IR 05000219/1985012
| ML20127D768 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/03/1985 |
| From: | Joshua Berry, Keller R, Kister H, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20127D757 | List: |
| References | |
| 50-219-85-12, NUDOCS 8506240330 | |
| Download: ML20127D768 (78) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N (OL) FACILITY DOCKET N FACILITY LICENSE N DPR-16
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LICENSEE: GPU Nuclear Corporation l P. O. Box 388 ' Forked River, New Jersey 08731
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FACILITY: Oyster Creek Nuclear Generating Station l EXAMINATION DATES: April 8-12, 1985 CHIEF EXAMINER: [. 5 ,79 f D. in ge! R Engineer (Examiner) / Dite ' REVIEWED BY:
! c/M : . Be'rry', Lead ctor Engineer (Examiner) ' trate ARfYb R. M. Keller, Chief, Project Section 1C thkW ~
Date APPROVED BY: _ B. TisterkChief, Project Branch N h2b H'. / 7 ate < SUMMARY: Operator Licensing Examinations were conducted at the Oyster Creek Nuclear Generating Station during the period of April 9-12, 198 Eight i Reactor Operator candidates and three Senior Reactor Operator candidates were administered written and oral examinations. All candidates successfully passed both the written and oral examination During the oral examinations, the Senior Reactor Operator candidates demon-strated an overall strength in their ability to use Technical Specifications and Emergency Plan procedures. During grading of the written examinations, an overall strength was noted in the area of Plant Design / Instrument Controls, for both Reactor and Senior Reactor candidates. No generic weaknesses were noted i for either the Reactor or Senior Reactor Operator oral or written examinations, t 8506240330 850604 PDR ADOCK 05000219 0 PDR w ________---_____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ ___ _ _ ._ - _ _ _ - - _ _ _ - - _ - - _ - _ _ - _ _ _
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REPORT DETAILS, Type of Exams: Replacement
- 2 .- Exam Results
, e l R0 l SR0 l l Pass / Fail l Pass / Fail l l l l
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1 l I I l Written Exam 1 8/0 1 . 3/0 l
- 1 I l l I -- 1 I i 10ral Exam I 8/0 .I 3/0 'I I I -l i I I I I-10verall l 8/0 1 3/0 l-
' I I I I l Chief Examiner at Site: D. Lange, USNRC, Reactor Engineer Examiner i Other Examiners: B. Hajek, NRC Consultant Examiner i- D. Hill, EG&G Idaho, In . A. Mendiola, USNRC (OLB) Examiner Trainee } Dyster Creek Entrance Meeting:
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- 4 NRC Attendees
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D. Lange, USNRC Region I, Reactor Engineer Examiner B. Hajek, NRC Consultant Examiner . Facility Attendees
- Rod Davidson, Training Supervisor Derrick Wilson, Training Instructor
i An entrance meeting was. conducted immediately. fo11o' wing the start of the j R0/SR0 written exam. -A tentative schedule for oral exam assignments was , discusse The two hour. exam review was scheduled at the training 1-department following completion of -the exam. A tentative exit' meeting was set up for 8:00 AM Friday morning, April 12, 1985.
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i l : Summary of strengths noted on oral exams:
' An overall_ strength was noted in all candidates ability to use piping and instrument ' drawings and their' ability to use normal and emergency procedure '
The SR0 candidates demonstrated a strong safety awareness to overall plant conditions and abnormal event Candidates did very well using the control room plant reference materia The SR0 candidates did very well using__ Technical Specifications .and emergency plan procedure . Exit Interview Details Personnel Present at' Exit Interview: NRC Personnel D. Lange, Chief Examine J. Berry, Lead BWR Reactor Engineer (Examiner) i A. Mendiola, OLB, Examiner Trainee W. Bateman, Senior Resident Inspector !
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NRC Contractor Personnel
B. Hajek, NRC Consultant Examiner-Facility Personnel P. B. Feidler, Vice President and Director, Oyster Creek I R. Davidson, Operations Training Manager D. Wilson, Training Instructor (Initial Operator Licensing) W. Stewart, Operations Manager D. Holland, Supervisor of Licensing \ , '
-Summary of Comments made at exit interview:
The Chief Examiner advised .the facility of the preliminary results 'of the i oral examination ' l l The Chief Examiner noted the generic strengths observed during - the SRO '
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The Chief Examiner noted a problem area, with the noise._ level of facility personnel in the control room, during the oral exams. This created a problem, for both candidates and examiners, to a point where examinations had to be stopped and.the problem brought to the Senior Shift Supervisor's attentio The Chief Examiner also noted that certain control room piping and instrument diagrams needed to be replaced' due to their worn conditio The Chief Examiner thanked the facility for the use of the conference room and commented on the cleanliness of the plan The Training Department submitted the facility comments for both the R0 and SR0 written exams. The facility wanted an indication of when .the next ' round of Operator Licensing Certificate presentations would be conducte The NRC Lead Examiner explained how this would be accomplished.
, The facility noted that the problem areas identified during the Oral Exams
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would be addressed.
Attachments: . Written Examination and Answer Key (SR0/RO)
' Facility Comments on Written Examinations Made After Exam Review and NRC ,
Resolutions to Those Comments
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Nuclear m w an e Subject: Exam Comments April 11,1985 From: R. Davidson Location: Operator Training Manager Forked River To: D. Lange NRC Examiner General C_omments Oyster Creek especially in therecognizes the difficulty of writing a valid license exam area of procedure , We have recently spent about 400 manhours developing the six exams for the Licensed Operator Requalifica program and experienced many difficulties in this are However, we have some concerns about Section 7 of the SRO exam . our concerns lie in the selection of the procedures to be examine ForGenerally, example, the questions on EPIP-7 asked detailed specifics of how to trea injured during workers emergency and also condition guidelines for extremities and skin exposu but do not need to memorize its contents or reproduce parts o . A question on ASN-7 asked what plant parameters must be checked wh unexplained reactivity change occurred at powe been very few, if any, unexplained reactivity changes and ifwere oneTo our knowle experienced, taken in accordance thewith Reactor Engineers would be notified and appropriate the procedur this question was unreasonaol To expect the SRO's to respond to knaning why cautions and limitations are there.Some e que For example, one question asked them "to state the two precautions and limitations prior to putting the static charger in service."
Thisisisunreasonacl a relative cbscure procedure and to expect regurgitation Some questions required immediate operator actions where ynone are procedur memorization.Our training program stresses the use of procedures but not rote A0000648 8 83 _ _ _ _ _ _ _ .
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. , ' , OYSTER CREEK NUCLEAR EERATING STATION NRC EXAMINATION OF APRIL 9,1985 EXMITERS: RO - BRIAN HAJEK . SRO - DAVID LANZ FACILITY CDMENTS Specific _ E;xamination Que.stion_ Comme _n RO Examination .Section. _1 1.01 Major part of answer should be towards the periphery and not requim towards the top since the peak shifts axially with core age / rod positio .03 a. part of answer - should accept inadequate volume in the Acetnulator to give proper scram insertio b. expand band to ll50(Proc. 302.1)
1.04 a. Also acceptable - Negative reactivity due to doppler coefficien .05' b. Answer key should be changed to reflect the fact that to increase flow it is less efficient to change ptsnp spee .09 a. Should accept any transient which may impact pressure and/or temo. The judgement as to which is most severe is open' for discussio Section 2 \ 2.07 psid due to modification 2.11 b. Answer key changes .to Block of Low flow /Hign temp trip Manually start the backup ESW ptn . Section 3 , 3.10 Should also accept valve being pinned as reason for not movin .
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4.02 Other possible answers such as loss of feedwater heater i 4.05 Ans. Breakers remain closed
; Breakers may trip on regaining D .06 Answers should be for turbine trip not generator tri Feeowater heaters may not be on at 30E . . (Step 4.19 Procedure 201.3)
l Possible answer - Reactivity increase due to pressure spike from step valve closule.
. 4.07 Answer should rot include 100# 4.08 First part of answer is not necessarily an immediate actio Suggest it be delete l 4.09 Additional answer is that the oncoming operator must be alert and t
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- SRO EXAMINATION i
! S_ection 5 1 5.07 Voids don't change during this transien Presence of void is the answer.
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' Rod worth at power lower due to volds XO will also be the primary ' l effect of turning power.
) 5.08 a & b One heat source-is the control rods & NI's.
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t 6.02 High drywell pressure setpoint by tech specs is 2.4#. , 6.06 - thit substations are only 460V breakers supplied by 125V DC for :
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Should not grade on position indication.
! , , Section 7 7.02 Pressure greater than 850# prior to going to range 10 i Tech spec answer should be acceptable ] l Proce& re and alarm 1 ' . 7.03 . Also vibration induced failure of tubes ~ Secondary is Fire Brigade - i 7.04 Other loads on 24V DC should be acceptable
- Other cautions in proceW re
, 7.06 Qualified first aid member only required answe ;
- Coordinator should be acceptable
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] Hospital should be acceptable 7.07 Level restored should be acceptable f
; 7.08 Other reasonable answers % should not be required
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, Section 7 (Continued)
7.09 b. More answers than listed 7.10 More answers than listed 7.11 Alert and coherent should also be acceptable (Procedure 106 Step 4.3.1, Pg. 24) Section 8 8.04 Exact limits should not be required to be memorize .05 a. When fire watch requimd b. Comunications 8.06 This 15 day requirement is not asked for in the question and should not be require .11 a. Question asks for the temperature, not a definition, therefore tnere could be confussion on this par b. Should also accept 500F T for meirc pump start and 1000F Heatup/Cooldown
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NRC Resolution to Facility Comments (R0 Exam) Category 1 e 1.01 Near the Top of the Core is not required for 1.1(2).
1.03 a, b Comment accepte .04 (a) Add, negative reactivity due to dopler coefficient during normal operating condition .05 (b) Not accepte This is true for decreasing flow, but not for increasing flow. Will consider candidates answer on pump law .09 (a) Considered during gradin Category 2 2.07 (c) Accepted with documentation provide Due to NO modificatio .11 (b) Comment accepte Category 3 3.10 Comment accepte Will accept, inadvertently left pinne Category 4 4.02 Alternate reasonable answers accepte .05 Comment accepted if candidate describes correct logic sequenc .06 (a) Turbine Trip accepte (b) Considered during grading if candidate assumes that the Heaters are not in service at 30*4 powe .07 (c) Comment accepted if candidate says that the MPR is kept higher than Rx. Pressur .08 (b) Comment accepted and point value re-distributed for remainder of answe .09 Alert and coherent accepted as an additional answer.
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NRC RESOLUTION TO FACILITY COMMENTS (SR0 EXAM) Category 5 5.07 - The presence of voids, during this transient is the acceptable answe .08 - The heat source, of control rods and N.I., will be accepted in addition to gamma and beta heating.
