j b. Core flow would increase (0,5) due to an increase in natural ! circulation (0.5).
REFERENCE VY LOT-02-204, Flow Measurement, Two Phase Flow, pg 8 CRO Student Objective 2 VY LOT-03-OO7, Reactor Reci r cul at i on System, pg 28 CRO Student Objective 9 K/A 293008 K1.28 (2.3/2.5), K/A 293008 K1.37 (3.2/3.4) 293OOOK128 293OOOK137 ...(KA'S) ANSWER 1.08 (2.50) c. BETA - the delayed neutron f raction - the f raction of neutrons in the core that were produced by delayed neutron precursors. (0,5) b. Reactor reTponse is slower (0.25). As BETA increases, the reactor period associated with the same reactivity addition is longer (0.75).
c. DECREASE (0.25). As Pu-239 production increases (0.25) and U-235 depletes (0.25), the core average beth will decrease due to Pu-239's beta being smaller (0.25).
REFERENCE VY LOT-02-OO9, Prompt and Del ayed Neutrons, pgs 7 & 9 CRC Student Objectives 3, 5&6 K/A 292003 K1.06 (3.7/3.7), K/A 292003 K1.04 (2.5/2.5) 292OO3K104 292OO3K106 ...(KA'S) _ _ _ - _ _ - _ _ _
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1___EBINCIELES_gE_NQC(E68_EQWEB_EL@NI_gEEB811gNt PAGE 23
* IHE60QQYN@dlCSt _dE@I_169NSEE6_@NQ_E(ylg_E(QW
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ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
ANSWER 1.09 (3.00) c. Decreases (0.5). The HPCI flow controller maintains a constant flow rate (0.5), thus the pump is pumping against less head and the discharge pressure is decreased (0.5).
b. Decreases (0.5). Flow rate remains constant (0.5), but the change in reactor pressure causes the pump to have to do less work to maintain the same flow rate, so turbine speed decreases (0.5).
REFERENCE VY LOT-02-205, Pump Char acteri sti cs, Pump Head, Pump Laws, pgs 10 & 11 CRO Student Objective 3 K/A 291004 K1.13 (2.6/2.7), K/A 293006 K1.29 (2.6/2.7), l K/A 206000 K4.09 (3.8/3.9) 206000K409 291004K113 293OO6K129 ...(KA*S) ANSWER 1.10 (2.00) c. FALSE (0.5) b. TRUE (0.5) c. FALSE (0.5) d. FALSE (0.5) REFERENCE VY LOT-02-115, Reactor Heat Balance, pgs 5 & 6 CRO Student Objective 1 K/A 293007 K1.13 (2.3/2.9) 293OO7K113 ...(KA'S) I ! _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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- IdE6dQQyN@dlCS t_dE@I_16@NSEE6_@ND_E(QlD_E(QW i .
ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
ANSWER- 1.11 (3.00) A- Rx power increases due to increased subcooling't+p) then (0.75) decreases sharply due to scram from turbine trip (hi level).
B- FW flow decreases due to FW pump trip caused by high RPV water (0.75) level (+177"). C- Core flow decreases due to recirc pump runback (<20% FW fl ow) (0.75) D- Steam flow drops to zero upon turbine trip then oscillates as (0.75) the BPVs open to reduce reactor pressure.
REFERENCE VY LOT-09-OO4, Operational Transients I, pgs 9, 10, 18 & 19 CRO Student Objective 3 K/A 259002 K3.01 (3.8/3.8), K/A 259002 K3.02 (3.7/3.7), K/A 259002 K3.06 (2.8/2.8), K/A 259002 K3.05 (2.8/2.9) 259002K110 259002K302 259002K307 ...(KA'S)
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE '25 ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
ANSWER 2.01 (1.50) c. Operation of the pum with the discharge valve sh could damage the pump bearing (0.5) y//3 g us 4e ex ce asW e-. crcd bu.s b.- Could lif t the relief valve at the discharge of the seal purge flow control station. (0.5) c.- Such operation can shorten seal life. (0.5) REFERENCE VY.OP 2110, Reactor Recirc System, pg 6 VY LOT-03-OO7, Reactor Recirculation System, pgs 30 e 23 CRO Student Objective 6 K/A 202001 G.10 (3.5/3.7), K/A 202001 K4.04 (3.0/3.1), 202OO1GO10 202OO1K404 . ...(KA'S) ANSWER 2.02 (1.50) g,, Tem spec. m e N .~ L a. ret w s. % h a. The rod will scra but t ; alc_;r ..Le ti.e....m 'JRO.75) b. The rod will ram but ri Pi't i. cl er! y .75) ci a!!sto w raw h w \ REFERENCE Teh,,a .spe.,4 . g..o , pg.73 VY LOT-03-OOS, Control Rod Drive Hydraulics, pg 11 CRO Student Objective 3 K/A 202001 K4.05 (3.8/3.8), K/A 295006 AK2.03 (3.7/3.8), K/A 295006 AA2.04 (4.1/4.1) 202OO1K405 295006A204 295006K2O3 ...(KA'S) ANSWER 2.03 (1.50) c. High outlet temperature (0.2) Low i nl et pressure (0.2) Low stator cooling water flow (0.2) b. Generator runback to < 4271 stator amps ( 29'/. ) (0.5) c., Turbine, Trip (due to failure to runback) (0.4) REFERENCE VY LOT-05-204, Stator Cooling Water System, pgs il & 12 CRO Student Objective 1 K/A 245000 A2.01 (3.7/3.9), K/A 245000 K3.OB (3.7/3.8), K/A 245000 K6.05 (2.9/2.9) 245000A201 245000K308 245000K605 ...(KA'S)
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, 2___PL@N]_DEg]QN_JNCLUDJNg_g8EE]y_6ND_EDER@gNCy_gygIEdg PAGE 12 6 , ANSWERS'-- VERMONT YANKEE -87/07/14-LUMB, T.
ANSWER' 2.04 (3.00) c.- 1 g. 3 b. 3 h. 2
.c. 1 i. 1 d.. 1 j. 4 a., 3 k.- 3 f. 4 1. 2 (0.25 each)
REFERENCE VY LOT-05-301, Circulating Water System, rg; 6 CRO Student Objective 1 VY LOT-05-303, Service Water, ;ng 6 CRO Student Objective 1 VY LOT-05-304, Reactor Building Closed Cooling Water System, pgs 8 L 9 CRO Student Objective 2 VV LOT-05-305, Turbine Building Closed Cooling Water System, pg 5.
CRO Student Objective 1-K/A 203000 K1.16 (3.1/3.2), K/A 204000 K6.01 (3.1/3.3), K/A 259001 K6.05 (2.7/2.7)- 203OOOK116 ANSWER 2.05 (2.00) 1) Leakage rate measurement is accomplished by measuring the time interval between two different l evel switch actuations in the sump as the sump fills with leakage. Whenever the time interval decreases to a prescribed setpoint' (indicating excessi ve leakage rate), an alarm annunciated in the Control Room. or pq *rtaw 4eo ch - basca on a4lw uw& s% h A P H s PS-(1.0) 2) The alarm will also sound if a sump pump runs longer- than a preset time interval. (1.0) REFERENCE VY LOT-05-413, Drywell Equipment and Floor Drains System, pg 10 CRO Student Objectives 3 L 4 K/A 223001 K1.04 (3.2/3.3), K/A 223001 A1.10 (3.4/3.6) 223OO1A110 223OO1K104 ...(KA'S)
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2[__E(@N1_DE@l@N_lyC(UDIN@_@@ EELY _AND EMERGENCY SYSTEMS PAGE 27
. ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T. q
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l i
. ANSWER. 2,06 (1.50)
jw.g we ccO:.r on h ob c.AP 4 5 4 4.I we-
- SLC pump running light on - Squib valve continuity light off - " Squib Valve Continuity Lost" Annunciator - SLC tank level decreasing - Reactor. power l evel decreasing - SLC pump indicates normal running discharge pressure - SLC inj ecti on . fl ow li ght illuminated (6 at.O.25 each)
REFERENCE VY OP 2114, Operation of Standby Li quid Control, pgs 4 & 5 VY LOT-05-409, Standby Li quid Control System, p gs 14 + to CRD Student Objective 4 K/A 211000 A4.OB (4.2/4.2) 211000A408 ...(KA*S)
-ANSWER 2.07 (3.00)
c..Will continue to operate (0.5). Operation under these conditions could allow contamination of the RCIC room and atmosphere from turbine and valve. steam leakage (0.5), b. Will continue to operate (0.5). Only adverse effect is the leakage f rom the CST or the from the condensate ECCS pressurizing station (i f the.
manual valve lineup was perf ormed because CST < 35% full or RCIC suction is f rom torus) (O.5). 01- Wo oAue.ss e e6e.sJr on 'Rcac. ope.raWen c. RCIC will not continue to operate (0.5). RCIC turbine will trip due to insuf ficient oil pressure to hold trip throttle valve open (0.5).
REFERENCE VY LDT-03-302, Reactor Core Isolation Cooling, pgs 24, 28 & 29 CRO Student Objectives 2, 5& 13 K/A 217000 K1.OB (3.3/3.4), K/A 217000 K4.05 (3.2/3.5), K/A 217000 A2.07 (3.1/3.1) 217000A207 217000K108 217000K405 ...(KA'S) l
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2 __gLeNJ_pg!]QN_JNgLUg}Ng_ggggly_gNp_ggggggNgy_gy!JEDE PAGE.'28
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ANSWERS -- VERMONT' YANKEE -87/07/14-LUMB, T. - l f ANSWER 2.08 (2.50) c. Normal - RPS MG sets (0.5) or- N V t 15 6 M NAC.C. % Alternate - 480V MCC BB (0.5) Yvd4 no ciben - Pc.t 5, % s.oku ku."E-b. - half scram
- start SBGTS or 'pc.i3 h4 TII iso \cRon - trip reactor building ventilation - i sol ate Of f-Gas System (valves OG-516A & OG-516B) - pow led h AEc6 reaw~he (3 9 0.5 each)
REFERENCE c+ z isq ,12.ec..hr hk.c.b'.~ Sekw , Pg l' VY LOT-03-108, Reactor Protection System, pas B, 9 & 31 CRO Student Objective 5 K/A 212000 A2.02 (3.7/3.9) 212OOOA202 ...(KA'S) ANSWER 2.09 (3.00)
'RHR Status RHRSW Statust RHR PUMPS A & C TRIPPED RHRSW PUMPS TRIPPED MOV-17 CLOSED MOV-89 CLOSED MOV-18 CLOSED , MOV-13 CLOSED MOV-65 OPEN MOV-27 OPEN MOV-25 OPEN-MOV-15 OPEN (10 at 0.3 each)
REFERENCE VY' LOT-03-306, Residual Heat Removal, pgs 11, 12, 16, 21, 22 & 27 CRO Student Objectives 2 & 3 _K/A 203000 K4.01 (4.2/4.2), K/A 203000 A3.08 (4.1/4.1), K/A 205000 K4.03 (3.8/3.8), K/A 205000 A2.05 (3.5/3.7) 203OOOA308 203OOOK401 205000A205 205000K403 ...(KA'S)
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ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, ANSWER 2.10 (3.00) c, yes (1.0) b. no (0.5) must insert the rod (O.5) c. no (0.5) must raise the grapple fully (0.5) or- m964,% w 5..m core REFERENCE ' VY LOT-05-408, Refueling Operations, pgs 5 & 6 CRO Objective 1 K/A 234000 K5.02 (3.1/3.7), K/A A3.02 (3.1/3.7) 234000A302 234000K502 ...(KA'S) ANSWER 2.11 (2.50) c. Level will increase (0.5) because the FCV's will lockup providing constant feed flow while power (steam flow) is decreased. (1.0)
(alt reason: FCV's drift open thus FW flow increases)
b. Flow will decrease.(0.5) because the flow control valve fails closed on a loss of air (0.5) REFERENCE VY LOT-05-109, Feedwater System, pg 19 CRO Objectives 1 &5 VY LOT-03-OO5, Control Rod Drive Hydraulics, pg 12 CRO Objective 3 K/A 259001 K3.01 (3.9/3.9), K/A 259001 K6.01 (3.0/3.0)
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K/A 201001 K6.03 (3.0/2.9), K/A 201001 K1.09 (3.1/3.2) 201001K109 201001K603 259001K108 259001K601 ...(KA'S)
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. -ANSWERS:--LVERMONT~ YANKEE- > -87/07/14-LUMB, T. it t j
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L ANSWER' 3.01. (3.00). > q c. 1) .'YES- '(0.5) ' J 2)' NO (0. 5)' 3)L YES, (0. 5)' 4) .NO (0.5)
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b. 'Halfiscram (0,5) rod block (0.5)
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REFERENCE l VY LOT-03-106, Aver.sge Power, Range Monitor (APRM), Transparency 7 CRO Student' Objectives 3 & 7 K/A 215005 K1.04 (3.6/3.6), K/A 215005 K3.01 (4.0/4.0), , K/A 215005 A2.03 ( 3. 6/ 3. 8 )' 215005A203 215005K104 215005K301' ...(KA'S) i ANSWER 3.02 (2 50), c. ADS valves remain as,is (0.5) j b. ADS valves close. (0.5) c. ADS valves close. (0.5)
-d./ ADS valves open (0.5)
p.iADS valves open 10,5) REFERENCE.
VY LOT-03-304, Automatic Depressurization System, pgs 12 - 14 CRO Student Objectives 2 & 4 K/A 210000 K5.01 (3.3/3.0), K/A 218000 K4.04 (3.5/3.6), i K/A 218000.A2.05'(3.4/3.6), K/A'218000 K6.02 (4.1/4.1), i K/A 218000 A1.05 (4.1/4.1)
'21BOOOA105 21BOOOA205 210000K404 21BOOOK501 218000K602 ...(KA'S)'
l l ANSWER 3.03' (2.50) j
.A11' inboard and outboard MSIV and recirc sample valves (0.5) to close (0.5) l These -switches are two position switches (AUTO /OPEN-CLOSE) thus this inter-lock' prevents the isolation vlaves f rom opening automatically af ter the ,
'
logic was reset. (1.0) Other GROUP I valve switches have three positions (CLOSE-NORMAL-OPEN) and require operator action to open. (0.5) ot- me-hod opesc-bA uc.1 o ue 45 h erw M mob,. ope.deh Mm u.0 texL is Pd. o^ REFERENCE ] VY LOT-03-211, Primary Containment Isol ati on System, pgs 14, 15, 37 & 38 CRO Student Objective 10 _ _ _ _ _ . _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ -
_____ - - - - _ hs L t c
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_ I N_ S_ T_ R_ _U. M_ E_ N_ T S_ _ A_ N_ D_ _ C_ O_ N_ T_ R_ O_ L_ S_ PAGE -31 L- cANSWERS -- VERMONT YANKEE -87/07/14-LUMB,.T.
l.
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.K/A'223OO2.K4.06'(3.4/3.5), K/A 223002 A4.03 (3.6/3.5)
223OO2A403 .223OO2K406 ...(KA'S)
..r..
ANSWER 3.04 (1.50) c.rTRUE (0.5) ,, b. FALSE (0.25)' the count circuit determines 2 f ' an LPRM is operable from the-input to the downscale trip unit (0.25) c.. FALSE-(0.25) the purpose of the RBM is to prevent the power in the
- bundles surrounding a control rod being. withdrawn, from approaching thermal limits (or to prevent fuel damage) (0.25)
REFERENCE
:VY. LOT-03-107, Rod Block Monitor, pgs 5, B& 11 CRO Student Objectives 1, 2&3 K/A 215002 G.04 (3.3/3.4), K/A 215002 K1.02 (3.2/3.1)
215002 GOO 4 215002K102 ...(KA*S) ANSWER' 3.05 (3.00) a.- Both recirculation pumps run back to 45%(0,5) as limited by the dual limiter on the output of the master contr ol l er (0. 5) .
'b. Recirculation 1 pump A runs back to 20% (0.5) due to the discharge. valve not full open bypass around the speed limiter not met (0.25). Recirculation pump B speed will be unaffected ( O.25).
c.. Both pumps will runback to 20% (0.5) because the FW flow signal to the recirt system is less than 20% (0,5) (the flow limiter is not bypassed when this signal is less than 20%) REFERENCE VY LOT-03-OO7, Reactor Recirculation System, pgs 22 & 23 CRO Student Objectives 5, 7, &B K/A 202002 K3.05 (3.2/3.3), K/A 202000 K6.04 (3.5/3.5)
.202OO2K305 202OO2K604 ...(KA'S) ;
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3 __lNS16UMENIS_gND_CQNI@gLS PAGE 32
.: ANSWERS -- VERMONT' YANKEE -87/07/14-LUMB, T.
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, ' ANSWER 3.06 (1.50) ,o. Increase (0.5> t b. Decrease (0.5) l c.. Increase (0. 5) .
REFERENCE VY LOT-03-OO2, Reactor Vessel I n st r umen t at i on , pgs 25-27 CRO Student Objective 7 K/A'216000 KS.01 (3.1/3.2), K/A 216000 K5.07 (3.6/3.8),
, . , '
K/A 216000 K5.10.(3.1/3.3), K/A 216000 K5.09 (2.9/2.9),
:K/A 291002 Kl.OO (2.8/2.9)
216000K501 216000K507 216000K509 216000K510 291002K108
...(KA'S) . ANSWER 3.07, (2.50)
Indicated feed flow would decrease while actual feed flow would remain et 100' percent. (0.5) The FWLCS would increase FW according to the change in indicated feed flow. (.045)' Lovel will begin to increase, generating a level error signal. (0.5) Fsod flow will decrease in an attempt to bring level back to normal . -(0. 5) Final level will signi ficantly higher (30") than ori gi nal , l evel '#(by"e es abl i shedr25Tl'$ fps and main turbine trip at 177" causing areactorSCRAM)G.25).JA.g f REFERENCE VY LOT-05-110, Feedwater Level Control, pg 19 CRO. Student Objective 5.d ) K/A 259002 K1.02 (3.2/3.3), K/A 259002 K1.03-(3.8/3.9), K/A 259002 K1.04 (3.5/3.6), K/A 259002 K3.01 (3.8/3.8), K/A 259002 K6.04 (3.1/3.1), K/A 291002 K1.07 (3.2/3.2) 259002K102 259002K103 259002K104 259002K301 259002K604 291002K107 ...(KA*S) l L (. _ _ _ . . . _ . - . _ - - _ _ _ _ _ - - _ - - _ _ - -
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3 f 3 __1NSl@UMENig,@ND,CQNIBg6g grAGE 33
. . ANSWERS -- VERMONT' YANKEE -87/07/14-LUMB, T._ .g ,
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3.00 -(2.50)
(Aug5 ANSWER) #
lq 1. - Service Water Pump A or C will start (depending on current pump
'
J 0J f'g'f) operation) (0.5) T
*
2. - RHR pump A starts.(0.5) .
;
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3. - RHR pump.B starts (0.3) af t'er. 5 second time delay (0.2) f)/ , 4. --Core Spray Pump A starts (0.3) after 10 second time delay (0.2) '(q 5. - Station Air Compressor cooling transfers from TBCCW-to service water (0.5)
(p. - RBu t.a ye A shssh Lo.Q Jrkr; i W& t wt. dalm.) ( c.2D REFERENCE vy ur c5.D4, R6c.c u , pg Iz.
VY LOT-05-208, 4KV Electrical Distribution System, pg 22 CRO Student Objective 7
/
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K/A 262001 A3.04 (3.4/3.6) 262OO1A304 ...(KA'S) ,
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, C ANSWER 3.09 (3.00) -
a c. hi r adi ati on at AOG inlet to final delay. pipe - 6.OE4 cpm ,
(0.3) ,)
or 9 . , monitor (RAN-OG-3127/3128) downscale or inop (f uncti on switch out of operat6) (0.3) FCV-11 (stack isolation) cloaes - - il (0.3) OG-3 (delay pipe drain to radwaste) closes (0.3) after 2 or 30minTDdependingonAOGS/Ubypassvalveposfyif'ys (0.3)
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b. hi radiation levels on refuel floor - 100 mr/hr (80 mr /he~ ' .CA ')a' (0. 3) ,
\
or hi radiation levels in rx b1dg ventilation exhaust duct ^14 mr/hr-(0.3)
(12 mr/hr -CAF)
YC55 Reactor Buliding Ventilation system shutdown (i sol ati on ) (0.3) 6**9 SBGTS starts (0.3) Group III (primary containment vent and purge valves) i sol ati on (0.3) WWr , REFERENCE VY ON 3152, Off Gas Hi gh Radi ati on, pg 1 ,,. VY LOT-05-405, Process Radi ati on ' dom i tori ng, pgs 27, 29, 3O(9 31
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CRO Student Objective 2 ) K/A 272000 G.08 (3.5/3.5), K/A'272OOO K1.02 (3.3/3.6), _.
K/A 272000 K1.06 (3.2/3.3) t ,, 272OOOGOO8 272OOOK102 272OOOK106 . . . ('t A ' S ) ? I
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.S ANSWERS -- VERMONT' YANKEE -87/07/14-LUMB, T.
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ANSWER I { 1 '. O
- (3.00) d c.\chd c. Bypass valves open 10% (or til limited by Reactor Flow Limit). Gi. ff )
Pr essur e decreases .>.t the ain Steam Line Averaging Manif old causing \ control valves to close o maintain pressur+ at EPR setpoint. (0.5) s No change in reagter p er. (0.5) t t
g b. Power decrease causes a pressure decrease which causes the pressure at - the Main Steam Line Averaging Manif old to decre.ne. (0.5) Control valves shut to maintain pressure at the EPR setdtint.j (0.5) Bypass v,
,.,
Wies remai n cl osed. (0.5) ,
,
REdNndolCE i , VY L07-05-004, Mechanical Hydraulic,Oontrol Systed,'pos 11 L 12
'
Q., CRO Student Objectivesi 1 ., 2, 3&4 , K/A 24100 K4.19 (3.6/3.7), K/A 24100 A1.02 (4.1/3.9), K/A 24100 A1.07 (3.8/3.7), K/A 24100 A3.09 (3.8/3.8) 241000A102 241000A107 241000A3OB ' 241000K419 ...(KA'S) d
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4 PAGE 35
* RADIOLOGICAL ---------------. 7CON..THOL -- .' ANSWERS'-- VERMONT" YANKEE -87/O//14 LUMB, T. :
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' REFERENCE ,
Vi LOT-06 -301, Operati ons Department ' Ad' mini strati ve Procedures, pg 12 ' Vermont Yankee Local Control Swithching Rules
, . . ' .
_'CRO Student Objective 3
'K / A ' 294001' K1,02. (3. 9/4. 5) ^
294001K102 h ..(KA'SS q ,
,. g., ANSWER: 4.O2 '12.005 700 psig is the minimum 'PV pressure that'Will pr'oduce adequate steam bO'
flow through one SRV td limi t_ peak. cl ad temperature to 2200 deg. - F.
or te inse cAgA corr., ucAig Co.5) N I@ r The surge of steam flow rgsulting f rom rapid depressurizatlon 'will ; lower fuel temperatures a providing additional time.to establish a-scurce of coolant i nj ecti on. (o.s') +t-t$N V t->l ; g.
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T f-)lREFE!\ENCE > y - '
, Vr}OE 3102, RPV Level Control, sheet 2 -'
VY. LOT-09-OO9, RPV Level. Control, pg 16
/
p: 4 CRO Student Objective A.2
+ K/A 295031 EK1.01 (4.6/4.7), K/A 295031 G.07 (3.7/4.0)
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41;_B899Epg8gp_ _Ng8D963_BBN9BD962_gDgBggNgy_9Ng PAGE 36 B99196991996_99NTBgh.
. ANZWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
,
. 1 ANSWER 4.03 (3.00) l '
a. - reactor.subtriticc1 (0.3)
- rx plant < 212^F (0.3) - rx coolant system vented (0.3) - no activity being performed which can reduce shutdown margin '-- bel ow that specified in TS (0.3) - irradiated. fuel is not being moved in the rx building (0.3) i or fwd cesk, b. - At least one door in each access opening is closed (0.5) - SBGT ' system is operable (0.5)
uAll reactor building automatic ventilation system isolation valves are secured in the isolated position. (Also building areop)erableor intact (0.5) REFERENCE EVV. Technical Specifications, pgs 3, 131 & 132 VY. LOT-03-210, Secondary Containment and SBGT Tech Specs, pgs 6 & 9 CRO Student Objectives 1 &5 K/A 290001 G.11 (3.3/4.2) 290001G011 ...(KA'S) ANSWER 4.04 (2.00)
. CRO - stati on himself at CRP 9-5 and be concerned with power l evel ,
control rod positions and reactor water level (1,0)
' ACRO - primary communi cati ons contact - handles all Control Room communications and any Gaitronics communications that cannot be answered by other personnel (0.5) - contacts security or plant telephone operator and directs them to limit unnecessary calls to the Control Room (0.5) - (monitors plant status to provide knowledgeable communications) - (is prepared to assume.SE duties in case of a fire - keep a rough log)
REFERENCE VY Standing Order #10 K/A 294001 A1.03 (2.7/3.7) 294001A103 ...(KA'S) l l
!
k. __ __ ___-__ - __ -
- - _ _ _ - _ - . - - - - - - - _ _
4-
,
p 4 PRDCEDURES - NDPMAL t_ABNQRMAlt_EMERGENQY_ANQ PAGE 37-
. ,BODlQ(QGlq@(_CQN16QL '
ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
l ANSWER 4.05 (2.00) c ;' 1. Scram and isolate the reactor (0.5) 2. Use RCIC and SRV-71A to control reactor level and pressure (0.5) 3. . Use the RHR system to cool the torus and when possible for shutdown cooling. (0.5)
-(concept of each objective will be accepted for full credit)
b. FALSE (0.5) REFERENCE VY DP 3126, Shutdown.Using Alternate Shutdown Methods, pgs 1 &3 VY LOT-09-013, Shutdown Using Al ternate Shutdown Methods, pgs 6 & 7 j CRO Student-Objective A.1 ' K/A 295016 AK2.02 (4.0/4.1), K/A 295016 AK3.01 (4.1/4.2) 295016K202 295016K301 ...(KA'S) n ANSWER 4.06 (3.00) a. Area where person may receive greater than 100 meem in one hour (1.0) b. (2 of below & 1.0 each)
-
A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- A radiation monitoring device which continuously integrates
.the radiation dose rate in the area and alarms when a preset integrated dose is received. (Entry into such areas with this monitoring device may be made after the dose rate levels in the area have.been established and personnel have been made aware of them.)