, Category 6 6.02(a) - 2.0# is accepted, 2.4# will be accepted if candidate identifies this change from 2.0# 4 2.4#. 6.06 - Unit substations - accepte Will not deduct points for improper position indicatio Category 7 7.02 a) Correct answer is, pressure > 850# prior to gaining to range 10 on IRM' b) Considered during grading if candidate explains responsibil-itie c) Caution in precedures accepted in addition to (2) alarm condi-tion .03 b) Maintenance of any ht. exch. assoc. prob. accepted if candidate explains prob with isolation valve c) Fire brigade response accepte .06 a) Reasonable answers accepted if candidate identifies that per-sonnel must be qualified in first ai b) Accepted, c) Accepte .07 a) Considered during gradin ' 7.08 a) Reasonable answers, affecting reactivity change, considered dur-ing grading, b) First immediate action of a rapid reduction in Recirculation flow accepte .09 b) Additional answers, only if correct as to actions performed before leaving control room, accepte .10 - Considered during gradin OFFICIAL RECORD COPY OC RESOLUTION - 0001. /07/85 . .
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NRC Resolution to Facility Comments 2 7.11 - Alert and Coherent is an acceptable answe Any (2) of the following three answers are correc . Alert and coheren . Fully qualified for that shift positio . Fully aware of plant condition Category 8 8.04 (b) Question asked for Limits, 110'.' of Federal and Admin. Limits considered during gradin .05 (a) Not accepte (b) Not accepte .06 - Candidate is not required to say 15 day requirement. The section, and most Limiting LCO is require .11 (a) Considered during gradin (b) Temp. thermal units for heatup and cooldown accepted as examples of temp requirements for various operating condition . D OFFICIAL RECORD COPY OC RESOLUTION - 0002. /07/85 ___ _ ___- _ _ _ _-_-__-_- _ -_- _ _ - -_ ... . _ _ _ . . . __ . _ _ _ _ _ _ _ . _ .
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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: OYSTER CREEK ________________ REACTOR TYPE: BWR _____-GE2 ____________________ DATE ADMINISTERED: 85/04/09 _________________________ EXAMINER: LANCE, _________________________ APPLICANT: ____ h [ [ [ [ _[_______ INSTRUCTIONS TO APPLICANT: __________________________ Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer sheet 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 start % OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY ________ ______ ___________ __.._____ ______________
_I"I"I_00 ___ I _II__0 ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS - I __ _ I ___________ ________ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _ I b__ _ I ___________ ________ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL _ Ibb__ _ I ___________ ________ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.00 100.00 TOTALS ________ ______ ___________ ________ FINAL GRADE _________________% All work done on this examination is my own. I have neither given not received ai ~ SEPLfC55T 5~5iG55TURE~~~~~~~~~~~~~~
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______________ l GUESTION 5.01 (2.00) l NOTE: Answer the following questions from a theoretical standpoint, not l from an Oyster Creek system design standpoin a. With the plant operating at 90 % power , extraction steam to one string of low press. feeduater heaters is removed. An engineer, observing that turbine load increased by 15 MWe after the extraction steam removal, concludes that this action has improved the plant's thermodynamic efficiency (NOT heat rate). Do you agree with this conclusion? Explain your answer full (2.00) OUESTION 5.02 (2.25) For each of the events listed below, state which reactivity coefficient will respond first and if it adds positive or negative reactivity.
! a. Relief Valve opening at 100 % power . (0.75) 6. Rod drop at 100 % powe (0.75) c. Isolation of a feedwater heater string at 75 % powe (0 75) GUESTION 5.03 (1.50) If the Main Condenser and associated systems were absolutely AIR TIGHT would there be any need for the Steam Jet Air E. lectors during full power operation ? (Explain your answer). (1.50) GUESTION 5.04 (3.00) Indicate whether the following will INCREASE or DECREASE reactivity during operation AND briefly EXPLAIN 9h a. Moderator temperature increases while below satruation temperatur (.75) b. Fuel temperature increase (.75) c. Loss of a feed 9ater heate (.75) d. A sudden reduction in reactor primary system pressur (.75) i
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GUESTION 5.05 (2 00) Concerning the term NPSH ; How would you account for the fact that the NPSH is different for a recirculation pump with the plant operatin3 at 100 % power versus 4 % power ? Include in your answer at which power level will tha NPSH be' hi3her, the operating conditions that are affecti.3 the chan3e' and the reason wh (2.00) GUESTION 5.06 (2.25) s. Following a Standby Liquid Control system initiation, from rated pow'er, list five (5) REACTIVITIES that the~ system must overcome to provide its tended functio (1.25) b. What is the bases for havin3 a minimum and maximum injection time re-quirement for the SDLC (1.00) GUESTION 5.07 (2.00) ~ What is the operational' requirement and main purpose for having the Rod Worth hinimi:er ? What is one of the principle reasons it is not required above 10 % power operation ? (2.00) QUESTION 5.08 (1.50) Concerning Reactor Core Flow I a. How is Core flow heated in the bypass reSion of the cor (0 50) b. Why is a certain amount of bybass flow require (0.50) c.. List the factors affecting flow resistance in a fuel bundle for both single and two phase flo (0.50) GUESTION 5.09 (2.00) During a cooldown of the reactor vessel fen = ^ " t c i i' t' - cen+eni cosm, racetor pressure decreased from 885 psis. to 595 psig. in one half hour.Has your reactor cooldown limit been exceeded 7 { show all work } (2.00) i I t
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! i I QUESTION 5.10 (2.25) Oyster Creek's reactor is operated within three (3) specified Thermal Lioits. List each of the three limits and explain the specific heat trans-for related problem that the limit protects agains (2.25) GUESTION 5.11 (2.00) The steady state MCPR limit given in Tech. Specs. is multiplied by a flow biasin3 correction factor Kf. Explain the bases for this correction factor including the events associated with i ( 2.00) GUESTION 5.12 (2.25) For the following changes in plant parameters will control rod worth increase, decrease, or not be affected? Briefly explain '4hy? An increase in moderator temperatur ( 0.75 ) An increase'in void conten ( 0.75 ) An increase in fuel. temperatur ( 0.75 )
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GUESTION 6.01 (2.25) Concerning the Main Turbine Control System i n. Following a Main Turbine ( lo- vacuum ) Trip f rom rated power , what pre-vents the turbine control valves and stop valves from popping open once the turbine is reset ? (0.75) 6. Briefly explain how a main Turbine Generator runback condition will occur. Include in your answer, the actuation signal, the response of the control and bypass valves, and when or if a turbine trip will occu (1.50) OUESTION 6.02 (2.75) Concerning the Standby Gas Treatment System ( SBGTS ) ; a. List four automatic si3nals that will place the SBGTS in servic NOTE: Setpoints require (1.00) b. With the SDGTS operatin3 at rated flow, which way will the inlet and outlet valves fail upon a loss of instrument air . Can the valves still be operated. (briefly explain). (1.00) c. If the SBGTS was automatically initiated by a reactor Low water level signal,what must be done to shut the system down once the Low -level signal clears ? (0.75) 00ESTION 6.03 (2.25) The EMRV *s have a three (3) position switch station on panel 1F/2 ( Automatic / Off / and Manual ) . List the functions provided by each Switch Positio (2.25) GUESTION 6.04 (2.00) Concerning the Isolation Condenser i e. Why would it be necessary to manually remove A and E Recirculation pumps from service after a manual initiation of the Isol. Con (0.75) b. All automatic isolation valves for the Isol. Cond. can be manually over-riden,(opened or closed), regardless of what is being called for by the logic cirev:try . ( TRUE or FALSE ) (0.50) Why is there a caution that directs you to maintain Reactor water level below JHt' Yaevay indicatio (0.75) lV T 74F l l t I
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . . . - - .__ GUESTION 6.05 (2.50) You are in the process of using the Alternate Shutdown Cooling method, ( cleanup system, condensate, and main condenser ) i - a. List two (2) methods you can use to decrease the temperature of the non-regenative heat exchan3er outlet as it approach's 130 de (1.00) b. During this shutdown cooling method why might you want to increase hot-well level 7 (0.50) c. Why might you want to place the mechanical vacuum pump in service, and what precaution should you be aware of when placing a negativa pressure on the reacto (1.00) GUESTION 6.06 (2.50) During the 4:00PM-12:00 MID shift you experience a total loss of all 125 VOC power; a.For each of the following components listed below, indicate whether there is a loss of position indication, control power, and/or reset capabilit NOTE; If more than one condition applies, indicate s (2.50) 1. Shutdo9n coolin3 system isolation valve . Main Steam Isolation Valve . Isolation Condenser, isolation valve . All 4160 V and 460 V breakers .
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QUFSTION 6.07 (3.00) Concerning the Reactor Level and Feedwater Control System i (Reactor Power 100 %, steady state) For the following, Loss of Signal failures to the Feedwater Control Sys., indicate in which direction reactor water level will responde(inc. or dec.)
the reason for the change, and the probable automatic action that will oc-cur with no operator actio (3.00) Feedwater Flow Steam Flow Reactor Water Level Steam Pressure Feedwater Temperature t i i
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l QUESTION 6.08 (1.'50) During a Reactor Isolation, with the Isolation Condensers in servicer you roceive a triple -low water level alarm. If'all control' room level indicat-ors appear to be normal; a. What caused this alarm condition ? (0 75) b. Why the apparent contsidiction between this alarm and indicated : level ? (0.75) QUESTION 6 09 (3.00) -
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a. For the following Reactor Level instruments, indicate where the water level is being sensed and if there is any compensation involve (1.50) 1. Wide Range GE-MA l 2. Narrow Range GE-MAC l 3. Lo-Level Yarway : I 4. Lo-Lo-Level Yarway
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6. Fuel Zone Rosemounts a GE-MAC's f Concernin3 the Fuel Zone Level Instrumentation i 1. What signal (s) is required to turn the system on ? (0.75) -
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2. Why won't the HIGH PRECISSION requirement of this system work if the Core Spray pumps are running ? (0.75) i r i f r I
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______________________________________________________ QUESTION 6.10 (2.25) Concerning the Standby Diesel Generators ; a. Explain the automatic sequential actions of the Diesel Generator for the fo110 win 3 two conditions i Limit answer to operation of D/G onl . Inadvertant LOCA signal, immediately clearing, with no loss of off-site powe (0.75) Loss of off-site power followed by a LOCA signal, with the Core Spray system loads already picked up, and then additional loading causes the Diesel Generator to stall and output breaker to tri (0.75) If an inadvertant, false, start signal is received by the D/G , how can it be manually shut down ? Would you want to do this immediately ?
( yes or no and WHY ) (0.75)
GUESTION 6.11 (1.00) Operating Procedure i 308 e ( Emergency Core Cooling Operation ), provides direction on controllin3 Core Spray injection to maintain reactor water leve What action must be taken to manually take control of the Core Spray Syste (0.50) 6. One,e manual control of the Core Spray system has been taken, what C/S components are used to control injection to the vesse (0 50) , I ,
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____________________ QUESTION 7.01 (2.00) According to procedure ABN-3200.13,(Response to Loss of All 125 VDC), a caution exists, that directs you to either maintain the plant in a steady state condition or enter a prompt reactor scram, irrespective of the Tech-nical Specification requirements. What two (2) specific criteria are used to make this determination ? (2.00) GUESTION 7.02 (2.25) l According to procedure 201.1 (Approach to Criticality) i i a. Why is there a caution in this procedure that instructs you not to ' switch to range 10 too early while in the startup mode. Briefly explain , the bases for this cautio (0.75) i b. If the RWM becomes inoperable before the first teelve (12) control rods i have been withdrawn , can the startup eoame*+e ? ( If not why ? If so C#"""# under what conditions ? ) (0.75) c. What two (2) warning indications are available to assure that the min-imum permissable positive period of 30 sec. is not exceede (0.4d-)ygf'
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GUESTION 7.03 (2.50) Concerning the 1983 Outage ' Control Room Modifications .'; c. Where are the new,' Loss of Power Alarms ', for the Vital Instrument Panel power supplies located ? (0.75) b. Why were modifications made to the Shutdown Cooling System and what equipment / components had to be replaced and/or installed as a result of this modificatio (1.00) What is the PRIMARY and SECONDARY fire protection system for the New Cable Spreading Roo (0.75) i OUESTION 7.04 (2.25) Concerning Procedure 340.2 ,( 24 VDC Distribution System ) i a. Can you commence a startup with the 24 VDC power panel 'A' inoperable ? l l
( 24 VDC power panel 'B' is operable ) (0.50)
6. A loss of all 24 VDC will cause what three (3) systems to be lost or cetivated i (1.00) c. Whrt two precautions / limitations exist concerning the operation and/or removal from service of the STATIC CATTERY CHARGERS . (0.75)
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t PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 10
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--- EA5i5t55iEAt E5nTR5t ____________________ QUESTION 7.05 (2.50) According to procedure EPIP-2 ( Emergency Direction ) , the Emergency Dir-cetor has certain responsibilities and/or authorities that nay NOT be del-osated to a subordinate during emergency conditions. List five (5) of these ections that may not be delegate (2.50) GUESTION 7.06 (1.75) Concerning Procedure EPIP-7, (Personnel Injury); a. What plant personnel are to be used to make up the FIRST AID TEAM.(0.50 b. During an emer3ency condition, (activation of the OSC), who would be re-sponsible for coordinating the off-site medical transport for injured personne (0.75 c. If unsuccessful attempts have been made to de-contaminate an injured person requiring off-site medical attention, the Emergency Director should be communicating with what off-site personnel ? Where ? (0.50 QUESTION 7.07 (3.00) Concerning ENG-3200.01 ( RPV Control / RC/L Level Control) i a. Under what conditions can the automatic controls of the Core Spray System be placed in its manual mode of operatio (1.00) b. List the Entry Conditions for this procedur (2.00) OUESTION 7.08 (3.00) ry / Concerningg p,rocedure ABN-3200.07 (Unexplained Reactivity ChanSe ) I List er . e( 5 ) plant parameters / indications that should be checked if an unexplained reactivity chan3e should occur at rated powe (1.50) Dependin3 on the ma3nitude of the reactivity change list four alarms that may be initiated. ( prior to a reactor scram ) (0.00) If this reactivity chan3e is a result of decreased temperature, due to a loss of a feedwater heater string , and a reactor scram has not oc-cured, what would be your next immediate actio (.70)
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I PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 11
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____________________ GUESTION 7.09 (2.75) je Concerning procedure ABN-3200.0y,(Control Room Evacuation) i e. Following a control room evacuation how is reactor level controlled ? , ( include location and components ) (0.75) ! b. Being forced to evacuate the control room, an attempt should made to bring the plant to a safe shutdown condition before leavins. List, eight (8) operator actions / verifications to be attempted prior to leavin3 (2.00) GUESTION 7.10 (2.00) List six (6) automatic actions you would want to verify upon receiving a reactor level Lo-Lo condition durins a small pipe break LOCA. Include those actions that may have been initiated from other signals as wel (2.00) QUESTION 7.11 (1.00) According to Administrative Procedure 106,( Conduct of operations ), what two criteria, concernin3 the on-commin3 Operators, must the off-soins oper-ators be satisfied with before relinquishing their responsibilitie (1.00) O
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t ADhINISTRATIVE PROCEDURES, CONDITIONS, AND LIhITATIONS PAGE 12
QUESTION 8.01 (2.00) Concerning the Limiting Condition For Operation ( Operability Requirements) as described in Technical Specifications. When a system, subsystem, train, cocponent or device is determined to be inoperable solely because its emer-gancy power source is inoperable, or solely because its normal power source is inoperable, it may be considered operable for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation only if two (2) conditions are vatisfied . LIST THESE TWO CONDITION (2.00) QUESTION 8.02 '(2.50) During you'r shift, 12:00mid - 8:00AM , at 100 % steady state power, one of your operators informs you with thatthethe plangEME o position indicat-ion li3ht for one EMRV is not indicatin The problem is determined not to be a burned out light bulb but a possible logic circuit problem . This problem cannot be worked on until 8:00 am, when maintenance personnel are availabl Using the attached sections of Tech. Specs , can the plant continue to operate under this condition ? Fully reference all sections of T.S related to this situation, giving a brief description and bases for any actions you would tak (2.50) GUESTION 8.03 (2.50) Concerning the use of Radiation Work Permits and the Control of Locked High Padiation Areas i a. List three conditions that must be adhered to prior to performing maintenance activities under the use of an Extended RW (1.50) 6. List two operational conditions that must be satisfied prior to entry into the TIP SHIELD too (1.00) 00ESTION 8.04 (2.75) Concerning Administrative / Federal RAD. dose limits and EMERG. Exposure's.; a. List the Federal Occupational Dose Limits and the administrative dose limits normally provided to site workers fori WHOLE DODY, SKIN, and EX-TRIMIT (1.00) b List the emergency exposure guidelines for whole body, extremeties and thyroid that apply for CORRECTIVE ACTION and LIFE SAVING ACTIO (1.00) Who , by title, is responsible for authorizing emergency exposures and can this responsibility be delegated ? (0.75)
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! ADhINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 13
__________________________________________________________ QUESTION 8.05 (2.00) c. What criteria is used to determine if a continuous FIRE MATCH is requir-ed ? (0.50) What communication and /or documentation would be required on your shift for a continuous FIRE WATCH ? (0.75) If a fire were to occur within the station, what caution is required concerning the automatic fire protection system on elevation 119 '?
(briefly explain why ) (0.75)
GUESTION 8.06 (2.50) n NOTE: USE THE ATTACHED SECTION OF THE TECHNTCAL SPECIFICATIONS TO *
* ANSWER THE FOLLOWING GUESTIO FULLY REFERENCE ALL SECTIONS YOU USE. *
During a shift turnover, with the plant operating at 75% power, you are informed that the Quarterly MSIV closure Surveillance test has exceeded the maximum allowable extension interval,and must be performed on your
*
shif Halfway through the test, ONE' Outboard MSIV FAILS to meet the specified closing time. In accordance with the Tech Specs: a. What situation exists due to the surveillance test being outside of the test frequency schedul (4 00) 6. What actions must be taken due to the fact that the MSIV has failed it's closing time test ? (l.KO) GUESTION 8.07 (2.00) In accordance with the Technical Specifications, the reactor uos scrammed due to Suppression Pool water temperature being greater than 110 degrees F. The reactor is now in HOT SHUTDOWN, Suppression Pool Coolin3 is ON, and Suppression Pool water temp-orature is 95 degrees Using the attached section of Technical Specfications, can you commence a startup ?? (Fully Explain) (2.00 QUESTION C.08 (2.50) During the beginning of your shift it is determined that one of the Con-toinment Spray pumps in Loop 'A' is inoperable. Half way through your shift the Diesel Generator associated with Containment Spray Loop * A ' fails its oporability Surv. Test.---Using the attached sections of Tech. Spec's, what action is required due to this situation ?
(2.50)
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S . ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 14 ________________ _________________________________________ GUESTION 0.09 (1.75) With the plant in cold shutdown, conditions in the reactor shiuld be ceintained to prevent thermal stratification of the reactor water level and inadvertent repressurization. List two (2) operating methods, briefly explainin3 how they are being used, to maintain this cold stable conditio (1.75) QUESTION 8.10 (3.00) Concerning Technical Specifications---> Safety Limits and Specifications i a. What is the bases for keepin3 a minimum of two (2) Recirculation Loop suction and discharge valves oPen at all times ? (1.00) b. During all modes of reactor operation with irradiated fuel in the ves-sel what is the minimum water level to be maintained, and how does this level compare to the lowest point at which water level can be presently monitore (1.00) c. What are the Tech. Spec. requirements for SRM operability during a core alteration involving the removal of one (1) control rod ? ' (1.00) GUESTION 8.11 (1.50) a. What is the NIL- DUCTILITY Transition Temperature ? (0.50) What Limitations are placed on plant operations, at Oyster Creek, to minimize the possibility of Brittle Fracture. (three required) (1.00) , e P m m-._ -.-_m s- , _
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. I TABLE II-3-1 PEPERTIES Or SAWRATED STEAM AND SAWRARD WATER (TEMPERATURE)
{l volume. ft'/ib enth.'py, Sture Entro,y, Stupb a F T ;' ;,' o.,,, t.., St..e w.i., c...