- A H.P. qualified individual with a radiation dose rate monitoring device who is responsible f or providing posi tive control over the activities within the area and who will perform periodic radiation surveillance.
REFERENCE
.VY, Technical Specifications, page 201 VY LOT-06-305, Chemi stry & Heal th Pnysi cs Admi n Procedures, pgs 13 & 15 CRO Student Objective 1 K/A 294001'K1.03 (3.3/3.8)
294001K103 ...(KA'S)
-_ _ _ _ _ -_ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ . _ - . _ _ _ _ _ _ _ __ __ - . _ _ _ _
r F . '*- 4 t__PQQQEQQBES_;_NQBd@Lt_@BNQSd@(t_EdEgGENgy_@NQ PAGE 38
* 599196QGlq@(_qQN16QL , " ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
' ANSWER' 4.07: (1.00) c. Scram reset' switch to-reset (both ways --1-4, 2-3) (0.5) b., Must wait until MSOP has' developed enough discharge pressure to reset the acceleration relay pressure switches. ( 0. 5 ) - REFERENCE VY OP 0101,' Reactor and Generation System Heatup to Low Power, pg 2 VY LOT-07-OO1, Preparation for Plant Startup, pg 22 CRO Student Objectives A.2 &.A.3
'K/A 294001'A1.02-(4.2/4.2), K/A 245000 K3.07-(3.6/3.7)
245000K307 294001A102 ...(KA'S).
. ANSWER 4.08 (1.50) y e-r-n
'(035 each 4m ..ny a) -
th: c: au i : 7~ F rc
- To. inert the containment f ollowing a startup or drywell entry (no longer than 24 hrs) -
To de-inert (purge) the containment prior to a shutdown (no longer than :24 hrs)
-Tu _ t i r ' '/ T er' air:1 Sp::i'ir tion, cur /ri!!aar= r;ui :::nts pa ySt7 REFERENCE VY OP-2115, Primary Containment, pgs 3 & 4 .
VY LOT-07-OOS, Turbine Generator Operation, pg 9 I CRO Student Objectives A.4, A.5 & A.6 K/A 223001 G.07 (3.7/3.8) 223OO1 GOO 7 ...<KA'S) l
_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - -
.- , ,
l
.4; -PROCEDURES - NORMAL _ ABNORMAL <
t _ tEMERGENCY _AND .PAGE H39 ! l 4 RADIOLOGICAL CONTROL ANSWERS -- VERMONT-YANKEE -87/07/14-LUMB, T.
ANSWER 4.09 (2.00) c.. Alert b. Unusual Event c. General Emergency d. Site Area Emergency (0.5 each) REFERENCE VY-AP 3125, Emergency Plan Classification & Action Level Scheme, pgs 1 &2 VY; LOT-06-307, Introduction to the Emergency Plan, pgs 6 & 7
. .
CRO Student Objective 2 ! VY LOT-09-OO3, Power Maneuvering.with Malfunctions, pgs 19, 20 & 21 ! CRO Student Objective A.2 K/A 294001 A1.16 (2.9/4.7) 294001A116 ...(KA'S) ANSWER 4.10 (1.50) a. 3100, 3101 (0.15 each) b. 3100, 3102 (0.15 each) c.'3100, 3103 (O.15 each) d. 3104- (0.3) e. NONE (0.3) REFERENCE VY LOT-09-OO1, Reactor Scram Response, pgs 12 - 14 CRO Student Objective A.3 K/A 295006 G.11 (4.3/4.5), K/A 295006 G.12 (3.8/4.4) 295006G011 295006G012 ...(KA'S) _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ - _ . -_ _- _ _ _ _ _ _ _ _ _ . ..
. * .
4. PROCEDURES -' NORMAL _ABNQRMALt_ t EMERGENCY _AND PAGE 40 4 88D1969GICBL_CQNIBQL
.
ANSWERS -- VERMONT YANKEE -87/07/14-LUMB, T.
ANSWER' 4.11 (3.00) 1.~ Perform OE 3100, SCRAM concurrently (0.25) 2. For the failed DG, energize the bus as f ollows: a. check open the normal supply breaker (0.25) b. manually start the DG from the control room (0.25) c. close the DG output breaker (0.25) d. if the DG restart fails, energize the bus from the Vernon tie (0.25) , i I 3.' Start or verify auto operation of the following: a. 2 service water pumps and SW2O closes (0.25) b, transfer of station air' compressor cooling water (0.25) c. restart station air compressors A and B -(0.25) d. Emergency DC equipment: 1) turbine emergency bearing oil pump (0.25) 2) emergency seal oil pump (0.25) 3) recire MG LO. pump (0.25) 4) vital MG set shift to DC (0.25) REFERENCE VY OT 3122, Loss of Normal Power, pg 1 VY LOT-09-OO6, Operational Transients III, pg 15 CRO Student Objectives 2 K/A 295003 G.10 (3.9/4.1), K/A 295003 AA2.04 (3.5/3.7) 295003A204 295003G010 ...(KA'S) ANSWER 4.12 (2.00)
(0.5 each) ,
c. YES I b. NO c. YES d. NO REFERENCE VY OP 2110, Reactor Recirc System, pg 3 VY LOT-07-OO2, Approach to Criticality and Heatup, pg 19 CRO Student Objective B.1 K/A 202001 K1.17 (3.1/3.3) 202OO1K117 ...(KA'S) l l l' J
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TEST CROSS' REFERENCE PAGE -1 s y, QUESTION: VALUE, REFERENCE- ________ ______ __________
:01.01 2.50 TELOOO1089 01.02- ' :2. 00 ' TELOOO1135 01.03 .3.00 -TELOOO1132-01.04' 1.00 .TELOOO1138 01.05~ 2.50 .TELOOO1115 101.06' 1.50- TELOOO1151
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TEST CROSS REFERENCE PAGE 2
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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _yERMgNT_YANKE@__________ REACTOR TYPE: _pWR-Gg3_________________ DATE ADMINISTERED: _@Zfg?f13________________ EXAMINER: _HQWE 1 _A.________________ CANDIDATE: _ _________________ JN@J@yCJ1gN@_JQ_CBNpJg81El Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each ( question are indicated in parentheses after the question. The passing \ grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF
.ATEGORY % OF CANDIDATE'S CATEGORY
_YBLUE_ _Jgl@L ___SCg6E___ _y@LUE__ ______________C@lEGQ@Y_____________ 39199__ 29199 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS l 39199-- _39199 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 29199-- _39199 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, l EMERGENCY AND RADIOLOGICAL ' CONTROL 29199-- 29199 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
.99199__ ___________ ________% Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
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l ! * NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS ] suring the administration of this examination the f ollowing rules apply:
. Cheating on the examination means an automatic denial of your application )
and could result in more severe penalties. )
'. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
i. Use black ink or dark pencil gnly to facilitate legible reproductions.
5 Print your name in the blank provided on the cover sheet of the examination.
i. Fill in the date on the cover sheet of the examination (i f necessary) . 2 Usa only the paper provided for answers.
. Print your name in the upper right-hand corner of the first page of each section of the answer sheet, l. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last answer sheet.
. Number each answer as to category and number, for example, 1.4, 6.3.
O. Skip at least three lines between each answer.
1. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
2. Use abbreviations only if they are commonly used in facility litetatute.
. 3. 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.
4. 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.
(6. If parts of the examination are not clear as to i ntent , ask questions of the examiner only.
7. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assi st anc e in completing the examination. This must be done after the examination has ' baan completed.
i
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8. When you complete your examination, you shall: a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including fi gures whi ch are part of the answer, b. Turn in your copy of the examination and all pages used to answer the examination questions.
c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions. i d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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5. THEORY OF NUCLEAR POWgR_ PLANI _gPgRAIJgN 2 _FLUlpSi _ANp PAGE 2 THERMODYNAMICS
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QUESTION 5.01 (3.00) Following an AUTO INITIATION of HPCI at a reactor pressure of 800 PSIG, rocctor pressure decreases to 400 PSIG. HOW are the following parameters offscted (INCREASES, DECREASES, REMAINS CONSTANT) by the change in reactor pressure? BRIEFLY EXPLAIN your choice.
ASSUME the HPCI System is operating in automatic as designed.
i a. HPCI pump discharge head (assuming NPSH remains constant). (1.5) b. HPCI turbine RPM. (1.5) QUESTION 5.02 (2.00) e. The reactor is operating at 100% power and flow. Explain WHAT happens to core flow (INCREASE, DECREASE, or NO CHANGE) and WHY it happens, with a reduction in power by control rod insertion.
Assume recirculation pump speed remains constant. (1.0) b. At low power conditions, an increase in reactor power by control rod withdrawal will (INCREASE, DECREASE, or NOT CHANGE) flow through the core. Choose the correct answer and BRIEFLY explain your choice. Assume recirculation pumps are running. (1.0) l l
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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i :, , . 5. THEORY'OF NUCLEAR POWER PLANT OPERATION g _FLUJpS1 _ANp PAGE 3 , ,_ THERM ____ _O_D_Y_N_AM_I_C_S _ ! l. QUESTION 5.03 (3.00) Match each of the lettered items with one of the numbere'd itsms.. '(Numbered items may be used more than once or not at all ao appropriate.) (3.0) 1. FRP 5. PCIOMR 2. APLHGR 6. APF 3.. CPR 7. , TPF 4. GEXL 8. LHGR _____a.- Parameter. by which plastic strain and deformation are limited to less than 1%. _____b. Ratio of bundle power required to produce onset of transition boiling somewhere in the bundle to actual bundle power.
_____c. Parameter ' by which peak clad temperature is maintained less than 2200 degrees F during postulated design basis accident.
_____d. Contains guidelines restricting power ramp rates above the threshold power.
_____e. Parameter by which fuel f ailure involving Pellet Clad Interaction is prevented.
_____f. Limi t changes over core life due to changes in the fuel's ability to transfer heat.
(***** CATEGORY Ob CONTINUED ON NEXT PAGE *****) _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ .
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' 5. THEORY OF NUCLEAR POWER PLANT OPERATIONg _ FLUIDS2 ,AND_ PAGE 4 THERMODYNAMICS ,
. . . QUESTION 5.04 (2.00)
During your Shift, an ADS relief. valve inadvertently opens. The. reactor is et 100% power.and 1000 psig. Use a Mollier Diagram or the Steam Tables to answer the following: . l a. STATE the tailpipe temperature, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurization. (0.5) b. If the Suppression Pool Pressure were to. INCREASE, STATE whether the Tailpipe Temperature would INCREASE, DECREASE, or REMAIN THE SAME. (0.5) c. If the r eactor starts to depressurize when the SRV is opened, STATE whether the Tailpipe Temperature will. INITIALLY INCREASE, DECREASE, or REMAIN THE SAME. (0.5) d. STATE the Reactor Pressure at which the Tailpipe Temperature would be at its MAXIMUM value (during the depressurization) . (0.5)
(ASSUME A SATURATED SYSTEM AND INSTANTANEOUS HEAT TRANSFER)
QUESTION 5.05 (1.00) Which one of the following conditions would tend to INCREASE the Critical Power l evel assuming all other variables remain unchanged? (1.0) NOTES ASSUME NORMAL FULL-POWER OPERATING CONDITIONS 1. High pressure feedwater heating is lost 2. Turbine control failure causing control valves to open 3. The axial power peak is RAISED 4 Recirc pump speed recuction using the master controller (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) __ __--___ _ __ _ -
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5.J THEORY OF NUCLEAR PQWER_ PLANT _gPERATIgNt_FLUIQSt_ANQ PAGE 5
,, ,IHERdQQXN@dlCS ' QUESTION. 5.06 (3.00).
Ascume the plant-had been operating at full power for one month and thGnescrammed. .The plant went through a normal cooldown and stayed shutdown for ten days. Given the f ollowing curve plotting delta k/k vsreus time, answer the f ollowing: a) At point (1), list five contributors to the negative reactivity? (1.0) b) ' What. causes the reactivity change between points (2) & (3)?' (0.5)
-c) What causes the increase in negative reactivity between points (4) . cn d (5)? (0. 5 )'
d) State the shutdown margin requirements according to the Technical Specifications ? BE SPECIFIC. (1.0) NOTE: SCRAM OCCURRED AT TIME ZERO (O)
-10 1-9 1 % -8 1. (1 )
DK -7 1 .
~/ -6 1 / \(3).
k -5 1 (2)
-4 1 . -3 1 \N -. (5) -2 1 (4)\.- -1 I_______________- _________________________________
O O 16 24/ HOURS 1 2 3 4 5 6 7 8 9 10 DAYS I
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5. THEORY OF NUCLEAR POWER PLANT OPERATION3 _ FLUIDS2 _ANp PAGE 6 _TH_ER_M_O_D_YN_A_M_IC_S _ _ _ _ l
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l DUESTION 5.07 (3.00) I R;garding reactor startups a) Does the magnitude of the initial level of source range l
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counts affect the estimated critical rod position? WHY? (1.0) b) The reactor is brought critical at 40% on IRM range 2 with q the shortest permissible stable positive period allowed by l OP 0100, REACTOR STARTUP TO CRITICALITY. Heating power is determined to be 40% on range 8 of IRM's. ;
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****SHOW ALL WORK **** I
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1) What is the doubling time if the period remains constant ? (1.0) 2) How long will it take for power to reach the point of adding heat if the period remains constant ? (1.0) QUESTION 5.08 (3.00) Figures 1, 2, and 3 (attached) contain charts of several key reactor pcrameters following a Feedwater Controller Failure to Maximum Demand.