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s, r,, e, he se o 82 008S59 0 01602 3305 3305 -O 02 107 .5 0.0000 2.1873 2.1873 32 35 0.09991 0 01602 2948 2948 3.00 107 .0061 2 1706 2.1767 35 40 0.12163 0.01602 2446 2446 8 03 107 .1432 2.1594 40 45 0 14744 0 01602 203 .8 13.04 1068 1 10E .1164 2.1426 45 50 0 17796 C 01602 1704 8 1704 8 18 05 1065 3 10634 0.0361 2.0901 2.1262 50 50 0.2561 0 01603 120 .6 2806 1059 7 106 .0555 2 0391 2.0946 60 70 0 3629 0 01605 865 3 865 4 38 05 10540 10921 0 0745 1.9900 2.0645 70 80 0565 0016:7 633 3 633 3 45 04 104E 4 1095 4 0.0932 19426 2 0355 30 90 0 69il 0 01610 465 1 46I 1 SE 02 1042 7 110: 6 01115 18970 2 0~,56 to 1C0 09452 0.01613 350 4 35 4 62 00 103 .1295 18530 1.9225 100 110 12750 C.01617 2654 265 4 77.98 1031 4 1109 3 0.1472 1.8105 1.9577 110 120 1 6927 0.01620 203 25 203 26 87 97 1025 6 1113 6 0.1646 1.7693 1.9339 120 133 2.2230 C.01625 157.32 157 33 97.96 101 .8 0.1817 1.7295 1.9112 130 140 2.8832 0.01629 122 95 123 00 107.95 101 .6910 1.8E35 140 150 3.718 0.01634 97.05 97.07 117.95 1008 2 1126 1 0.2150 1.6536 1.8685 150 160 4.741 0.01640 77.27 77.29 127.96 100 .6174 1.8487 160 170 5.993 0 01645 62 04 62 06 137.97 996 2 1134 2 0 2473 1.5822 1.8295 170 180 7.511 C 016:1 50 21 50 22 148 00 990 2 1135 2 0 2631 1.5450 18111 180 130 9 34, C 01657 40 94 4 96 155 04 954 1 !!42 1 0 27E7 1 514? I7934 190 200 11 526 C01664 33 62 33 64 16E 09 977 9 1146 0 0 2540 1 45*4 1.7764 200 ;
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210 14 123 C 01671 27 80 27.E 2 17E.15 97 .4509 1.7600 210
, 212 14 696 C.01672 26 7E 2( 50 :5017 970 3 l* 5: 1 0 2121 1 4447 1.756E 212 220 17.156 0 01678 23 13 23 15 '8E23
. 965 2 1153 4 03241 14201 1.7442 220 230 20779 0 01685 19.364 19.351 195 33 95 ES 13902 1.7290 230 240 24 968 0 01693 16.304 16.321 20845 9521 1160 6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13.802 13 819 218.59 945 4 1164 0 03677 1.3323 1.7000 250 26~ , 25 427 C.01709 11.745 11.762 225 76 935 6 116 4 02519 1.3043 16562 260 4 270 41.E56 0 01715 10 G42 10.060 235 95 931 7 1170 6 0 3960 1.2765 1.6729 270 280 49200 0.01726 8.627 6 644 249.17 924 6 1173 6 0 4095 1.2501 1.6599 280 290 57.550 0 01736 7.443 7.460 259 4 91 .2238 1.6473 290 300 67.005 0.01745 6 44E 6.466 26 .7 0 4372 1.1979 1.6351 300 310 77.67 0.01755 5.609 5 626 280 0 90 .1726 1.6232 310 320 39.64 0.01766 4.896 4.914 290 4 894 8 118 .4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3 788 3113 878 8 119 .4902 1.0990 1.5892 340 360 153.01 0 01811 2.939 2.957 33 .1 1194 4 0.5161 1.0517 1.5678 360 340 195 73 0.01636 2.317 2335 35 .5 1195 0 0.5416 1.0057 1.5473 380 400 24726 0 01864 1.8444 1.8630 3751 825 9 1201 0 0 5667 0 9607 1.5274 400 420 30E 78 0 01894 1.4805 1.4997 39 .5915 0.9165 1.50SO 420 440 381.54 0.01926 1.1976 12169 4190 7854 1204 4 06161 08729 14890 440 460 46 .01 % 0.9746 09942 44 .2 1204 8 06405 0.8299 1.4704 460 480 566 2 0.0200 ** 0.7972 0.8172 46 .1 0 6648 0.7871 1.4518 480 500 680 9 0 0204 06545 0.6749 48 .4333 500 520 81 .0209 0.5356 0 5596 51 .0 1199 0 02133 0.7013 1.4146 520 540 962 8 00215 0 4437 0 4651 536 8 65 .3 0 7378 0.6577 1.3954 540 ,
560 1133 4 00221 03651 0.3871 562 4 6253 118 .7625 0.6132 13757 560 1 540 13262 0.0228 0.2994 03222 58 .9 1179 0 0.7876 0.5673 1.3550 540 Goo 1543 2 00236 02438 02675 61 .7 0 8134 0.5196 1.3330 600 i 620 178 .0247 0.1962 02208 646 9 50 .3092 620 , 640 2059 9 00260 0.1543 0.1802 67 .8686 0.4134 12821 640 J 660 2365 7 0.0277 0.1166 01443 71 .1 110 .8995 0.3502 1.2498 660 680 2706 6 0.0304 0.0808 0.1112 75 .5 0 9365 0.2720 12086 680 s 700 309 .0386 0.0752 822 4 17 .2 0.9901 0.1490 1.1390 700 . 70 .2 0.0508 0 0.0508 90 .0612 0, 1.0612 70 I
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ANSWERS -- OYSTER CREEK -85/04/09-LANGE, ~ ANSWER 5.01 (2.00) A. NO , Thermodynamic efficiency is a comparison of energy in versus energy out. CO.53 The increase in generator output resulted from decreasing the amount of steam diverted to the LP FW heaters.CO.53 This condition requires additional ener3y output from the reactor to raise FW temp to the same sattration temp as-before CO.53'Thus, thermodynamic efficiency of the plant has gone down. CO.53 More delta T across the heater would have caused more extraction steam to have been removed from the turbin n d h m ex t nt lp . (0.50) p, fr4 B, / M i 7g/o, d[d REFERENCE Oyster Creek LP. objective # 27, FW heater failur Basic Turbine / Plant thermo. effecienc ANSWER 5.02 (2.25) VOID COEFFICIENT (0.50), adds neSative reactivity (0.25)
. FUEL TEMP'. CdEFFICIENT (0.50), adds negative reactivity-(0.25) MODERATOR TEMPERATURE COEFFICIENT (0.50), adds positive' reactivity (0 25 REFERENCE Oyster Creek LP. 6 300.08, Rx. Coefficients and Control Rod Wort ANSWER 5.03 (1.50)
YES. (0.50) To maintain the removal of non-condensable gasses produced from the decomposition of water , activation products and noble' gasses produced in the fuel and leaking into the coolant via. cladding cracks.'(1.00) REFERENC Oyster Creek LP $ 68, pg. t5 , Design Bases for SJAE.
9
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' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 16
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4 . ______________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, D.
:
i ANSWER 5.04 (3.00) a. Adds negative reactivity CO.253 due to the increase in neutron leakage - Moderator temperature coefficient. CO.503 b. Adds. negative. reactivity CO.25] due to the increase in neutron capture in the fuel - Doppler coefficient.-[0.50] Adds positive reactivity CO.25] due - to the. decrease in neutron ] leakase - Moderator temperature coefficient. CO.50] Adds negative reactivity E0.25] due to the increase in neutron leakage - Void coefficient. [0.50 i REFERENCE
^
j Oyster Creek LP. Reactor Theory CH. 48 ( 8.12) Plant Parameter Effects on Control Rod Worth and Reactivity Coefficient . ! : ANSWER 5.05- (2.00) "
, At 100 % powe (0.75) At 4.% pcuer, you are at operating pressure but at low feedwater. flow rate. NPSH is low due to T inlet beins hish. As j power increases, pump inlet temperature is reduced-due to mixing in the i downcomer. T-inlet is lower so P-sat. at inlet is lower, therefor-NPSH is
- highe (1.25) , REFERENCE l Oyster Creek Fluid Flow CH # ANSWER 5.06 (2.25) . Rated void collapse.- i Fuel doppler effect decrease.
- 3.-Xenon deca . Temperature decrease to 125 de '
-l S. Shutdown aargin 4 % delta (5 correct ans.9 0.25 each) i b. Minimum- To prevent power chussin3 caused by uneven mixin3.- (0.50) Maximum- To compensate for cooldown following Xenon Pea (0.50) REFERENCE Oyster Creek LP.# 53,SBLCS.
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. . * . . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 17
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ANSWERS -- OYSTER CREEK -85/04/09-LANCE, D.
ANSWER 5.07 (2.00) Purpose & Req.-> Serve as a backup to procedural controls, limiting control rod worths, to assure that in the event of a control rod drop the reactiv-ity addition rate would not lead to significant fuel damag (1.00) One of the principal reasons that it is not required is due to the presence of coolant voids. The rod drop accident would be less severe.[As the voids increased durin3 the transient, the power excursion would be greatly limit-ed.] (1.00) REFERENCE Oyster Creek LP 4 49, RWN.
ANSWER 5.08 (1.50) This flow is heated by neutron and gamma heating. + c. m 2ct mh 0 (0.50) Core bypass flow is needed for cooling of components in the bypass res-ion and to prevent excessive voiding in the region which could affect the accuracy of the nuclear instrumentatio (0.50) Single phase flow- friction, acceleration, and local restrictions.(.25) Two phase flow- same factors as single phase determined by a multiplier looking at, coolant quality, bundle flow rate and sys. pres (0.25) REFERENCE Oyster Creek LP. Thermo. Ht. Transfer, Fluid Flow CH.S 9.
ANSWER 5.09 (2.00) First, convert psi to psia. by adding 14.7 psi. Then,refering to the steam tables; 900 psia. = 532 de psia. = 488 de des.F - 488 des.F =44 des.F / half hour, or 80 des./hr (1.50) NO. The cooldown limit of 100 des.F/hr has not been exceede (0.50) REFERENCE Oyster Creek LP . Thermodynamics and Heat Transfer (Steam Tables)
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. _ _ _ . . . _ _ _ _ _ . , _ . _ . _ _ . _ ~.. ., _ .. . . . ._ - . . . - . . . . . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 18
____ _ ______________________________________ ______________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, ANSWER 5.10 (2.25) 1. MCPR protects against the onset of transition boilin ( .75 ) 2. LHGR protects against exceeding 1% plastic strain on the clad 4 due to excessive heat generation in the fue ( .75 ) 3.MAPLHGR- ensures that peak fuel clad temperature will not exceed 2200 degrees F during a DBA-LOC ( .75 ) REFERENCE Oyster Creek Heat Transfer and Thermo. CH.19, Learning Objective 9-lb . ANSWER 5.11 (2.00) This flow adjustment factor increases the MCPR limit at core flows less than rated. Events such as LOSS OF FW. HEATING and TURBINE TRIP without bypass become less severe when initiated from power levels less than the design value. This is due to decreased steam flow. But events such as inadvertant start up of an idle recire. pump, recire. flow controller failure (increased flow ) and FW flow controller failure (man) can be-come more severe than transients which are limiting at design condit-ion ( 2.00 ) REFERENCE Heat Transfer and Thermal Limits, CH t 9 , Oyster Creek . ANSWER 5.12 (2.25) increase (0.25) As moderator temperature inc. so does thermal diffusion lenSth which increases the leakage of thermal neutrons from the fuel bundles and into the control rod regions.Thus rod worth in (0.5).
b. decrease (0.25) Moderator density decreases resulting in more fast neutrons and fewer thermal neutrons leaving the bundle. Since control rods are thermal absorbers, overall control rod worth decreases (.5).
c. no effect (0.25) Since fuel temp. affects primiraly fast neutrons, which are resonantly captured, and control rods are thermal neutron absorbers, fuel temp. and rod worth are essentially independent of each other. (.5 ). REFERENCE < Oyster Creek RX.Tneory CH #0, pg. 964 a6 I ! l
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. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 19
______________________________________________________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, ANSWER 6.01 (2.25) c. The Load Limit handwheel trip prevents this actio ( Trip oil pressure must first be restored and the handwheel moved to the closed position, relatchin3 the handwheel spindle.) to.75) b. 1. Runback- Caused by loss of stator coolin (0.50) 2. Control Valves move in the closed directio (0.25) i Bypass Valves move in the open directio (0.25) ! 3. Turbine Trip- If, after three (3) minutes, the stator current is greater than ( 4800 amps. ) (0.S0) REFERENCE l Oyster Creek LP 4 73, Turbine Control System, Rev. #1 Ps.-2 ANSWER 6.02 (2.75)
- Reactor building vent High Radiation. > 13 mr/h (0.25)
Reactor buildin3 Operating floor (elev.119) > 70 mr/hr after a 2 minute time dela High drywell pressure > 2.49 psi yo sj, (0.25)
(0.25)
Lo-Lo reactor water level ( 86' above the TAF ) (0.25) Valves fail as-is. Accumulators on both inlet and outlet valves allow for five (5) valve cycles, after a loss of inst. ai (0.75) The drywell isolation reset button must first be reset. (3-S-2) (0.50) j REFERENCE l Oyster Creek Oper. Procd. # 330 rev.14 ( SBGTS ) ANSWER 6.03 (2.25) l 1. Auto. (0.25), allows auto operation of EMRV's for ADS and OVER-PRESSURE l relief function. (0.50) l 2. Off. (0.25) , allows auto action of ADS but defeats pressure relief l function. (0.50) 3. Manual. (0.25) energizes the solenoid valves to open the EMRV's. (0.50) l REFERENCE Oyster Creek LP i 5 . AOS.
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i - . PLANT SYSTEMS DESIGN, . CONTROL, AND INSTRUMENTATION PAGE 20 ______________________________________________________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, , i ANSWER- 6.04 (2.00) To prevent an inadvertant system isolation caused by induced high flow { ' j-
'
through the condenser (0.75) TRUE (0.50)
; c. The steam lines to the isol.cond. come off directly at an elevation .correspondin3 to approx. 95' Yarway indication. Operatins above this
, water level will'cause a potential for damage due to water hammer.(0.75) REFERENCE ', Oyster Creek LP 4 21e Isol. Con ANSWER 6.05 (2.50) l
- Increase RBCCW flow by opening the NRHX outlet valve, (V-5-122). (0 50)
! Reduce letdown flow by using FCV-ND-22 . (0.50) ] b. To minimize the Oxygen addition to the condensate syste (0 50) c. This is an attempt to reduce the Oxygen level content in the reactor
- coolant water.-(0.50). With a negative pressure in the reactor, boiling
, will occur at a lower temperature than 212 de3 F.(0.50) (1 00) i i REFERENCE Oyster Creek , Proed. # 203.3, Alternate S/D Coolin j ANSWER 6.06 (2.50) i . Loss of osition indt' catio . Loss of osjAion kndd and reset capabilit ! 3. Loss of osition,)nd} and reset capabilit ; 4. Loss of control power._Fy4,wec-ee -
- 5. Loss of control power. g* * ve nd- , . (0.50 for each correct ans.)
- REFERENCE "##'*' i Oyster Creek, 2000-ABN-3200.13. Response to loss of All 125 VDC. p . i
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. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUhENTATION PAGE 21
______________________________________________________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, ANSWER 6.07 (3.00) a. FW-Flow > Level increases (0.20), due to large flow error signal causing FRV to open fully (0.20). Probable Turbine Trip on high level.(0.20) STM-Flow > Level decrease (0.20), due to large flow error signal causing the FRV to close (0.20). Possible Low level Scram. (0.20) Rx. Water Level > Level increases (0.20), due to large level error signal causing FRV to open fully (0.20). Probable Turb. Trip on hi-level. (0.20) d. STM-Pressure > Indicated level will decrease (0.20), due to the decrease in indicated steam flow there will be a resultin3 overfeeding and under-feeding signal being sent to the FRV (level vs steam flow signal)(0.20).
Depending on the power level,(steam flow), and level being maintained at the time of the incident, actual level may increase or decrease. ( Possible Turbine Trip or Reactor Scram (level inc. or dec),but in either case less significant than a loss of the basic three element (0.20) Feedwater Temp.> Level increase (0.20), due to the temp. signal failing high causing an indicated decrease in FW-flow and opening the FRV.(0.20) Probable Turb. Trip on high level. (0.20) REFERENCE Oyster Creek LP. 844, Attachment #2 Loss of Input Signal to FW-Contro ANSWER 6.08 (1.50) - No return flow path VIA Recireviation loop piping exist (0.75) This is because the Tripple low sensor monitors level within the Shroud while all other sensors monitor annulus leve (0.75)
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REFERENCE Oyster Creek Proed. # 307, p . Isolation Condenser System.
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.. _ _ _ _ . _ . - . . . - _ . _ . ~ _ _ _ . . . -- . _ .. ._.._ _ ..~.L . . * . . . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PACE 22
______________________________________________________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, . ANSWER 6.09 (3.00)
. For instruments # 1,2,3,4, water level is sensed in the annulu For * 4 5,6, level is sensed in the core shrou Compensation; 1. none density compensated (ig sing MSL pressure.)
3.& compensated (usins heated ref erence les.) none (calibrated for o p e r a t i n g c o n d i t i o n s .) density compensated (usin3 reactor p r e s su r e .)
(0.25 for each full correct answer) ,
b .1., System turns on only when all five recire. pumps tri ( ) Mith tk- ccirc. ; ; ,r -
..o us e rel;c 1reci .jcit
^
a i j.c l ic Onn- , arn+=d +>r is fc-rc;d ciim lativo ...3 .wav1 Lina pr= w r: dc y ure;; tum .mm_ em-,,,te (0.50) 3. The variable les of the transmitter is sensed at the Core Spray spar-ser, therefor a false high level signal is generated due to the C/Sy3 pump discharge hea ( 0.M ) REFERENCE Oyster Creek LP. # 64, ps.7-16, and pg.3 . ! ANSWER 6.10 (2.25) i . The D/G immediately starts, goes to idle, the automatically shuts
! down after a time delay of ( 11.5 Min. ) (0.75)
2. The persistant undervoltase condition will cause an automatic re-i start of the D/G and begin loading in a timed sequenc (0.75)
- The Control Room start /stop switch or the fuel oil cut-off switch in the
D/G engine compartment.(0.25) j ' NO , The 11.5 min. idle feature allows engine cooldown before shutting down. (0.50) i REFERENCE Oyster Creek Oper. Proed. # 341, ps.1-26.
) ANSWER 6.11 (1.00) i i a. Must have cleared or overridden all initiating and auto start signals.
l (0.50) b. Controlled by openin3 and closins the C/S parallel valve (0.50)
- REFERENCE l Oyster Creek, Oper. Procd. # 300rsec.5.3.5, and LP. I 10, C/S logic.
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*- . .- , . ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23 ~~~~ ~~~~~~~~~~~~~~~~~~~~~~~~
RA65 L665CAE~C6UTR6L ____________________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, ANSWER -7.01 (2.00) 1. If a reactor limiting safety system setpoint has/has notebeen exceede (1.00) 2. If reactor vessel pressure boundary integrity or fuel inte3rity is threatene ,
(1.00)
REFERENCE Oyster Creek-Proed. ABN-3200.13 , Rev. O, pg.4&1 ANSWER 7.02 (2.25) Switching the ranse switch too early may result in an automatic closure of the HSIV's and MSL drain valve (Jechnical Sp cifications require a minimum recirculation flow'of'?" '" '^ ' E
.'nhile i t. the startup mode ~before switchin3 any IRM range switch to position 10.) (0.75) Yes. . (0.25) Only with the permission of the Manager of plant operations, or his designee in accordance with the approved procedure (5218). (0.50)
c. 1. SRM PERIOD SHORT alarm on panel 3 ( 0 J5$') 2. Amber peri.od alar on panel 4 (0 316 3. B y Pau L .p'. 8. t . REFERENCE (* * "8 Oyster Creek Proed. # 201.1, ps. 1-1 ANSWER 7.03 (2.50) Annunciation for these new alarms are on panel 9-x-7 in the Main Con-trol roo (0.75) The system could not be properly isolated to perform maintenance of the SDC heat exchan3er and the inside containment M */>isqlation valves and pump inlet valves needed repai VG M J (0.25) Replacementofinsidecontainmentisolation} valve (0.25) 2. Replacement of pump inlet valves for each loo .
(0.25)
3. Installation of flow ind. (local) for each of the (3) SDC loops.(.25) c. Primary- Automatic wet pipe sprinkler syste , (0.50) Secondary- Manual wet pipe hose station. ( pd( R e;*: -4 hI- (0 25) REFERENCE Oyster Creek LP. 9 57, Handout 1210 03., Learnins Objective .
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, . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY fND PAGE 24 ~ ~~~~R dD56L6656hl 66NTR6L'~~~~~~~~~~~~~~~~~~~~~~~
____________________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, ANSWER 7.04 (2.25) NO, (Both A & B 24 VDC panels must be energized before the RX. is made Critical ). 3,m s (0.50) b. 1. Loss of neutron monitoring syste y ,7 % 2. Loss of the area radiation monitoring syste Mf9*/acapZu( (0.33)
(0.33) Initiation of the Reactor building ventillation Isolation. 7}] . (0.33) . Do not operate the Static Char 3er s without its battery beins connect-e (0.375) Do not remove from service unless its respective battery is supply-ins power to the affected DC bu (0.375)
REFERENCE Oyster Creek Proed. 340.2, pg.t 1- ANSWER 7.05 (2.50) 1. Classification of the even . Approving and directing official notification to off-site agencie * ' ' information releases to the medi . Approving and conveying protective action recommendation to the office of Emergency Hgn . Directing on-site evacuation at the Alert or lower level to non-assigned personne . Authort:ing emergency workers to exceed 10 CFR 20 Rad. exposure limit (any 5 et 0.50 each) REFERENCE Oyster Creek EPIP-2, pst 3 ANSWER 7.06 (1.75) a. A member of the plant shift, qualified in first aid, and a qualified Radiation Control Tec (0.50) The Operation Support Coordinatore(The Emer3 Dir. would have this re-sponsibility during non-0SC activation). (0.75) The Nursing Service Supervisor at Community Hemorial Hospita (0.50) REFERENCE Oyster Creek EPIP-7, Personnel Injur r
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. . . . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25 - ------------------------ ~~~~R5656LU55CdL C5RTR6L -
____________________ ANSWERS -- OYSTER CREEK -85/04/09-LANCE, ANSWER 7.07 (3.00) . Misoperation in automatic is confirmed by at least two independent process parameter indication (0.50) [. Core cooling is assured AND this procedurer( g g.y p),e irects you.(0.50 -, ., . RPV Level below + 138 TA . Drywell Pressure above 2.00 psi . A condition that requires a reactor scram and power above 2 %. RPV pressure above 1050 psi . MSIV closur . A condition requiring a scram , to conserve inventory or reduce re-lease of radioactivity to the enviroment, determined by the operator.(0.50 REFERENCE Oyster Creek EMG-3200.01 RPV Control.
ANSWER 7.08 (3.00) . Main Gen. Outpu . Steam flow or tem . LPRM reading . Feedwater flow or tem . Reactor powe (5 correct ans at .3 ea.) . LPRM - HI , , 2. APRM - HI, (3) IRM - HI , (4) SRM -HI,inopeshort (four correct at 0.20 each) Reduce power to 80 % of the power level prior to the change using Recirc flow. (0.70) REFERENCE ysTv.9, GRes.K,32c047 Mj 1- to (Lg c. 3 2co.16
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____________________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, ANSWER 7.09 (2.75) a. Usins the Control Rod Drive System , elev.23' in the reactor bids. by confirmins both CRD pumps are running and throttlein3 the by-pass valv (0.75) . Scram the Reactor 2. Verify all rods i . Trip all Recire. pump . Trip the Main Turbin . Close the MSIV's 6. Trip all FW pumps Trip all condensate pump , . Initiate both isolation condenser ( ei3ht correct answers at 0.25 each ) REFERENCE Oyster Creek , ABN 3200.30 (Control Room Evacuation) ANSWER 7.10 (2.00) 1. Core Spray system initiatio . 2. Diesel senerator start and idl . Recire. pumps tri . Primary Containment Isolatio . Secondary Containment Isol. and start of SBGT . Reactor Isolatio . Isolation Condenser initiatio (any 6 at 0 33 each) REFERENCE Oyster Creek 0 a A bank , and LP. for Reactor Level Instrumentation.
ANSWER 7 11 (1.00) 1. Is fully qualified and/or licensed, to assume the shift position. (0.50) 2. Is fully aware of ev.istin3 Plant / equipment condition (0.50) 3. B e 444[dM. ,g) . REFERENCE
.