For the areas marked, give the cause of each parameter change as eteted below. Initial reactor power is 100%. All automatic systems function normally. No operator action.
A- State WHY reactor power rises then sharply DECREASES. (0.75) , i B- State WHY feedwater flow DECREASES sharply. (0.75) f C- State WHY core flow DECREASES. (0.75) D- Explain the variations in steam flow. (0.75)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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~ 5 t__IMEQBy_QE_NgGLE86_EQWEB_E(@N1_QEE6@llgN t _E(gips _@NQ t PAGE 7 )
.IdE@dQQyN@dlGS QUESTION 5.09- (3.00)
a. Which one of the following best describes the core response to a shallow rod inserted two notches ? (1.0) 1. The axial flux peak will move up axially and total core power will decrease.
2. The axial flux peak will move up axially and total core power will not change.
3. The axial flux peak will move up axially and total core power will increase.
4. The axial. flux peak will not change and total core power will decrease.
-5. The axial flux peak will move down axially and total core power will not change.
b. For each of the following, CHOOSE the situation with t se greatest rod worth and explain WHY.
1. During a startup,
- the second rod withdrawn in a group-OR- - the third rod withdrawn in a group. (1.0)
2. During power operations, a rod
- in a control cell with 11% exit quality-OR- -in a control cell with 13% exit quality. (1.0)
QUESTION 5.10 (2.00) The reactor is operating at 80% power. Recirculation flow is increased to raise power. Describe how the various reactivity coefficients function to cause and terminate the power increase. Include in your answer each effect that causes a reactivity change, and why that effect causes its associated reactivity change. (2.0)
(***** END OF CATEGORY 05 *****)
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QUESTION 6.01 (1.50).
Concerning the Recirculation System, STATE the adverse affmets, if any, that operation under the f ollowing conditions could cauze. Consider each set of conditions separately.
l a. Recirc Pump A.was started at 12:50. The discharge valve.
l (RV-53A) . was opened at 13:08. (0.5) b. Recirc Pump B was~ isolated at'10:10. Seal purge to Recirc
' Pump B was i solated at 110: 12. (0.5)
c.. Recire Pump B is operating at minimum speed with a suction i
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pressure of.275 psig. (0.5) QUESTION 6.02- (3.00) The Residual Heat Removal (RHR)-System loop A is operating inLthe
' Shutdown Cooling mode with full flow'through the heat exchanger.
A valid LPCI initiation' signal' occurs. State the final pump and valve-alignment for the RHR and RHR service water systems (LOOP A ONLY) as a result of this event. Consider affected flowpaths only. BE SPECIFIC. A diagram of the RHR system is attached (Figure 4.) for your reference. ASSUME NO OPERATOR ACTION. (3.0)
-QUESTION '6.03 (2.50)
In order to reset a GROUP I isolation, certain valve control switches must be repositioned. Briefly describes
- WHAT switches must be repositioned, - to WHAT position, , - the reason WHY this is done, - and WHY-only certain Group I valve control switches require repositioning (2.5) ,
QUESTION 6.04 (2.00)
'The. reactor is operating at 66% power. Flow convertor A fails I such that its output is downscale. State ALL trips which )
will occur and the CAUSE(S) for each. A center rod is selected. (2.0) i l l
)
)
(***** CATEGORY'06 CONTINUED ON NEXT PAGE *****)
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6. PLANT SYSTEMS DESIGN _CQNTROLt_AND_t INSTRUMENTATION PAGE 9 G * l-f.
i QUESTION 6.05 (3.00) For each of the following situations, determine whether or not.the activity can occur. For those activities which cannot i occur, state the condition (s) which must change to allow it to occur.
j i
.a. Refuel bridge is over the vessel and in motion toward the ;
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fuel pool with the fuel grapple loaded. All rods are inserted. Mode switch is selected from SHUTDOWN to j STARTUP. Will' the bridge continue to move ?' (1.0) b. Refuel platform near the core. The f rame mounted hoist is loaded. One rod is at position 30. Can the load on the hoist be lowered into the vessel ? (1.0) c.' Refueling platform over the core. Mode switch in, REFUEL. , Grapple fully lowered and unloaded. Can a control rod be I withdrawn ? (1.0) I l I
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QUESTION 6.06 (2.50) I a. HOW will vessel l evel respond (DECREASE, NO CHANGE, INCREASE) if recirc flow is slowly reduced using the recirc flow master contr ol l er and control air pressure is lost to the Feedwater Control Valves ? WHY ? (1.5) b. HOW will cooling water flow respond (DECREASE, NO CHANGE, INCREASE) if control air pressure is lost to the CRD Flow Control valve. WHY ? (1.0) I i QUESTION 6.07 (2.50) l
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i a. STATE the normal and alternate power supplies to the RPS system. (1.0) b. WHAT automatic actions could occur on a manual transfer of RPS bus A f r om its normal to alternate power supply? List three (3). (1.5)
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6, PLANT SYSTEMS DESIGN _QQNTRQLt_AND_ t INSTRUMENTATION PAGE 10
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QUESTION 6.08 (2.50)
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A Loss of Normal Power (LNP) signal is present for Essential Bus 4 and reactor vessel water level is 80" with reactor pressure at 325 psig.
What five (5) ACTIONS (load sequencing) will occur after the accociated diecel generator is supplying pow 7r to the bus and load shedding . i s compl ete. Include any associated time delays. (2.5) i I QUESTION 6.09 (3.00) For each of the f ollowing annunciators associated with Process Radiation Monitoring, STATE ALL signals that will cause the alarm and WHAT AUTOMATIC ACTIONS will occur, if any. Include setpoints and any associated time j dalays.
s. AOG SYS OUT RAD TIMER START (1.5) b. RX BLD VENT CH A RAD HI and RX BLD VEN's CH B RAD HI (1.5) QUESTION 6.10 (2.50) For the following situations, state whetner the Automatic Depres-surization System (ADS) relief valves will OPEN, CLOSE or REMAIN AS IS.
Con si d er each set of conditions separatel y, a. ADS initiating signal sealed in, ADS valves open . . . . reactor water level then rises to 177 inches. (0.5) b. ADS initiating signal sealed in, ADS valves open . . . . ADS timer reset buttons,are then depressed. (0.5) c. ADS initiating signal sealed in, ADS valves open . . . . then a DC power failure occurs that affects all busses supplying ADS valves. (0.5) d. . ADS initiating parameters present, a loss of the containment air system supply to the drywell has occurred, 120 second timer timing out . . . . then the 120 second timer times out. (0.5) c. ADS initiating parameters present, all ECCS pumps are secured except for CS pump B which is running with a discharge pressure of 175 psig, 120 second timer timing out . . . . then the 120 second timer times out. (0.5)
(***** END OF CATEGORY 06 *****)
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ZL- PROCEDURES' _NQRL1Al t_A_BNQRMAl t_ EMERGENCY AND < ~( FAGE M 1 RADIOLOGICAL CONTROL 5, i (-
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QUESTION 7.01 (2.00) ) o J.
a. What are the three (3) basic objectives of OP 3126, SHUTDOWN ,:
( 'g, f '
USING ALTERNATE SHUTDOWN METHODS? t, r, (1.5)
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b. TRUE or FALSE? Engaging RCIC, RHR or the DG transfer switch $s will cause a loss of the control function in the Control Room but will not affect the automatic functions or svstem' interlocks. ' (0.5) \t f f< A G,' fy;- r l pf[,
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QUESTION 7.02 (2.00) t
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OE 3102, RPV LEVEL CONTROL, directs the operator to (steam tool & the reactor if all possible injection into the reactor becomes unevai1able and 1evel drops below -87. This is accomplished by ,d opsning one SRV. If during steam cooling, RPV presst re drops - bel'iow y
.700;psig, the procedure directs the operator to emergency derpressurize.
a , {3, ,, / s WHAT is significant about 700 psig and WHY is RPV depressurization required below 700 psig. (2.0)
(
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,N QUESTION 7.03 (1.50) ,j'
OP 2115, PRIMARY CONTAINMENT, and MOO Directive 79-4, Rev. 2 allow ,' f f' bypassing the interlock, which keeps the containment vent and purge > velves closed while in the run mode, for several reasons. ',., '
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,b l Give three (3) of these reasons. 8 < ( {. 5 ) , -
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7. PR[jjgphRgS,,_NgRMAL _ABNgRMAl 2 _ iEMERGENCY _A3p PAGE 12
.Ud919'=gGIcGy_ggNTRg6 c-t',, s I' s
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's QUESTION 7.04 (2.00) '
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f l S(Jtewhether each of the f ollowing gets of plant operating conditions l
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would be likely to cause thermal s t r'p t i.41 c at i on (per OP 2110, REACTOR V 'v RECIRC SYSTEM) . Consider eOch set at gonditions separately.
' , a. 0%" power, IMMEDIATELY following!a r'eactor scram during I I which both recire pumps trippedi yte.rtup of both recirc '
/ (0.5?
oumps is in progress. ,
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b. 20% power, both recirc pumps ruening at minimum speed. (0.5)
#
c. . 5% power, plant heatup in progresa, one recire pump running 4 'O.5) at minimum speed. '
$
d. k65% power, one recire pamp running al/SOZ rated speed. -
(0.5) '
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GLESTION 7.05 (1.50) / i Whic of the following is (are) symptom (s) that the #1 Scifice of a recirculation pump seal is plugged ? I (1.5)
%
Ed.ni seal pressure decreases ' e \ It t 3 i e
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seal pressure i ncria.> anes ( b.[41 f;. c. yd2 seal pressure dNcreases
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(: d. 47 seal pravsure increases ,
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e. controlled leakage through #2 ori ficd ,6ecreases f f cont'r olled leakage through #2 orifich increases. /' g .1- neal temperatures decrease
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-OUESTION 7.06 (3.00) i
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The Unit has had a loss of normal power and one diesel generucor has failed to start. WHAT are ALL of your i mmedi at e s a i t i nn7 'i,.sv'
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4GT 3122 LOSS OF NORMAL POWER 7 BE SPECIFIC. , (3.0)
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a. . . 7. . PROCEDURES V NORMAL.u A BNQRMA k EMERGENCY _AND- PAGE 13 R@ Dig (QM C@k,CQt!!6QL Y.
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p QUESTION 7.07. (2.00) i<
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Lig The reactor has scrammed on high &essel pressure and'you are performing 4.fi" actions.per.OE 3100 SCRAM. You notaa that drywel1 pressury i s 2. 7 psi g.
% What'OE procedure (s) do you enter'and WWY ? (2.0)
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,; QUESTION 7,08 , (2.00) .m .
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M Procedure O.P. 0111 Shutdown to Cold Shutdown cautions the opsrator to limit stack gas release rate to less than the 3Joch. Spec. Limit. Why is.<this caution emphasized during the
. UkNhutdown/cooldown evolution? >
(2.0)
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.. ( s A ;.;s Q(JE3 TION 7.09 (2.00) .,
a If. drywell . temperature should exceed the limits impoded by the "RPV Saturation . Curve" of. procedure OE 3103 DRYWELL PRESSURE AND TEMPERATURE CONTROL,- the' operator wills be directed to emergency depressurize and than flood the RPV. What is the gasis f or RPV emergency g 'depressurizatlon in this case? (2.0) ..:, W- 5 ipuESTION 7.10 (2.5DI 1 , y La. Why should manual starting and stopping of reactor feed pumps be ?) accomplished withLthe discharge. valve closed ? (three reasons) (1.5) h >; ! 1 .* -
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. i N b', ; The plant is shutdown, condenser hotwell levels are inte the i circulating water (CW) tube bundles and CW is secured. WHY are you cautioned to sample the CW system (per OP2170 j Condennate System) prior to the restart of the CW system 7 (1.0)
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.s ; .s '7 . - PROCEDURES'- NORMAL 2_ABNgRMAL _gMgRGgNCY_@Np 3 PAGE 14 ,
R__A_D_I_OL_O_G_I_C_AL__ _ _ __ CON _T_R_O_L-QUESTION 7.11 (3.00) a.- Why should valve HPCI-25 (pump minimum fl ow) be checked closed following a HPCI system isolation or trip ? (1.0)
.b. The. HPCI system is operating af ter an automatic start. You are performing actions per the OE procedures. Adequate core cooling is NOT yet assured. Assume other systems are operable.
-You 'are informed of the following HPCI parameters:
- pump suction pressure 22 psig ' steam supply pressure 553 psig - turbine exhaust pressure 171 psig -
turbine RPM 3876 RPM
- turbine vibration 1.3 mils Would you continue HPCI operation ? WHY OR WHY NOT ? (2.0)
QUESTION 7.12 (1.50) The reactor l's at full power. As a part of rounds this shift, the AO
.is essigned to monitor performance of RCIC components while RCIC full flow testing is in progress. Is a Radiation Work Permit (RWP)
required 7 If not, justify WHY not. If so, state the type of RWP to use and justify WHY it should be used. (1.5) l (***** END OF CATEGORY 07 *****) l
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QUESTION B.01 (3.00) Stcmding Order #10, dated 10/15/85, gives guidelines for operator responsibilities during accidents when EOP's are used.
o. What are the initial duties of the SCRO until j reassignments are made by the Shift Supervisor ? (1.5) I b. What are the initial duties of the Shift Supervisor ? (1.5)
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QUESTION 8.02 (2.00) For each of the functions described below, MATCH it with the appropriate ) teg that would be used. (2.0)
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FUNCTION TAGS c. Provide a reminder that special consideration 1. White tag is required before operation 2. White tag with b. Indicate tne presence of an installed ground red border c. Provide visual indication that operation is 3. Yellow tag with not allowed for the protection of personnel black lettering or equipment or necessary to maintain system integrity 4. Orange tag with black lettering d. Provide visual indication that operation is ; not allowed except by the Local Permissive 5. Red tag with Test Person (LPTP) or a person he directly black lettering asks to operate it for him I QUESTION 8.03 (3.00) During a tour of the plant you notice that Secondary Containment Integrity is NOT established, c. What five (5) plant conditions must exist to ensure that Technical Specification requirements are met without Secondary Containment Integrity ? (1.5) b. What three (3) conditions must be met to establish Secondary l Containment Integrity? (1.5)
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QUESTION B.04 (3.00) Tho reactor is at 100% power. RHR service water pump A i s out of service
; for bearing replacement. Diesel Generator B has just been declared inoperative due to a leaky fuel line. WHAT are ALL the actions required according to the Technical Specifications ? (3.0) *************************************************************************** * NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS * * TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS * * OF THE T.S. THAT YOU USE TO DEVELOP YOUR ANSWER. * ***************************************************************************
OUEST10N 8.05 (2.50)
.The reactor is shutdown. A test of the refueling interlocks, which requires placing the mode switch in startup, is necessary. All APRM's aro inoperable. Can the refueling interlocks be tested ? If not, WHY not; if so under WHAT conditions 7 (2.5) *************************************************************************** * NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS * * TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS * * OF THE T.S. THAT YOU USE TO DEVELOP YOUR ANSWER. * ***************************************************************************
QUESTION 8.06 (3.00) The reactor is at 100% power. You are informed that a RCIC MAIN STEAM LINE TUNNEL TEMPERATURE instrument i s inoperabl e. According to the Technical Specifications, WHAT actions are required 7 (3.0)
*************************************************************************** * NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS * * TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS * * OF THE T.S. THAT YOU USE TO DEVELOP YOUR ANSWER. * ***************************************************************************
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QUESTION 8.07 (2.50) c. For each of the f ollowing situations given below, classify the ovent using TABLE I of AP 3125 " EMERGENCY PLAN CLASSIFICATION AND ACTION LEVEL SCHEME" (attached).
1. The reactor has scrammed and the MSIV's have closed due l
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to high radiation. Primary leakage is 80 gpm. (0.5) 2. HPCI system fails during a surveillance test. (0. 5) ' 3. A man is injured while working on contam4nated equipment and he needs evacuation to the hospital. (0.5) 4. Offsite power lost at 1:20 AM with failure of D/G's to start. It is now 1:40 AM. (0.5) b. TRUE or FALSE. You have just declared a site area emergency. The Vermont State Police should be notified before the NRC. (0.5) OUESTION 8.08 (2.00) Tha reactor scrammed from 100% power due to high neutron flux cauced by a complete MSIV closure. Has a saf ety limit been eax e s e d e d ? WHY or WHY NOT ? (2.0) QUESTION 8.09 (2.00) Tha plant is in an outage and all fuel is removed from the reactor.
R moval and replacement of the IRM detectors and detector tubes is in progress. Is a dedicated SRO required to be in charge of this operation ? Justify WHY or WHY NOT in your answer. (2.0)
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I- QUESTION 8.10 (2.00)
:A 22 year.old contractor has a. job in a 150 mr/hr "High Radiation Area".
He has received 600 mrem in the current quarter at another facility.
H3 does.not have'an up to.date NRC Form 4. His li f etime exposure is 19.25 REM.
9 9 $HOW 4L.(. W W sc 94 a. In' accordance with Vermont Yankee Administrative whole body exposure limits, how long can .this person work in this area 7 (1.0) b. In accordance with 10CFR2O whole body exposure limits, how long can this person work in this area ? (1.0)
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..ISEBdQQyN@dlGS ANSWERS'-- VERMONT YANKEEE -87/07/14-HOWE, A.
-ANSWER 5.01 (3.00) e. Decreases-(0.5). The HPCI flow controller maintains a constant flow rate (0.5), thus.,the pump is pumping against less head and-the discharge pressure is decreased (0.5).
b.- ' Decreases (0.5). . Flow rate remains constant (0.5), but the change in reactor pressure causes the pump to have to do'less work to maintain the same flow rate, so turbine speed decreases (0.5).
REFERENCE VY LOT-02-205, Pump Characteristics,. Pump Head,. Pump Laws p.10,11 SCRO Student Objective 3, 4 K/A 291004 K1.13 (2.6/2.7), K/A 293006 K1.29 (2.6/2.7), K/A 206000 K4.09 (3.8/3.9) 206000K409 291004K113 293OO6K129 ...(KA'S) ANSWER 5.02 (2.00)- a. Core flow would increase-(0.5) due to a decrease in two phase flow resistance-(0.5).
b. Core flow would increase (0.5) due to an' increase in natural circulation (0.5).
REFERENCE.
\A/ LO1-02-204, Flow Measurement, Two Phase Flow p.8 SCRO Student Objective 2-LOT-03-OO7, Reactor Recirculation System, p.28 SCRO Obj. B
'K/A 293008 K1.28 (2.3/2.5), K/A 293008 K1.37 (3.2/3.4)
293OOOK128 293OO8K137 ...(KA'S) 1-
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..,. -5g_'THEORYJOF NUCLEAR-POWER PLANT OPERATIONt _FLgIQS t_ANQ : PAGE 20 ;IHERMQQYN@M[CS ANSWERS -- VERMONT YANKEE -87/07/14-HOWE, A.
! ANSWER 5.03 (3.00)
~e. 8 b. 3 c.- 2 d. 5-s. O cr 5 f. 2 (0.5 each)
REFERENCE VY LOT-02-312, Thermal Limits p.19,36,40,41
.SCRO Student Objectives 1, 10, 11 & 12 VY LOT-02-405, PCIOMR p.O SCRO Student Objective 3 ,K/A.293OO9 K1.08 (3.0/3.4), K/A'293009 K1.06 (3.4/3.8), .K/A 293009 K1.10 (3.3/3.7), K/A 293009 K1.18 (3.2/3.7),
K/A 293009 K1.36 (2.8/3.4)
;293OO9K106 293OO9K108 293OO9K110 293OO9K118 293OO9K136 ;...(KA'S)
i ANSWER 5.04 (2.00) a. 295 deg F-(+- 15 deg F)- (0.5) b .- Increase (0,5) c. Increase (0.5)
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I d. 450 psia (+- 50 psia) (0.5) REFERENCE VY LOT-02-104, Mollier Diagram,p. 4,5, SCRO Obj. 1 VY LOT-02-105, Steam Tables p.6,10 SCRO Student Objective 2 K/A 293003 K1.23 (2.8/3.1) 293OO3K123 ...<KA'S) )
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ANSWERS -- VERMONT YANKEE -87/07/14-HOWE, A.
ANSWER 5.05 (1.00) 2 or i REFERENCE VYNPC, LOT-02-312 p. 17, 18 SCRO Obj. 1.e, 3 KA 293009 K1.21 3.1/3.6, K1.22 2.9/3.3, K1.23 2.8/3.2, K1.24 2.7/3.2, K1.26 2.6/3.1 293OO9K121 293OO9K122 293OO9K123 293OO9K124 293OO9K126
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ANSWER 5.06 (3.00) a) 1) Control Rods 2) Mod. Temp. React, Coeff. contribution at 545 deg.F (-4% delta k/k) 3) Fuel Temp. React. Coeff. (d opp l er ) at 545 deg.F ( .4% delta k/k) 4) Xenon (-2.7% delta k/k) 5) Samarium ( .4% delta k/k) (.20 each) (1.00) b) Plus reactivity from cooling (.25) interacting with the negative reactivity being added by Iodine decaying to Xenon.(.25) (0.50) c) Samarium buil<lup (f rom decay of Promethium). (0.5) ,
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d) Shutdown Margin (required by Tech. Spec. 4.3.A.1)
---demonstrate a shutdown margin of 0.25 per cent delta k (0.5) ---at any time in core lif e (0.25) ----with the highest worth operable control rod withdrawn (0.25)
REFERENCE VYNPC; LOT-02-011,p.4,5,6, SCRO Obj. 2.; -012,p.4,5,6, SCRO Obj. 23-015,p.8,9,10, SCRO Obj. 43 -017,p.8, SCRO Obj. 4,
-018, p . 0, 4, SCRO Obj. 6.
LOT-02-OO5,p.8,9, SCRO Obj. 7, T/S 4.3.A.1 KA 292006 K1.13 2.6/2.6, K1.2Q 1.6/1.7, 292004 K1.02 2.5/2.6 292004 K1.00 2.2/2.4; 292002 K1.10 3.2/3.5 292OO2K110 292OO4K102 292OO4K108 292OO6K113 292OO6K120
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ANSWERS -- VERMONT YANKEE -87/07/14-HOWE, A.
ANSWER 5.07 (3.00) a) No (0.50). The critical control rod position is a f unction of
;the- Keff or reactivity of the reactor and is not a function of the source count rate (0.50). (1.00)
b.
1) From OP 0100, shortest permissible stable period equals 30 sec.(0.5) Thus. Doubling time equals 30/1.44 = 20.8 seconds. (0.5) 2) 40% range 2 is equal to 0. 04*/. on range B P(0) =-0.04 P(t) = 40 Period = 30 seconds ) P(t). = P(0) e ^(t/ period) 40 = 0.04 e ^(t/30 sec) Ti me = ' 1 '.O. 5 ;;o-dr :- ' -d ; . . 00.5 .2 07, A A n, J A 7,2 A ( 1. 0 )
(NOTE: Grade method if period is different)
REFERENCE VYNPC, LOT-02-010 p.6,.SCRO Obj. 1; LOT-02-OO7 p.4-6, SCRO Obj. 1 ' KA 292003 K1.OB 2.7/2.8, 292008 K1.04 3.3/3.4 292OO3K108 292OOBK104 ...(KA'S) ANSWER 5.08 (3.00) A- Rx power. increases due to increased subcooling (+p) then *(0.75)
' decreases sharply due to scram from turbine trip (hi level). j B- FW flow decreases due to FW pump trip caused by high RPV water
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(0.75)
level (+177"). C- Core flow decreases due to recirc pump runback. (<20% FW fl ow) (0.75) D- Steam flow drops to zero upon turbine trip then oscillates as (0.75) the BPVs open to reduce reactor pressure.
REFERENCE VYNPC, LOT-09-OO4, Operational Transients I, p.9,10,18,19 SCRO Obj. 3 259002 Rx Water Level Control System (Group I Systems) ;
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KN of effect of loss of FWCS on following parameters K3.01 3.8/3.8, K3.02 3.7/3.7, K3.06 2.8/2.8, K3.05 2.8/2.9, 259002K110 259002K302 259002K307 ...(KA'S)
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5. THEORY OF NUCLEAR POWER PLANT OPERATIONt _ FLUIDS t_AND PAGE 23
.IHE@dODYN@dlCS ANSWERS --- VERMONT YANKEE -87/07/14-HOWE, A.
! ANSWER 5.09 (3.00) a. 2 (1.0) b. 1. The second rod withdrawn (0.25). As each rod is withdrawn, the rod density decreases thus core average flux increases.(0.25) Thus the ratio of local flux to core flux is higher with higher rod density resulting in higher rod worth. (0.5) 2. In a cell with 11 */. ex i t quality. (0.25) This cell has less voids (0.25) thus the percentage of thermal neutrons is higher (due to increased moderation) resulting in higher rod worth because the local flux to core average flux is higher. (0.5) l REFERENCE VYNPC LOT-02-207 p. 45,46, SCRO Obj. 3,4 LOT-02-016 p. 8,9,11,12,13 SCRO Obj. 4,5 KA 292005 K1.09 2.5/2.6, 292005 K1.12 2.6/2.9, 292008 K1.27 3.4/3.5 292OO5K109 292OO5K112 292OOBK127 ...(KA'S) ANSWER 5.10 (2.00) As flow increases, the void content in the core decreases (0.25), causing a positive reactivity addition due to the void coefficient (0.25). As pow:;r starts to increase, the fuel temperature begins to rise (0.25). As the fuel temperature increases, the fuel temperature coefficient (doppler) adds negative reactivity (0.25) to slow the power rise. As the higher fuel temperature is transferred to the coolant (0.25), void generation increases (0.25), which also adds negative reactivity. The power increase is terminated by the combined negative effects of the void and doppler coefficients (0.5).
REFERENCE VYNPC, LOT-02-207, p.14,15,16,17,18 SCRO Obj. 4 KA 202001 K1.02 4.1/4.1, 292000 K1.20 3.3/3.4 202OO1K102 292OOOK120 ...(KA'S) l L______ _ l
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ANSWER 6.01 (1.50) a.- Operation of the pump with the discharge valve shut could damage
'the pump bearings. 5)
b.
h( xcessn% e at d NVfs O~r-4-!? e Could. lift the relief valve at the discharge of the seal purge flow control-station. (0.5) c.' Such operation can shorten seal life. (0.5) REFERENCE VY OP 2110, Reactor Recirc System, pg 6 VY LOT-03-OO7, Reactor Recirculation System CRO Student Objective 6 K/A 202001 G.10 (3.5/3.7), K/A 202001 K4.04 (3.0/3.1), 202OO1G010 202OO1K404 ...(KA'S) ANSWER 6.02 (3.00) RHR Status: RHRSW Status: RHR PUMPS A'& C TRIPPED RHRSW PUMPS TRIPPED MOV-17 CLOSED MOV-89 CLOSED MOV-18 CLOSED MOV-13 CLOSED MOV-65 OPEN MOV-27 OPEN MOV-25 OPEN MOV-15 OPEN (10 9 0.3 EACH) REFERENCF, VYNPCI LOT-03-306 RHR,p.11-13,15-17,21,22. SCRO Obj.2, 3.