A.P-106, Sec. 4.3. bs: o r w.. a r u s o u!**l kt ? J*b
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- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 27
___________________________________________ ______________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, D.
ANSWER 8.01 (2.00) 1. Its Corresponding NORMAL or EMERGENCY power source is operable. (1.00)
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2. All of its redundent system (s), subsystem (s), train (s), component (s) and device (s) are operabl (1.00) REFERENCE Oyster Creek, TS.Sec. 3.0, Operability Requirements. Specification ANSWER 8.02 (2.50) With no indication of valve position and possible logic or maintance prob-l 100 existing and unable to be corrected, the valve has to be considered in-operable according to the definition of operability. (0.50) The LCO for pressure relief systems-Solenoid Actuated Relief Valves , as stated in section 3.4-4 , requires all (5) valves to be operable when you are at operating temperature and pressure.(0.50) The primary bases is for depressurization to allow for full flow core spray operation in the i event of a small line break when the feedwater system is not active. (0.50)
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Section 3.4.B-2 allows for continued operation for up to (3) days provided the motor operated isolation and condensate makeup valves in both isolation condensers are demonstrated to be operable. (0.50) If specification 3.4.B 1 & 2 are not met the reactor press shall be re-duced to less than 110 psis. in 24 hours. (0.50) ' REFERENCE Oyster Creek T.S. sec. 3.4-4 , LCO for ADS.
ANSWER 8.03 (2.50) a. No contamination or potentially contaminated system openin (0.50) Must be a job of short duratio (0.50) The work is not expected to cause any significant change in the static radiological condition in the are (0.50) 6. 1. TIP detectors are either in their chamber shields or inserted to i 3reater than 400 '. (0.50) 2. Caution tags are placed on the drive control drawers (panel 4R)(0.50) REFERENCE Oyster Creek Proed. 915 12 (RWP) and ADM-4110.06 (Control of Locked Hi-RAD Areas.)
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ANSWERS -- OYSTER CREEK -85/04/09-LANGE, . ANSWER 8.04- (2.75) a. Federal:. Administrative! Whole Body- 1.25 R/qt (3R/qt max.NRC-4) 1R/qt-4R/yr (0.33) Skin - 7.5 R/qt 3R/qt (0.33) Extremenity- 18.75 R/qt 15R/qt (0.33) b. Whole Body- 25 R 75 R (0.33) Exrtemenity- 100 R 300 R (0 33) Thyroid- 125 R No Limit (0.33)
(protective ) (life naving) .
c. Emergency Director (0.50). NO, (0.25) (0.75) REFERENCE Oyster Creek ADH-4000.01, ps # 3-5 and EPIP-7 ost ANSWE .05 (2 00) a.~ A continuous fire watch is required when the fire supression system in that area has been isolate (0.50) b.(One hour) communication to the control room operator. (0.50), who-doc-unents conformation in the control room 103 (0.25) This system shall be valved out of service and only used as a manual system when the new fuel storage vault contains new fuel and is un-covered. Evacuation of the area is necessary before manual action.(0.75) REFERENCE Oyster Creek Proed. # 120.4 Fires pg #2 and i 120.2 Cont. Fire Watc ANSWER 8.06 (2.50) I S e. The LCO for the HSIVs are sta ted in %+pec, b+e ++,+A.T , T*lete
. Two mainJ.I.2 steam gj .
line isolation valves per main steam line shall be operable times 3r ater than or equal to 3 secs and less than or equal (with to 10closin,3 sec.)
Operatina outside of this specification is a violation of T.S. and therefore is a ,, rtable occuranc (1.00) b. Action req.of ktates that with the one HSIV INOP'due to exceedin3 the allowable closing time, the affeggg*d steam line shall be isolated . If the problem isnot. corrected /,'~ilttsateanorderlyshutdown. Table 3 1.1 also allows one hour to complete the test without having to trip the inoperable trip syste (1.85) REFERENCE . Oycter Creek Tech. Spec. Table 3.1.1 and LP 0 23, pg. 3 .
e ( -a -.m - - ' w
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l l l' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 29 __________________________________________________________ ANSWERS -- OYSTER CREEK -85/04/09-LANGE, D.
l l l ANSWER 8.07 (2 00) NO.(.75) Operation shall not be resumed until the pool temperature is re-l duced to below the power operation limit specified in section 3.5.A. This section only allows power operation to be resumed if pool temperature is BELOW the 95 de . (1 25) REFERENCE Technical Specification 3.5.A.1, specfication ANSWER 8.08 (2.50) With the Diesel g,ailins its surv. oper. test the 'A' Containment Spray Loop becomes inoperab1' (0.75). T.S. 3.4.c.3 requires that plant operation can only continue for 7 ays and that the C/S loop 'B' be demonstrated operable daily.(1.00) This LCO is different for only one pump bein3 inoperable as in the beginning of the shift where plant operation could continue for a period of 15 days providin3 the other pump in 'A' loop be demonstrated op-erable dail .S 3.4. (0.75) REFERENCE Oyster Creek T.S. Sec. 3.4.c.3,4,5.
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i ANSWER 8.09 (1.75) By increasing Reactor Water Level---> This helps maintain and promote , natural circulatio (0.875) l 2. Maintaining forced circulation ---> By using one recirculation pump or . un-throttled shutdown coolin3 flo (0.075) ' REFERENCE Oyster Creek Thermo.& Heat Transfer, CH.4 9, pg.125, Chapter Summar ,
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, ANSWERS -- OYSTER CREEK -85/04/09-LANGE, .
- ,
. ANSWER '8.10 (3.00) [
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i a. This assures that an adaquate flow path and communication exist from the j annular space and the core shroud to core re3 i on. (0.50) This is needed
to be assured that reactor water level instrument readings are indicat- ! ive of water level in the coret (0.50) ,
; b. 1. 4 ft, 8' above the TAF.(0.&g t), esion.Water level can presently be ponitored
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3 y y g gyv.p..., . AF r--4 0. 50 ) The SRM nearest the core alteration must be operabl a, F~ c y -[, . ) - {/. (0.50
' ! Two SRM's must be operablei one in the core quadrant and one in the ,
' adjacent quadran (0 50) REFERENCE Oyster Creek, Tech. Spec.- bases and spec. see.2.1rSafety Limits and Fuel l Clad Integrit i , ANSWER 8.11 (1.50) The temperature at which metal suffers brittle fracture without exhibit-
; in3 ductile yeil (0.50)
- b. 1. The temperature at which the head closure studs can be tensioned.(.33 j 2. The minimum temperature for a Siven reactor pressure under various '
; operating condition (.33) ; 3. The minimum temperature at which the reactor can be made critical (.33 i
J REFERENCE j ' Oyster Creek Thermo. and Heat Transfer, Learning Coals and Chapter Summary for LP. # 1 d&~g + u n . rvA. 4} - l- o . j ,
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l TEST CROSS REFERENCE PAGE 1 l QUESTION VALUE REFERENCE
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05.01 2.00 DJL0000185
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, 05.02 2.25 DJL0000186 l 05.03 1 50 DJL0000187 l 05.04 3.00 DJL0000188 l 05.05 2.00 DJL0000196 05.06 2.25 DJL0000197 i 05.07 2.00 DJL0000198 1 05.08 1.50 DJL0000211
- 05.09 2.00 DJL0000212 l 05 10 2 25 DJL0000213
' 05.11 2.00 DJL0000214 05.12 2.25 DJL0000215 ______ 25.00 06.01 2.25 DJL0000199 06.02 2.75 DJL0000200 06.03 2.25 DJL0000201 06.04 2.00 DJL0000202
'06.05 2.50 DJL0000203 *
06.06 2.50 DJL0000204 06.07 3.00 DJL0000208 06.08 1.50 DJL0000209 06.09 3.00 DJL0000210 06.10 2.25 DJL0000218 06.11 1.00 DJL0000235 ______ 25.00 r 07.01 2.00 DJL0000205 07.02 2.25 DJL0000206 07.03 2.50 DJL0000207 07.04 2 25 DJL0000219 07.05 2.50 DJL0000230 l 07.06 1.75 DJL0000231 i 07.07 3.00 DJL0000232 07.08 3.00 DJL0000233 07.09 2.75 DJL0000234 07.10 2.00 DJL0000236 07.11 1.00 DJL0000237 ______ l l 25.00 l 08.01 2.00 DJL0000216 i 08.02 2.50 DJL0000217 08 03 2.50 DJL0000221 l 09.04 2.75 DJL0000222 08.00 2.00 DJL0000223 08.06 2.50 DJL0000224 08.07 2.00 DJL0000225
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TEST CROSS REFERENCE PAGE 2 QUESTION VALUE REFERENCE ___ ___ ______' __________ 08.08 2.50 -DJL0000226 08.09 1.75 DJL0000227 08.10 3.00 DJL0000228 08 11 1.50 DJL0000229 ______ 25.00 ______ ______ 100.00
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U.S. NUCLEAR REGULATORY COMtISSION REACTOR OPERATOR LICENSE EXAMINATION Facility: oyster creek I Reactor Type: BWR
- Data Administered
- April , 1985
- Examiner
- Brf an K. Ha ick j
Candidate:(Print) MAsrine ) INSTRUCTIONS TO CANDIDATE: ! Use separate paper for the answers. Write answers on one side only.
j Staple question sheet on top of the answer sheets. Points for each t j question are indicated in parentheses after the question. The passing
' l1rade requires at least 705 in each category and a final grade of at '
east 805. Examination papers will be picked up six (6) hours after the examination start I i X of Category X of Candida'te's Category Value Total Scori Value Category ! 4, 25 25
! ! Principles of Nuclear Power
' D. + ; -
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Plant operation. Therno- ! ! $namics Heat Transfer ! ! and Fluid Flow t l 25 25 l Plant Design Including
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Safety and Eme'rgency
,
j , Systems 7
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25 25 1 Instruments and Controls
- 25 25
! Procedures - Normal i Abnormal. Emergency,, and ;
Radiological Control ; 100 I TOTALS ! l
- Final Grade % ,
All work done on thfs examination is my ow I have neither given nor t j received ai !
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t i i candidate's signature ! ! ! : ! [ !
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.. . ,A ! April, 19G5 Oyster Creek
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$ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMIC ! I HEAT TRANSFER,-AND FLUID FLOW (25) ! . .P. 201.1, Approach to Critical, cautions "under !
- conditions of high Xenon' concentration, the source l
- orange count rate may be quite small as criticality !
is approache It is, therefore, necessary to l i monitor-source range instrument responsiveness to i
- control rods with extreme care. . . " .
!
i Explain the reason f or this caution, and how the '
- conditions of high Xenon concentration can af f ect t
! individual control rod notch worth (3.0) i
.
i Core orificing is used to assure unif orm flow i through all fuel elements, somewhat independent of i
;
the power variations in individual fuel bundle i
4 Explain how the resistance to flow changes as i i the power increases in a fuel bundl (1.0) ' How does core orificing assure relatively i
, uniform flow independent of plant operating >
l conditions and location in the core? (2.0)
e j 7h The HCU accumulators are charged with nitrogen to (
( jy> assure rapid insertion of the control rod I a .
,
- When the accumulators are procharged, !