LOT-03-306H , p.3-6,10-12.
KA 203000 K4.01 4.2/4.2, A3.OB 4.1/4.1 205000 K4.03 3.8/3.0, A2.05 3.5/3.7 203OOOA308 203OOOK401 205000A205 2O5000K403 ...(KA'S) ..
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. ANSWER $ 6.03 (2.50) All inboard.and' outboard'MSIV and recirc sample valves (0.5)
'to cl ose - (0.5) . ; These' swi tches are two posi tion switches (AUTO /OPEN-CLOSE).thus this interlock preventsithe isolation valves f rom opening automatically after the logic was' reset. - (1.0) Other GROUP I valve-switches'have three positions (CLOSE-NORMAL-OPEN) and' require operator action to o (0.5)
c< ~ % ' & pen. m k $ "N W Y & y'= W & w U s ' AM *
= REFERENCE-VYNPC; LOT-03-211 p.14 SCRO Obj.10 'KA'223002 K4.06' 3.4/3.5, 223002 A4.03 3.6/3.5 1 223OO2A403 '223OO2K406 ...(KA'S)
ANSWER 6.04 (2.00) ROD BLOCK' - due to flow comparator mi smatch >7% (0.5)
(p.12; 10% APPENDIX B of handout ***CAF**)
--duesto flow biased trip (66) (.66W} + '42; W=O) ( 0. 5)'
- RBM: flow biased. trip (0.5)
ONE-HALF SCRAM Ch. A - due to-flow biased trip- (66> (.66W} + 54; W=0) (0.5) REFERENCE VYNPC;' LOT-03-106 p.12,13,25 SCRO'Obj.3,5,6; LOT-03-OO6 p.7, CRO Obj. 7-LOT-03-107 p.9, CRO Obj. 5 KA 215005 K1.16 3.3/3.4, 215005 K6.07 3.2/3.3, 215000 K1.10 3.3/3.3 215002 K1.01.2.9/3.0, 212000 K6.02 3.7/3.9 215000K110' 215005K116 215005K607 ...(KA'S) ANSWER 6.05 (3.00) a. yes (1.0) b. ' rso '( 0, 5 ) must insert the rod (0.5)
- c . ' no (0.5) must raise the grapple fully (0.5)
REFERENCE .j VYNPC; LOT- 05-400 p. 5,6 SCRO Obj. 1 j
'KA.234000 K5.02 3.1/3.7, A3.02 3.1/3'.7 l-234000A302 234000K502 ...(KA'S) j
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,AN9WERSr-- VERMONT YANKEE. -87/07/14-HOWE, A.
-ANSWER '6.06 (2.50)
*a.. Level wil'1. increase (0,5)'because the FCV's will lockup providing constant-feed flow while power (steam flow) is decreased. (1.0) . (alt reason: FCV's! drift open thus FW flow increases)
b .n Flow will decrease (0.5) because the flow control valve fails closed on alloss of air (0.5) REFERENCE VYNPC; LOT-05-109 p.19, SCRO Obj. 1,5 LOT-03-OO5 p.12, SCRO Obj. 2-KA 259001 K3.01 '3.9/3.9, K6.01 3.0/3.0 201001 K6.03 3.0/2.9, K1.09 3.1/3.2 201001K109- 201001K603 259001K100' 259001K60) ...(KA'S) ANSWER 6.07 (2.50)
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Normal - RPS MG sets '( 0. 5) or #kC OA # "' #IfC 9 a.- Alternate - 480V MCC BB (0.5)
.- AsIf lavia[s% Pcts 6ne ws' eb % cz, nt .b. - half scram - start SBGTS }or Acfy gaogr LI bc/a#ew - trip reactor building ventilation) - i sol ate Of f-Gas System (valves OG-516A & OG-5168) (3 @ 0.5 each)
REFERENCE ! VY LOT-03-108, Reactor Protection System, p. 9, 31 SCRO' Student Objective 5 K/A 212000 A2.02'(3.7/3.9) i 212002.K2.01 (3.2/3.3) 212OOOA202 212OOOK201 ...(KA'S)
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6. PLANT SYSTEMS DESIGN2_CgNTBgL 1_@Np_INSTRUMENI@IJgN PAGE 27 ANSWERS -- VERMONT YANKEE -87/07/14-HOWE, A.
i \ l i ANSWER 6.08 (2.50)
(A"1 h 1. - Service Water Pump A or C will start (depending on current pump operation) (0.5)
2. - RHR pump A starts (0.5) ] 3. - RHR pump B starts (0.3) after 5 second time delay (0.2) 1 4 - Core Spray Pump A starts (0.3) after 10 second time delay (0.2) 5. - Station Air Compressor cooling transfers from TBCCW to service water (0.5) gbcca y & A (o.d [ * / M. CL N ,V
,
k O' N REFERENCE vy .t.or . c s -so y , gacw p, tL-VY LOT-05-200, 4KV Electrical Distribution System, p.23 SCRO Student Objective 7
K/A 262001 A3.04 (3.4/3.6) 262OO1A304 ...(KA'S) ANSWER 6.09 (3.00) a. hi radiation at AOG inlet to final delay pipe - 6.OE4 cpm (0.3) or monitor (RAN-OG-3127/3128) downscale or inop (f unction switch out of operate) (0.3) FCV-11 (stack i sol ati on) closes (0.3) OG-3 (delay pipe drain to radwaste) closes (0.3) after 2 or 30 min TD depending on AOG S/U bypass valve position (0.3) I b. hi radiation levels on refuel floor - 100 mr/hr (80 mr/hr - CAF) (0.3) or hi radiation levels in rx bldg ventilation exhaust duct - 14 mr/hr (0.3)
(12 mr/hr -CAF)
YD Reactor Buliding Ventilation system shutdown (0.3) H SBGTS starts (0.3)
\gfM Group III (primary containment vent and purge valves) isolation (0.3)
REFERENCE VY ON 3152, Off Gas High Radi ation, p.1 VY ON 3153, Excessive Radiation Levels, p.1 VY LOT-05-405, Process Radiation Monitoring, p.27,29,30 SCRO Student Objective 2 K/A 272000 G.08 (3.5/3.5), K/A 272000 K1.02 (3.3/3.6), K/A 272000 K1.06 (3.2/3.3) 272OOOGOOB 272OOOK102 272OOOK106 ...(KA*S)
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A ~. .. '4 6. ~ PLANT SYSTEMg;pggJgNg_ggNIBQb, ,.Np_JNj]ByMgNIAIJgN PAGE 28 AMBWERS -- VERMONT YANKEE .-87/07/14-HOWE, A.
ANSWER 6.10 (2.50)-
's. - ADS valves' remain as i s' (0.5) -b. ADS valves close (0.5) -~c. ADS valves close . (0. 5).
d. - ADS ' val ves open (0.5) e. ADS: valves open (0.5)
' REFERENCE.
VY. LOT-03-304,. Automatic Depressurization System, p.11,13,14, SCRO. Student Objectives 2 & 4 K/A 218000 K5.01'(3.3/3.8), K/A-218000 K4.04;(3.5/3.6), K/A 218000 A2.05f(3.4/3.6), K/A 218000 K6.02 (4.1/4.1),
-K/A 218000 A1.05 ~ (4.1/4.1)
218000A105 218000A205 218000K404 218000K501 21BOOOK602 c...'(KA'S) l, 1.
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7. PROCEDURES - NORMAL t_ABNQRMAl _EMERQENQY_AND t PAGE 29 I 50919L991Ce6_CQN16QL ANSWERS -- VERMONT YANKEE -S7/07/14-HOWE, A.
ANSWER 7.01 (2.00) a. 1. Scram and isolate the reactor (0.5) 1
'
{ 2. Use RCIC and SRV-71A to control reactor level and pressure (0.5) 3. Use the RHR system to cool the torus and when possible f or j shutdown cooling. (0.5)
(concept of each objective will be accepted for full credit)
b. FALSE (0.5) REFERENCE VY OP 3126, Shutdown Using Alternate Shutdown Methods, pgs 1 &3 VY LOT-09-013, Shutdown Using Alternate Shutdown Methods SCRO Student Objective A.1, A.4 K/A 295016 AK2.02 (4.0/4.1), K/A 295016 AK3.01 (4.1/4.2) 295016K202 295016K301 ...(KA'S) ANSWER 7.02 (2.00) 700 psig is the minimum RPV pressure that will produce adequate steamhms) flow through one SRV to li,,mit peak clad temperature to 2200 deg. F.[e.s) +1 r&Nr (or do aavee. odeyk con coolip) f* n The surge of steam flow lower fuel temperatures,gesulting from rapidtime providing additional depressurization to establish will a source of coolant i nj ecti on. (o.5) 4tvog,, sm-n REFERENCE VY OE 3102, RPV Level Control VY LOT-09-OO9, RPV Level Control, pg 16 SCRO Student Objective A.2 K/A 295031 EK1.01 (4.6/4.7), K/A 295031 G.07 (3.7/4.0) 295031 GOO 7 295031K101 ...(KA'S)
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LJ.__PBQCEgyBgS_;_N96D9L 2 @BN9BD9L1 _EdgBGgNCY_gNp PAGE 30 599 19LgGJC@L_CgNIBgL.
ANSWERS.-- VERMONT' YANKEE -87/07/14-HOWE, A.
ANSWER 7.03 (1.50)
. g g 4 0-47 (025 each f;r eny 3) -
t r a ny r au i r :d - 769 F-a a -p- r
- To inert the containment following'a startup or drywell entry (no longer,than 24 hrs) -.To de-inert (purge) the containment prior to a shutdown (no longer than 24 hrs) - Tu . timiy T:rhai e=1 An-ri-f;m.iivo m u r # s i l l e r.c qu i r ; ... , , t a *FF- %n REFERENCE VY OP.2115, Primary Containment, pgs 3 & 4 VY LOT-07-OO5, Turbine Generator Operation SCRO Student Objectives A.4, A.5, A.6, and A.7 K/A 223001 G.07 (3.7/3.8)
223OO1 GOO 7 ...(KA'S) ANSWER 7.04 (2.00)
(0.5 each)
e.. YES b. NO c. YES d. NO REFERENCE VY OP 2110, Reactor Recire System, pg 3 VY LOT-07-OO2, Approach to Criticality and Heatup SCRO Student Objective B.1 K/A 202001 K1.17 (3.1/3.3) 202OO1K117 ...(KA'S) ! ANSWER 7.05 (1.50)
-A c, e, /'[h41 (0.5 each, all req'd for full credit; points deducted for incorrect responses)
REFERENCE VYNPC; ON 3142 Recirc Pump Seal Failure p.1, LOT-09-OO2 SCRO Obj. A.2
'KA 202001 A2.10 3.5/3.9 202OO1A210 ...(KA'S) ,
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- _ - - - - ._ - . - - - _ - - - - - - - - - - - - - - - - - - - - - - ., . , ,7g_ PROCEDURES - NORMAL _ABNQRMAlt _EMERGENgY_AND t PAGE 31- ;88 Dig (QG1C@(_CQNI@Q( . ANSWERS -- VERMONT _ YANKEE -87/07/14-HOWE, A.
ANSWER 7.06 (3.00) 1. Perform ~OE 3100 SCRAM concurrently (0.25) 2. For the failed D/G, energize the bus as follows: a. check open the normal supply breaker (0.25) 6. manually start the D/G from the CR (0.25) c. close the D/G output breaker (0.25) ; d. if the D/G restart fails, energize the bus from the Vernon ti e (0.25) 3. Start or verify auto operation of the followings a. 2 service water pumps and SW2O closes (0.25) b. transfer of-station air compressor cooling water (0.25) c. restart station' air compressors A and B (0.25) d. Emergency DC equipment 1) turbine emergency bearing oil pump (0.25) 2)' emergency seal oil pump (0.25) 3) recire MG LO pump (0.25) 4) vital MG set shift to DC (0.25)
' REFERENCE VYNPC; OT 3122 Loss Of Normal Power p.1, LOT-09-OO6 SCRO Obj. 2 KA 295003 G.10 3.9/4.1, AA2.04 3.5/3.7 295003A204 295003G010 ...(KA'S)
ANSWER 7.07 (2.00) Enter OE 3103 DRYWELL PRESSURE AND TEMPERATURE CONTROL (0.5) entry condition of 2.5 psig exceeded (0.5) Entar OE 3100 SCRAM (0.5). A second entry condition requires procedure ruperformance since plant conditions have changed and actions may be j different. (0,5) ' REFERENCE VYNPC)DE 3100, LOT-09-OO1 p.15 SCRO Obj.B.1, B.2 OE-3103, LOT-09-OOB, SCRO Obj. A.1, A.2 KA 295024 G.11 4.3/4.5, G.12 3.9/4.5 ; 295024G011 295024GO12 ...(KA'S) i j I I l I l
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17, PAGE 32
- PROCEDURES - NORMAlt_ ABNORMAL _tEMERGENCY _AND ,
R_A_D__I O_L_O_G_ _I C_A_L_ _C_O_N_T_R_O_L_ ANSWERS - VERMONT YANKEE -87/07/14-HOWE, A.
, ANSWER- 7.08 '(2.00) Dscreasing reactor system pressure can cause greater fission product. release through the fuel clad (0.75) due to the high fuel rod internal pressure compared to the lowering outside reactor pressure during shutdown /cooldown evolutions. (0.75) Also since the mechanical vacuum pump is on line which does not exhaust to off-gas filters. (0.50) REFERENCE VYNPC; O.P. 0111 Shutdown to Cold Shutdown p.1, LOT-07-OOB, SCRO Obj. A.3 p.8
. ANSWER 7.09 ( 2.~ O0 ) '(This curve is an: indication of RPV level indication accuracy.)
Should this curve.be exceeded, RPV level will be essentially
. unknown.(1.0) Therefore, to ensure adequate core cooling , the RPV is depressurized and flooded. (1.0)
REFERENCE VYNPC; 'OE 3103, Drywell Pressure and Temperature Control. Procedure LOT-09-OO8 SCRO Obj. 2 KA 295028 K1.01 3.5/3.7, K3.01 3.6/3.9, K3.02 3.5/3.8, G.7 3.4/3.8 29502BGOO7 295028K101 295028K301 295028K302 ...(KA'S) ANSWER 7.10 (2.50) e. - reduce motor statting current (0.5)
- eliminate check valve slam (0.5) - minimize flow transients in the FW system (0.5)
b. Sample for radioactivity since it is possible f or condensate to leak into the CW system via a tube leak. (1.0) REFERENCE VYNPC,'OP 2172 FW System p.10, LOT-05-109 SCRO Obj. A.31 OP2170 Cond Syst p.9, OP 2180 CW Syst p.3 LOT-02-205 SCRO Obj. 6 KA 256000 G10 3.1/2.9 259001 A4.02 3.9/3.7, A2.01 3.7/3.7 256000G010 259001A201 259001A402 ...(KA'S) l __________________________a
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L 7, PROCEDURES - NORMAL _t ABNORMAL _t EMERGENCY _AND PAGE 33
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RADIOLOGICAL CONTROL.
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ANSWERS'-- VERMONT YANKEE -87/07/14-HOWE, A.
. ANSWER 7.11 (3'00)
.
a. to prevent draining the CST to the_ Suppression pool (1.0) b. No'(0.5) HPCI has a misoperation'in the automatic mode-OR- the turbine exhaust pressure trip has failed (1.0) HPCI should be. secured to prevent further damage (0.5) REFERENCE VYNPC, OP 2120 HPCI System p.4, LOT-03-303 SCRO Obj. 6.b OE 31:00 SCRAM, LOT-09-OO1 SCRO Obj. A.4 KA 206000 K4.01 3.8/3.9, 295031 EA1.02 4.5/4.5 206000K401 295031A102- ...(KA'S)
-ANSWER '7.12 (1.50) ' Standard RWP.(0.5) This is used for non routine jobs where more unknowns exist about radiation' levels and changes. (i . e. hi ' rad from N-16 in RCIC steam line) (1.0)
REFERENCE VYNPC; LOT-06-305 p.19,20 SCRO Obj.1,2 AP 0502 p.2, 3 , 15 KA 294001 K1.04 3.3/3.6, 294001 K1.05 3.2/3.7 294001K104. 294001K105 ...('KA'S)
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at__8DdlNigl8@l1VE_EBQCEDQ8ES t_CQND111QNSg _@ND_Lidll@llgN@' PAGE 34 4-
. ANSWERS -- VERMONT YANKEE -87/07/14-HOWE, A.
!- l-
' ANSWER- 8.01 (3.00) '
_e.'SCRO - station himself in front of CRP 9-7 to. check turbine performance, D/G performance, status of MSIV's, recirc system status ana: electrical transfers ~ ( 0. 5 ) gdad* '~7M the initial actions of_the CRO and panel.9-5 operations
'
(0.5)
- move to panels 9-3.and 9-4 to assess ECCS, PCIS and RPS operation, actions as required at these panels (0.5)
b. Shift Supervisor - locate himself in the control room to best monitor system and plant performance but far enough away to monitor the overall plant (0.5) ,
- direct the1 actions of his operators and ensure their !
proper actions .
(0.5)
- use of the EDP's and check that plant response is as predicted in the EOP's (0.5)
REFERENCE' VY Standing' Order #10 K/A 294001 A1.03 (2.7/3.7), A1.02 (4.2/4.2), A1.09 (3.3/4.2) 294001A103 ...(KA*S) ANSWER 8.02 (2.00)
(0.5 each)
a. 5 b. 4 c. 1 d. 3 REFERENCE VY LOT-06-301, Operations Department Administrative Procedures, Lesson 1: Vermont Yankee Local Control Switching Rules SCRO Student Objective 6 K/A 294001 K1.02 (3.9/4.5) 294001K102 ...(KA'S)
.
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' - I j . -8 1 _0DdlNIS18@llVE_E8QCEDyBESt _CQNDillQNSt _@ND_Lidil@llgN@ PAGE 35 AN8WERS -- VERMONT. YANKEE -87/07/14-HOWE, A.
ANSWER 8.03 (3.00) a.~- reactor suberitical (0.3)
-
rx plant < 212^F (0.3)
- rx coolant system. vented (0.3) - no activity being performed which can reduce shutdown margin- . -
below that specified in TS (0.3)
- irradiated fuelf is not'being moved in the rx building (0.3)
or $ad cask b. - At.least one door in.each access opening is closed (0.5)
- SBGT. system is operable (0.5) - All reactor building automatic ventilation system iselation valves .are op rable or are secured in the isolated position. (Also building intact (0,5)
REFERENCE ! VY Technical Specifications, pgs 3, 131 & 132 VY LOT-03-210, Secondary Containment and SBGT Technical Specifications SCRO Qgjective 1, CRO Ob'Jective 5 KA 290001 G5 3.3/4.2 290001 GOO 5~ ...(KA*S) ANSWER. 8.04 (3.00) 3.5.C.2. 'Can operate for 30 days with RHRSW pump A out of service (0.75) 3.10.B.1 For one D/G inop. must satisfy 3.5.H.1 (0.75) 3.5.H.1 -With one D/G inoperable and all core and containment cooling subsystems powered from the other D/G not operable,(1.0) a shutdown shall be initiated and the reactor shall be in the cold . shutdown? condi tion; wi thin -24- hrs. (0. 5) !
. c+.: ~pmtsyw& +pFm ,- -. 3 m: -
REFERENCE VYNPC; T/S 3.5.C.2, 3.10.B.1, 3.5.H.1 l
{;&;} Q Q RT W - -- * l C b MQW .=
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8 __@pMJNJglB9IJVg_PBQGEpyBEg,_ggNp]IJgNgi_@Np_(JMJI@IlgNS PAGE 36 AN$WERS -- VERMONT YANKEE -87/07/14-HOWE, A.
.. n i
- ANSWER 8.05 (2.50)
Yea provided the following trips are available (1.0) 1. Mode switch to S/D scram 2. Manual scram 3. IRM HI Flux scram 4. SRM-HI Flux scram non-coincidence 5. SDV hi level scram 6. No more than two rods withdrawn and those may not be face or diagonally adjacent
, (6 & O.25 each)
REFERENCE-VYNPC; T/S Table 3.1.1 note 12, i ANSWER 8.06 (3.00) Psr T/S 3.2.B table 3.2.2, the RCIC isolation valves must be shut (1.0) and RCIC declared inop (0.5). May operate for 7 days provided HPCI is operable per T/S 3.5.G.2 (0.5). Per T/S 4.5.G.2 must demonstrate HPCI oparable immediately and daily thereafter (1.0).
REFERENCE VYNPC; T/S 3.2.B table 3.2.2 note 3, 3.5.G.2, 4.5.G.2 KA 223002 G.5 3.1/4.1 223OO2 GOO 5 ...(KA'S) l ANSWER 8.07 (2.50) a. (0.5 each) 1. Al er t 2. None 3. Unusual Event , 4. Site Area Emergency I b. TRUE (0.5) l s,gg ~,A.,,.g;m;;g;>' "#9
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AN5WERS -- VERMONT YANKEE -87/07/14-HOWE, A. I
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l I ANSWER 8.08 (2.00) i i i Ycs (0.5) the scram was produced by a high flux instead of l by the MSIV closure (0.75). If this happens, a safety limit ] is assumed to have been exceeded (0.75). j
REFERENCE VYNPC; T/S 1.1.c. p.6 ANSWER 8.09 (2.00) Yes (0.5). This work is a core alteration (0.75) and a SRO is required (withnoconcurrent.activitiehper T/S. (0.75).
REFERENCE VYNPC; T/S 6.1.D.3, Definition 1.B ANSWER 8.10 (2.00) a. VY limits of 250 mr/wk and 1000 mr/qtr (0,5) 250 mr/wk less than 400 mr remaining in the quarter (0.25) 250 mr / 150 mr/hr = 1.66 hr (0.25) b. 10CFR2O limits of 1250 mr/qtr and 5(N-18) total (0.5) 650 mr remaining in the quarter is less than 750 me remaining lifetime (0.25) . 650 mr / 150 mr/hr = 4.33 hrs. (0.25) REFERENCE VYNPC; LOT-06-305, Table 1 SCRO Obj. 1.
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.------- :------ ----------
05.01 3.00 AXAOOOO460
'05.02- c 2.00 AXAOOOO462 05.03. 3.00 AXAOOOO463 .05.'04 2.00 AXAOOOO464 05.05 1.00 AXAOOOO465 05.06 3.00 AXAOOOO471 05.07 3.00 AXAOOOO473 05.08 3.00- AXAOOOO488 .05.09 3.00 AXAOOOO489 05.10 2.00 AXAOOOO492 ------
, 25.00 06.01' 1.50 AXAOOOO476 06.O2 3.00 AXA0000478
~O6.03 2.50 AXAOOOO479 06.04 '2.00 AXAOOOO480 06.05 3.00 AXAOOOO481 06'.06- 2.50 AXAOOOO482
, 06.07 2.50 AXAOOOO483 06.08 2.50 AXAOOOO484 06.09 3.00 AXAOOOO485 06.10 .2.50 AXAOOOO487
------
25.00 l 07.01 2.00 AXAOOOO454 07.02 2.00 AXAOOOO455, 07.03 1.50 AXAOOOO457 07.04 2.00 AXAOOOO459 07.05 1.'50 AXAOOOO466 107.06 3.00 AXAOOOO467 07.07 2.00 AXAOOOO468 07.08 2.00 AXAOOOO469 07.09- 2.00 AXAOOOO470 07.10 2.50 AXA0000474 07.11' 3.00 AXAOOOO475
.07.12 1.50 AXAOOOO490 s' ------
25.00 i 08.01 3.00 AXAOOOO456 1 08.~ O2. 2.00 AXAOOOO458 l I 08.03 3.00 AXAOOOO477
:08.04 3.00 AXAOOOO493 08.05 2.50. AXAOOOO494 08.06 3.00 AXAOOOO495 08.07 2.50 AXAOOOO496 08.08 2.00 AXAOOOO497.
08.09 2.00 AXAOOOO498
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'dVERMONT YANKEE t w 'TM d4 e/UCLEAR POWEft s.CORPORATION _
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RD 5. Box 169. Ferry Roa . Brattleboro, VT 05301
.74 ,,
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1 July 22, 1967
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9 TDL-87.74 (802)257-5271 ,
'
Robert Keller, Chief .
,. Projects Section No. 1C ' ,
Division of Reactot Projects ,
'
Nuclear Regulatory Commission ' Region I ) < s , i 63k Park . // /enue s.
, t. , King of! Mussia, PA 19406 ij 1
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Dear Mr. 'Kellet:
I c ' In ockrdrnce with'NL1EG 1021, we ere submitting a detalked review of ' the U.S.N.R.C O ba'ctor Opehtol and Senior Reoctor Operator Licaisp fxarts' 4hich were conducteddti Verment Yank 6e 50 July 14, 19b7. The details of this Ieview are contain'ed in the attachments and should 0.learly delineate ou'r concernn i,ith the exam $.ations.
/ f you or your staff havd any questions about this cuatalalj please do not hesitate to call. . j
.s
,
,
very truly yours, ( x
'
E VERMONT YANKEL NUCLEAR POWER CORP.
y . i N \ W /rffN b ' Warren P M'.;rgdy
'llce Pru[ ice'ac \/
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Manage 7 c.5'yoerations ,
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REACTOR OPERATOR EXAM SeJtew
,s sq I uestion Q 1.')3a: Nuinber 5.should be acceptable. The concern during PCIOMR ramping is a cladding cracking due to pellet to clad int </ action. Le T - oz. - ws pp tr + 9 (_ r.L 4 pu k h ul E . Lud aced Question 1.05: Key contains a math error, specifically: (t) =.Poe UN ' '
n 40 = 0.04et/30 sec 4; in(1000) = t/30 se.
, 6.9078 = t/30 sec
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207.23 sec = t Answer key give 460.5 sec
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Question 2.Ola: Should also accept the answer "to limit axial thrust". This limit was given in L0T-03-007, Rev. 1, Page 23 of 32.
Question 2.02a: Rod will scram with normal scram time. See Tech Spec bases,
*
Control Rod Accumulators, Page 78.
Question 2.02b Scram slower than normal should be accepted.