1 instructions in O.P. 302.1, Control Rod Drive , j Hydraulic System, state that the accumulator l t - charging pressure be set only when the l i . temperature of the nitrogen has reached * i j equilibriu What is.the purpose of this
'
f instruction, and what would result if the ; 4 instruction was not followed? (2.0)
'
! What is the approximate accumulator nitrogen l pressure during normal reactor operation? Why l
- is it so dif f erent from the procharge pressure 7(1.0)
, l During a refueling outage, the core is loaded with ; i suf ficient f uel to achieve an effective ; multiplication factor (Keff) of about 1.25 i
I
- Since Keff in a critical reactor in 1.000, give *
l ! three reasons why the core is loaded with so j j much excess Kef (1.5) ! ! ! 1 What are the two means of compensating for '
this excess reactivity in the reactor? (1.0) j j Category Continued'on Next Page
i RO Examination Page 1 of 12 I . t ; J >
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, For the centrifugal pump characteristic curve shown in Figur.e 1.1, show how the pump operating ch vg will change, and explain why .
the change occurs, if a valve in the system is '
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throttled one-half closed, such as.might be ,.. -. done in the Feedwater Control System. Be sure ., to label all points and line (1.5)" The Recirc System uses changes in pump speed to control flow rate. Usina the same'fiaure, explain which flow control systemf,i s mor's efficient fece the standpoint of the amount of power require (1.5) ,
Head
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"A Figure Pump Characteristic Curve f
l , Consider a single control rod positioned near the radial center of the core. During a reactor startup near the end of life, this control rod will be the first control rod to b ithdraw This dk"*
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full withdrawal of the control 11 have a certain ! reactivity worth. After the reactor has been made critical and 100 percent power has been achieved, if this control rod is fully inserted while reactor power is maintained at 100 percent, will the reactivity worth of its full insertion be greater than, the same as, or less than it was at the time of reactor startup? Fully explain your answe (3.0) Describe the initial reactivi ty response (positive or negative) and the resultant effect on core power (increase, decrease, no change) for each of the f ollowing changes in plant parameter Include which reactivity coefficient is responsible for the chang Closure of one MSIV at full powe (1.5) Loss of feedwater heatin (1.5) Category Continued on Next Page RD Examination Page 2 of 12
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fE April, 1985 Oyster Creek
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3 The reactor is critical at "50" on Range d of the IRMs. A control rod is withdrawn three notches, resulting in a power increase with a stable reactor period equal to the minimum permissible sustained positive period permitted in OP-201.1, Approach to Critical.g Heating power is estimated to be at "30" on Range /. Show all your assumptions and work for the following calculation What is the doubling time? (1.25) How long will it take to reach haating power? (1.25) , The MCPR operating limit is established to ensure ' that the fuel cladding integrity safety limit is not exceeded for any moderate frequency transient.
, For operation of the Oyster Creek reactor, gave two examples of postulated pressure transients, and two examples of postulated temperature transients, and indicate which transient will have the most severe
: effect on NCP (2.0)
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RO Examination Page 3 of 12 1 .
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N April, 1985 Oyster Creek i
; PLANT DESIGN, INCLUDING SAFETY AND EMERGENCY SYSTEMS (25) Each EMRV discharge line is equipped with two vacuum breaker When and under what plant conditions will the vacuum breakers open? (1.5) What adverse condition could occur if the vacuum breakers did not function properly, and why must this condition be avoided? (1.5)
t The Shutdown Cooling System is used to complete plant cooldown after the main condensers are no longer acting as a heat sin When Shutdown Cooling is started, how must operation of the RBCCW System be changed? (1.0) What requirements are placed on the lineup and/or operation of the "E" Recirc loop? Why? (1.5)
- The combined Standby Gas Treatment System contains seven automatic isolation and purge valves.
j [# If the SBGT System is operating with its normal j g 3j
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valve lineup, how will these valves (damoers) position on a loss of (1) solenoid power, and i
(2) on a loss of instrument air? (1.5) What system operational capabilities are provided in case of a loss of instrument air? (0.5) hvo Scevice Water normally provides cool,kdg water to the RBCCW heat exchanger What 4hmee alternate functions may be provided by the Service Water System? Include the operating conditions under which these functions are provide (1.5) The Standby Liquid Control System requires heaters in the storage tan Why must an elevated tank temperature be maintained? (0.5) What will cause an automatic trip of the heaters? (0.5)
Category Continued on Next Page RO Examination Page 4 of 12 e
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-April, 1985 Oyster Croek i The. f unction of the vacuum priming system is to remove air from the Circulating Water System. The system includes two vacuum pumps, one of which (the
" pref erred" pump) normally runs continuousl > Under what conditions will the reserve pump start? (0.5) Explain how and why condenser vacuum may be affected if the vacuum priming system fails to operate properly, such as if a float valve should fail in the closed positio (1.5) Each Core Spray System contains main and booster pumps to provide water to the reactor under accident conditions.
, What will cause the start of the priority main ; i Core Spray pump? (0.5) -
- Under what conditions will the backup main Core ,
] Spray pump start? (0.5) When will a permissive to ADS be provided? (0.5)
'3g When will the parallel isolation valves open? (0.5)
, M .P. 308, Emergency Core Cooling Operation, ' j directs that Core Spray injection be controlled
- to maintain water leve What must be done to r
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! take manual control, and what Core Spray } component will be used to control injection to { the vessel? (1.0) , ] The Reactor Cleanup System has two types of system ,
- isolations. For each isolation type, (1) list the i
signals that initiate the isolation, (2) the ' actions that occur within the Cleanup System, and , j (3) how the isolations must be rese (3.0) Category Continued on Next Pag i 1 .
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! RO Examination Page 5 of 12 l
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April, 1985 Oyster Creek The Isolation Condensers are available to depressurize the reactor and to remove decav heat if the main condenser is unavailabl What is the normal valve lineup for standby operation of the Isolation Condensers? Why is this lineup required? (1.5) When the Isolation Condensers auto initiate, all five Recirculation System pumps trip. If the Isolation Condensers are manually I initiated, only one or two Recirculation System pumps may need to be trippe Briefly describe the required Recirculation System valve lineups
, for each of these operating conditions, and explain the reason for these lineup (1.5)
{
2.10 Three independent power supplies are provided for
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the fdur 4160 vac buse ' s_ With the plant operating at 100 percent power, what is the normal power source for each of the four 4160 vac buses? (1.0) With a loss of the normal power source, how vi S, may each bus be powered? Be sure to consider
,}, f all alternative (1.5)
2.11 The Containment Spray System provides for Containment cooldown following a loss of coolant acciden . What initiation signals must be present for Containment Spray to start automatically? (0.5) What will occur if the Emergency Service Water pump for either System I or II fails to start, and what must you do to assure system operation? (1.0) End of Category . RO Examination Page 6 of 12
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Oyster Creek INSTRUMENTS AND CONTROLS (25) High temperature in the drywell can affect reactor level indacatio . Explain why ine -indicatwd Level will not read true and in what direction it will read under conditions of high drywell temperature (1.0) Which level detectors (Varway, Barton, or GEMAC3) will be rest a,ffected by high drywell
"temperaturew? W6y? '
(0.5)
ci Hcw much level-errorfcan be introduced by high ocywell temperatures, and,what adverse effects can this havef on rectuired system actions? (1.0) Instrumsnt Air is required to.,osition or maintain ts powT tibn of a large r. umber of valves in the
' plant. (Fce a, loss of inr,trument air header pr esourt/, h ow.a will each of the following be affected? (2.5) .
Hots l'h level, control valve (V-2-16)-$%va.14 Ib. D;y4=11 Equipepnt Erain Tank valves I, - c. > CRD fIow ' control valves g(% %..?
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s d. . Isolation Cancenser vent valves
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e. j Cleanup Skutest , letdown @alve , 3 g, , ' Each AFAH channel cont 5 ins six~ trip units that provide rod blocks and reactor scram signal For each. trip unit, with the reactor operating in the R'M , mode , indi,cate.whether a rod block or a scram
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sFgnbl, is gaayratett, and whether that signal may be rewat. 'fSetpoints ore not require .
(3.0) - The c7 actor, initially operating at 100. percent <
powe wPhas just scrammed due to a load re;ec The rer.ctor pEctection system operaten as expected
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and 'tip/ reactor ih;,ttdown occurs normally'. Explain in detail what tscu must do to ' reset thig gettigglgt
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scram, incluc"tng any time delays that occur, as
' .,. well as the r,epecns for thase time;delaysc and any / interlocks thgt.msy be involved.*- (2.5)
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April,.1985 Oyster Creek l An Anticipated Transient Without Scram (ATWS) pump trip was added to the Reactor Recirculation System to reduce reactor power if the reactor does not scram when it shoul What parameters or plant conditions initiate the trip logic, and what sensors are used? (1.5)
. Briefly explain how a spurious trip is precluded on a loss of 125 VDC power to the Channel A trip circuitr (1.5) List two alternate instruments or instrumentation channels that could be used to determine reactor pressure in the event that normal reactor pressure instrumentation (wide and narrow range reactor pressure instruments) was inoperabl (1.0) You are withdrawing control rods during a power . increase from hot standb The reactor. power is about~15 percen What are three indications you would have that a control rod might be uncoupled? (Note: These .
indications might not occur simultaneously.) (1.0)
.&A (glm y b.' What action would you take to recouple the control rod, and what action is required if the control rod does not recouple-(assuming it can be moved with normal drive pressure). (2.0) Area Radiation Monitors are located throughout the plan However, automatic protective actions are initiated by_only 4uun of'the monitor mg. b) M ;. two Area Radiation Monitors ud1nitiate automatic protective actions? (1.0) What protective act!on will occur if the alarm trip point is reached on either of these instruments? (2.O)
Category Continued on Next Page RO. Examination Page 8 of 12
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Acril, 1985 Ovster Creek The RBCCW pumo breakers are eouipped with a two position control switch (Normal / Bypass). What is the function of this switch, and where is it positioned during normal plant operations? (1.0) If a loss of power and/or a LOCA occurs, how will the pumps respond for each of the two switch positions? Be sure to consider the response of the pumps for the switch initially being positioned in both Normal and Bypass, as well as the response of the pumps should the operator manually change the switch position after the event has occurre (2.0) 3.10 The reactor has been operating at 80 percent powe You are in the process of increasing power to 100 percent, and are currently holding at about 90 percen All three feed pumps are in operation with the Master Controller in Auto. You notice the flow for the 1C string is lower than the flows for the other two strings, which are well matche How can you balance the flows for the three
#T' . strings? (0.5)
k #~' r . -i If the flow in the 1C string does not respond to the action you have taken in Part (a) of this question, and it is determined that the lack of response is due to a Feed Reg Valve lockup, what two conditions could have
, caused the lockup? (1.5)
End of Category
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RO Examination . Page 9 of 12
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.'- April, 1985
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Oyster Creek l i ; I FROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL ' i CONTROL (25)
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4.' 1 0.P. 341, Standby Diesel Generator Operation, ! cautions that when testing is performed, the DG !