% '(; ' Question 2.06: Should also allow continuity ammeter indication insida 9-5 T panel as referred to in LOT-05-409, Page 10 of 14.
Question 2.07bs Operation of RCIC does not significantly affect the stay fill leakage so such leakage is not an " adverse effect" of RCIC operation. Should not be considered part of the answer.
Question 2.08a MCC8A and 9A should be acceptable. These MCC's feed the RPS MG sets. Refer to LOT-03-108, Transparency 1.
Question 2.08b: Half isolation should also be accepted. OP 2134, Rev. 8 lists the half isolation as a response. Refer to the precaution sec-tion. Refer also to LOT-03-211, Transparency 9. f Question 3.03: Rather than discussing the fact that the switches are 2 or 3 position, a reasonable response would refer to the MSIV and sample valves as air solenoid valves which will auto open if not in closed position. Additionally, the other valves are motor 7per.?ted valves which remain in position up 1 resetting of the iso l.ation. IRef,ere.n.ces sg ed . Thb rer ktc.s v/ c. L..am,A Question 3.07: The questi.'n specin'ically asks for FWLC response, but the key includes fee 4 pump a'd turbine response. Since this was not ; asked for, it si.wld at be considered part of the ar,swer. I
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Question 3.08: Should also accept a RBCCW pump start after one minute. Refer to LOT-05-304, Rev. 2, Page 12.
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REACTOR OPERATOR EXAM - (COWTINUED)
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Question 3.09: A Group III isolation includes a SBGT start and Reactor Building ] Ventilation System isolation. Refer to LOT-03-211, j Transparency 9. l Question 3.10a: This question has a number of problems which when combined, could lead to confusion of the candidate. Some of these prob-lems are: 1. MPR setpoint is not normal (should be ~ 969 psig)
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2. Bypass opening jack does not have setpoint indication, it has stroke (position) indication.
3. Speed / load changer demand is normally ~ 90% for full generator load.
4. To answer the question, many assumptions can be made as to j actual plant conditions, all of which will have different ! outcomes. Some of these assumptions / outcomes are listed-below. !
5. If I am reading the question correctly and making the most reasonable of assumptions, I cannot get the answer required by the answer key.
Assumption / Expected Outcomes of: a. Bypass opening jack demands 10% on the bypass valves.
NOTE: MHC functional diagram from lesson plan is attached.
1. Bypass opening jack "setpoint" of 0: Assume implies a setpoint less than that of EPR and MPR meaaing BPOJ has control of P-1 and since BPOJ is a manual valve positioner and not a regulator, a "setpoint" of 0 would result in almost full CV and BPV opening, therefore Rx pressure drops rapidly and Rx power follows. At 800 psig, a PCIS Group I isolation occurs.
NOTE: Since BPOJ setpoint is in the given material, this would be the same response in Part b.
2. Bypass opening jack setpoint of 0: Assume implies BPOJ stroke of 0 BPOJ demands 10% on bypass valves- assume implies 10% stroke on bypass valves.
This would probably be the most reasonable of assumptions.
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Expected outcome ]
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Since the Pressure Torque Bar (P-1) is controlled by the device with the lowest pressure setpoint which is indicated . by the device with the highest-indicated stroke, and con- ) ditions are given for 100% power meaning EPR stroke ~ 90%, then a stroke of 10% on BPOJ will not result in any effect on reactor power, control valve position, or bypass valve q position.
3 BP0J setpoint of 0: Assume implies BPOJ stroke of 0- BPOJ demands 10% on bypass valves- assume implies 10% position on +1 BPV.
In this case, the BPOJ stroke has increased such that it is now just slightly greater than that of EPR and MPR and it has therefore taken control of P-1 from EPR and MPR. Since the BPOJ is not a regulator, it is manually positioning valves more open than EPR/MPR. Pressure will slowly decrease causing power to decrease and there will be no change in BPV or CV position. Pressure will continue to decrease until the 800 psig in run PCIS Group I isolation.
4. BPOJ setpoint of 0: Assume implies BPOJ stroke of 0 BPOJ demands 10% on bypass valves- assume implies +1 BPV full open (10% steam flow).
This response will be the same as +3 except that pressure will drop rapidly to the PCIS Group I.
5. BPOJ setpoint of 0: Assume implies BPOJ stroke of 0 BPOJ demands 10% on bypass valves- assume reading between the lines, that examiner just wanted a BPV to open and show CV response. Therefore, assume BPV *1 fails 10% open without BPOJ movement. (Very similar to testing BPV's) In this case, I can support the answer key because the BPOJ has not taken control from EPR/MPR. Therefore, +1 BPV opening would cause a pressure decrease resulting in EPR positioning of P-1 to close control valves a comparable amount. The end result would be:
#1 BPV open 10%
Rx pressure- slight dip and return Rx power- slight dip and return Control valves- slightly fu-ther closed 6. Same assumptions as above (+5) except +1 BPV fully opens.
Expect the same response and a possible PCIS Group I isola-tion on high steam line flow depending upon speed of BPV opening and amount of steam line flow overshoot due to BPV opening and EPR/MPR response . lag time.
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REACTOR OPERATOR EXAM - COWTINUED
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My recommendation is that: 1. The confusion of initial conditions be taken into con-sideration during grading.
2. Part "a" be either deleted and point valve added to part
"b" or any response be acceptable for part "a" due to the numerous assumptions which can be made and that the exam answer key is incorrect for the most reasonable assump-tions.
3. This question be deleted from the exam bank; it is apparent that this is an EHC question modified for an MHC system.
One glaring difference is that in EHC, the BPOJ has direct positioning control of BPV's without effectina the EPR/MPR function. In MHC, the BPOJ can only take control from EPR/MPR and as such, can contrcl both control valves and bypass valve positions without reference to reactor pressure.
Question 4.02: Adequate core cooling is defined in the Emergency Procedure Guidelines as maintaining peak fuel clad temperature at or below 2200*F. Candidates should not have to make both state-ments since it is redundant.
Question 4.03: Should also accept fuel cask movement. Refer to VY Tech Specs, 3.7.c.1.0, Page 132.
Question 4.08: Question asks for 3 of 4 reasons to bypass interlocks which keep containment vent and purge valves closed when the mode switch is in run as per OP 2115 and M00 Directive 79-4, Rev. 2.
Answer key lists a paraphrasing of OP-2115 admin limit +8 which is a list of four restrictions which are placed upon the opera-tion of Containment Vent and Purge Valves. By OP 2115 and M00 Directive 79-4, Rev. 2, only two of these items require the , bypassing of the identified interlocks. Those are: {
- Inert the containment following s/u or d/w entry - De-inert the containment prior to s/d The other two items on the list do not require bypassing interlocks for valve operation. Refer to LOT-03-211, Transparency 10 and 11.
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SENIOR REACTOR OPERATQR
e Question 5.03: See Question 1.03.
Question 5.05: Answer 1 and 2 are both correct. Any drop in feedwater tem-perature (increased subcooling) will increase the critical power. uor -cz - 3 s z., pg n. cr ta pu Jelear w) 6. L: Mod} Question 6.03: See Question 3.03.
Question 6.07: Should accept Group III as being the same as a start of SBGT and trip reactor building ventilation. See Group III reference for RO exam.
Question 6.08: See Question 3.08.
Question 7.05: Answer key is incorrect. According to ON 3142, seal tem-peratures increase (answer h) rather than decrease (answer g).
Question 8.03: See Question 4.03.
Question 8.09: ' Since the question referred to a " dedicated SR0", the candidate should not be required to respond with "no concurrent activities...". i EVL 870721.1 -5-l
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ATTACHMENT 4 NRC RESPONSE TO FACILITY COMMENTS The following represents the NRC resolution to the facility comments (listed in Attachment 3) made as a result of the current examination review policy . Only those comments resulting in significant changes to the master answer key, or those that were "not accepted" by the NRC, are listed and explained below.
Comments made that were insignificant in nature and resolved to the satisfac-tion of both the examiner and the licensee during the post examination review are not listed (i.e.: typographical errors, relative acceptable terms, minor set point changes).
Reactor Operator Examination Question 1.03e: Comment accepted.
Question 1.05: Crmment accepted.
Question 2.01a: Comment accepted.
Question 2.02: Comments accepted. Information is not contained in the material referenced for development of this question.
l Question 2.06: Comment accepted.
Question 2.07b: Comment accepted.
Question 2.08a: Comment accepted.
Question 2.08b: Comment accepted. Alternate answer is not contained in the material referenced for development of this question.
Reference material does not give a clear explanation of the complete answer to this question. Also incorporated comment on Question 6.07 of the SRO examination.
Question 3.03: Comment accepted. Alternate explanation is not included in reference material provided for examination development.
Question 3.07: Comment partially accepted. Feed pump and turbine response is not required for a complete answer but stating that the l feed pump and turbine trip setpoint is not reached is an i incorrect response.
Question 3.08: Comment accepted. Alternate answer is not contained in the material referenced for development of this question.
i Question 3.09: Comment accepted.
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ATTACHMENT 4 2 Question 3.10a: Comment partially accepted. Question 3.10a deleted due to confusion in given parameters. Reference material provided for examination development does not provide information on normal parameters and system response to changes in these parameters.
Question 4.02: Comment partially accepted. Either "to assure adequate core cooling" or "to limit peak clad temperature to 2200 deg F" is required for a comp 7ete answer. Alternate wording is not provided in material referenced for development of this question.
Question 4.03: Comment accepted.
Question 4.08: Comment accepted. Only two items required for a complete answer.
Senior Reactor Operator Examination Question 5.03: See Question 1.03.
Question 5.05: Comment accepted.
Question 5.07: See Question 1.05.
Question 6.01a: See Question 2.01a.
Question 6.03: See Question 3.03.
Question 6.07: Comment accepted. Also see Questions 2.08a and 2.08b.
Question 6.08: See Question 3.08.
Question 6.09: See Question 3.09.
Question 7.02: See Question 4.02.
Question 7.03: See Question 4.08.
! Question 7.05: Comment accepted.
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Question 8.03: See Question 4.03.
Question 8.09: Comment accepted.
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ro t ATT/.CHMENT 5 SIMULATION FACILITY FIDELITY REPORT Facility Licensee: Vermont Yankee Nuclear Power Corporation Facility Licensee Docket No.: 50-271 Facility Licensee No.: DPR-28 Operating Tests administered at: Vermont Yankee Operating Tests Given On: July 15-17, 1987 and August 25-27, 1987 1. During the conduct of the simulator portion of the operating tests-administered July 15-17, 1987, the following apparent performance and/or
~ : human factors discrepancies were observed: --
Delays (approx. 5 brs.) in conducting the exams on 7/15/87 due to simulator repairs.
- Problems encountered when attempting to reinitialize the simulator between scenarios resulting in delays - i.e., a 45 minute delay between the first and second scenarios on 7/15/87 and a 30 minute delay between the first and second scenarios on 7/17/87.
- Spurious generator lockouts which interrupted the scenarios and affected the continuity of the examination - specifically in the first scenario on 7/15/87 resulting in an interruption of the scenario while the simulator was reset; in the first scenario on 7/17/87 resulting in omission of several planned events which required a third scenario to be prepared; and while the simulator was in freeze between the first and second scenarios on 7/17/87.
- Spurious FWRV lockups, Scoop Tube lockups, a Hotwell Level controller failure, and miscellaneous annunciators throughout the operating examinations which confused and distracted the candidates.
- Loss of the audible tone on several annunciator panels, spurious annunciators (some visual only) and other problems with the simulator performance that obscured the candidates' concept of the simulator as the real plant.
- Unstable drywell/ torus differential pressure during certain Initial Conditions (ICs) which distracted the candidates from planned " evolutions.
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ATTACHMENT 5 2 l
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Possible problems with the modelling of drywell pressure response for various events - i.e. during the first scenario on 7/16/87 the plant
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j scrammed on drywell pressure that had not yet reached the trip set- l point and during the second scenario on 7/17/87 drywell pressure i spiked to 40 psig when a safety valve lifted then immediately I reseated. {
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Failures of the HPCI. system - i.e. failure of the governor valve ' during a surveillance which caused the SRO to commence an unplanned ' shutdown and spurious HPCI turbine trips during an attempted full flow test. j
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Numerous surges in reactor power, core flow, recirc flow, jet pump flow and oscillations of the FW startup valve that caused the SRO to commence a plant shutdown during the first scenario on 7/17/87.
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Failure of several inserted malfunctions which caused deviations ] from the preplanned scenarios - i.e. HPCI controller failure, loss { of instrument air and EDG slow start.
- Differences between the plant and the simulator that obscured the candidates' concept of the simulator as the real plant - i.e., indications of containment parameters that exist in the plant but are not modelled on the simulator, individual rod scram capability not functional on the simulator.
- Personnel in adjacent computer room on 7/15/87 that distracted the candidates and the examiners. I 2. During the conduct of the simulator portion of the operating tests administered August 25-27, 1987, the following apparent performance and/or human factors discrepancies were observec: j i
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A startup transformer malfunction did not work properly. The supply j and output breakers should have opened, and loads from the j transformer which should have been lost functioned normally. The ) expected annunciators did alarm. 1
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Near the end of one scenario, at relatively low power, the simulator reset to a near full power condition. The simulator was frozen and the candidates left the simulator.
- A one hour delay occurred between scenarios after a spurious generator lockout occurred during setup. !
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During several scenarios the desk section indicating lights blinked i briefly and distracted the candidates and examiners.
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