, shall not be restarted in a fast start mode during I
' a period of 15 minutes to three hours after I
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shutdow i ; ! What is the reason for this caution? (1.0) , i Hoes does this caution . af f ect the release of the 3 second Diesel Generator for maintenance? Can j the second Diesel Generator be started and/or i operated? Explain.your answe (1.5)
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{ q If a Diesel Generator is required to supply l power to its respective bus during this three j hour period, are you permitted to make it ' available? (0.5) ; i i l 4.2' With the reactor operating at 90 percent power, , according to ABN-3200.07, Unexplained Reactivity ! Change, r } #it List four indications you would have if a j reactivity LDECEGEg had occurre (1.0)
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j What three systems should be checked, and what j conditions should be looked for, to determine j the cause of the reactivity addition? (1.5) , ! i According to 0.P.324, Thermal Dilution Pumps, the Dilution Pump " Auto-Bypass" switches should always be in the " Auto" position, except that there are i; . four exceptions when the switches should be in the j " Bypass" positio %
- . Why should the pumps normally be operated with
the switches in the Auto position? (0.5)
!
i What are two of the four times that the j . switches should be in the Bypass position, and ' what is the reason.for having the switch in the j . Bypass position for each of-these times? (2.0) i 4.4 .What-are the six entry conditions for the RPV , Conteal Procedure, EMG-3200.017- (3.0) '
Category Continued on Next Page l } v
1 < RO. Examination , Page 10 of 12 i -
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,' April, 1985 Oyster Creek I i- .i
! Operating Procedure 340.1, 125 VDC Distribution
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Systems "A" and "B", cautions against interrupting ' , 125 VDC to certain 4160 VAC switchgear during
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normal operatio Explain this caution and the consequences of interrupting this powe (1.5)
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py 3 With the reactor operating at 30 percent power, a ' y 4e / turbine trip occurs due to a Main Generator faul l g f According to ABN-32OO.10, .
( List five automatic actions that will occur4 (1.5) }
- ,/ke*/ What effect will this transient have on '{
/ reactivity? Why? (1.0) > During plant heatup after reaching critical, f according to D.P. 201.2, '
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j What are two methods of verifying that recirc , j flow is sufficient for operation in Range 10 of l I the IRMs? (1.0)
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i e ! j By what pressure must the Cleanup Auxiliary j Pump be removed from service? (0.5)
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! How are the bypass valves kept closed prior to i t',])-
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.'y the reactor pressure reaching 900 psig? (0.5)
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l 1 The reactor is operating at 100 percent power when ! j a single relief valve inadvertently lift !
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j To which Emergency Operating Procedure will you ;
, be directed, and what entry condition (s) i specific to this event will direct you there? (1.0) [
Ij s What immediate actions are you required to take
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t i according to this procedure? Do not include
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any actions'that may be required after a t
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reactor scram may occu (2.0) i
. ! According to Administrative Procedure 106, Conduct ~
of Operations, what iggg' statements complete the following sentence regarding Shift Turnover
; "All Off-Gaing Operators shall not relinquish their 1 responsibilities until ne/she is satisfied that the-On-Coming Operator: .. . . "
(2.0) >
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4 RO Examination- Patye-11 of 12
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s April, 1985 Oyster Creek l l 4.10 According to ABN-32OO.01, Reactor Scram, one of the primary concerns after a scram has occurred is to l l maintain the RPV pressure below 1050 psig. Section
I 3.7 states that if the Main Condenser is not available, gc if the Turbine Bypass Valves are inadequate, four other systems or components may be used to augment RPV pressure contro List these l four systems or components along with any s restrictions on their use or availabilit (3.0) ! End of Examination
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s f a ma v o s/t Cycle efficiency = (Network out)/(Energy in) w = mg 2 s = V,t + 1/2 at
E = mc XE = 1/2 mv a=(Vf - V,)/t A = AN A = Ag e'* PE = agn .
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Vf = V, + a t - w = e/t
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1 = an2/t1/2 = 0.693/t1/2 t:1/2'## " EIt19)I*b}3
, [(t1/2) * (*033 aE = 931 as
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L= I,e'"* Pwr = w an I = I, 10-x/TVL f l TVL = 1.3/u
! P = P,10 "#II) . HVL = -0.693/v l P = Pa et /T . St2 = 26.06/T SCR = S/(1 - Kdf) ! /*.g CR,=S/(1-Kgf,) . Q* SUR = 26s/t= + (s - o)T CRj (1 - K df1) = CR 2 II ~ "df 2I' T=(t*/s)+((s-o)/#3 M = 1/(1 - Kg f) = CR /CR j , ! T = 1/(o - s) - M = (1 - K I T = (s - e)/(lo) d fo /II ~ Edf1) SOM = (1 - Kgf)/Xdf a = (X ,ff-1)/Xd[ * ^5eff/X eff t' = 10^ seconds
,, T = 0.) seconds ~I i
a = [(t*/(T K df I ) * (Idf (1 /
+ AT)]
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Idgj=Id P = *( reV)/(3.4.210 10 I d'.2,,,Z-g2 d ' ' - ~ j 22 I = oN R/hr = (0.5 CE)/d 2(=,g,73y Water Parameters Miscellaneous Conversions I gal. = 8.345 lb I curie = 3.7.x 1010dps I ga;. = 3.78 liters 1 kg = 2.21 lem . I f t' = 7.48 ga I hp = 2.54 x 103 Btu /nr l Density = 62.4 lbm/ft3 i l Density = 1 sm/c..13 1 mw = 3.41 x 106 Stu/hr ' lin = 2.54 cm Heat of vaporization = 970 Stu/lom 'F = 9/5'C + 32 l
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Heat of fusion = 144 8tu/1bm *C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. H ___-_ ---____ - _ _ - _ - - _ _ _ _ _
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Acril, 1985 I Oyster Creek PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS, HEAT TRANSFER. AND FLUID FLOW (25) - ANSWERS (1) Under conditions of high Xenon concentration, such as following a shutdown, the Xenon will be at the highest levels where the neutrcn flux was the highes These locations will be in the central core regions where the SRMs are locate This will tend to suppress the flux around the SRMs, while the flux in regions of the core where Xenon is lower in concentration will have higher flu (1.5)
(2) Relatively higher notch worths will be found in core regions that had relatively low neutron, flux levels during the previous reactor ' .17 operation because under conditions of f 'f with high Xenon, the flux is now relatively astartup) , 4 * higher in these regions. This would be in the ]qq , 'tl[ peripheral areas of the core, and particularly \' #. near the top of the core where boiling had been N t Y occurin (Other reasonable explanations also ,/d$;g)
, ,/ g4 acceptable.) (1.5) t REFERENCE: 0.P. 201.1, Section Reactor Theory, Chapter 8 (4 ;; ppg. 60 - 6 .2 As power increases, the amount,of boiling in a flow channel increases, and causes an increase in the resistance to flow - that is, core dP increase . (1) Two regions are used for core orificin The central region consists of all but the outermost ring of perimeter fuel with the smaller or fewer orifices in the outer fuel position (2) The orifices installed in the fuel supports add a relatively large dP to the total dP across the core. The dP added by boiling is then insignificant compared to the dP of the orifices, and so the dP is nearly the same (as is flow) to all fuel cells under any condition of powe REFERENCE: Lesson Plan, NSSS, ppg. 13 - 1 OC Exam Bank, Duestion NS- RO Examination Answers Page 1A of 10A
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April, 1995 Oyster Creek 1.3 (1) As the high pressure gas in the supply bottle is released into the accumulator, it expands rapidly and thus cools. As it warms to ambient temperature, the pressure in the accumulator will increas CBy filling the accumulator slowly, the gas will tend to reach equilibrium temperature and establish the proper pressure more rapidly.3 Therefore, you must wait until equilibrium is reached before the called for pressure can be se (1.5) Cl (2) If it were permitted to fill rapidly to f the pressure given in the procedure [ 540 - 620 ' psig 3 , as it warmed to ambient conditions the . j d, ,g/h gas pressure would increase to a value higher 3 50 {, pu, , n" than the procedure calls for, and an overpressure condition might result. ; g)yy nd s' , 950 & rsuaws . 5V*'-) **i,)y p el,y W* b. Normal pressure t operating conditions is ' yy .w a' about 1000 to psi (0.25) The charging water decreases the accumulator gas volume by about hal Note that the gas pressure is not the same as the charging water pressur (0.75)
'". REFERENCE: 0.P. 302.1, Sections 3.2.1 and .4 a. (1) Negative reactivity added from moderator coefficient during heatu (2) Negative reactivity added due to fission product poison buildu (3) Negative reactivity added during power operation due to void coefficien (4) Negative reactivity added during power operation due to fuel burnu (d *
O* der N.1.,1p,w * r. * .- Jpdk* dB /N Any thr sW A4tNe'NYforcredik.due40 V b. (1) Control rods (2) Burnable poison REFERENCE: Reactor Theory, Chapters 7 and .
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RO Examination Answers Page 2A of 18A
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April. 1985 Oyster Creek I
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P.at A new yt4%n _ _ _ ___ __ __ _ 4. _ Head .
, . - E t b nas o,.c 43 p, nt
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GFM Figure Pump Characteristic Curve kv Throttling the valve causes the friction losses
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1g '#/>)#{ in the system to increase, and the system curve di h,h)gpE
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g f}*p/h/ to move to the lef ' ' s .
, 0g e Changing pump speed is more efficient since a . p ./ lower flow rate can be achieved without , ;gU g# gv <
increasing the pump hea Increasing the pump b h I head would require more powe REFEF.ENCE: GE Thermal Sciences Text, (*~*} . ppg. 7-114 - 7-12 . Its reactivity worth will be les (0.5) Rod worth is proportional to the (local flux . divided by the average flux) squared. With all rods inserted, the average flux is lo Then wnen the rod is withdrawn, it causes the local flun to
- be relatively high, resulting in a relatively high
, reactivity worth.
] If the same rod is inserted from the fully withdrawn position when all other rods are mostly withdrawn, the flux depression caused by the inserted rod will result in a small value of (l ocal flux / avg flux)^ The worth of the rod in this case is thus much less than in the above cas REFERENCE: Reactor Theory, Chapter 8.
i RO Examination Answers Page OA of 18A
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' April, 1985 Oyster Creek Reactivity increases Power increases Void coefficient Reactivity increases Power increases Moderator temperature coefficiont REFERENCE: Reactor Theory, Chapter .8 Min permissible stable period = 30 see (0.5) DT = period /1.445 or 1.443 0. W
= 30/1.445 = 20.9 sec (tze) t/T P = Po e P/Po = 3000/5 = 600 In 600 = t/T t = 30 * In 600 = 192 sec = 3.2 min REFERENCE: Reactor Theory, Chapter 1 .P. 201.1, Section f Pressure Transients (1) Turbine trip without bypass above 40 percent power (2) Turbine trip with bypass above or below 40 percent (3) Turbine trip without bypass below 40 percent power (4) Generator load reject (5) MSIV closure (6) Loss of vacuum (7) Pressure regulator failure (8) ENRV closure Temperature Transients (1) Feedwater Controller f ailing at maximum demand ,
(2) Loss of Feedwater heaters t (3) Inadvertent Isolation Condenser initiation
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0.45 for any two of each type of transient 0.2 f or indicating that turbine trip without bypass above 40 percent is the most limiting transient REFERENCE: GE Thermal Sciences Text, ppg. 9-94 - 9-9 Lesson Plan # 201, ppe. 19 - 3 End of Category l
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RO Examination Answers Page 4A of 10A l I
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' April, 1985 Oyster Creek PLANT DESIGN, INCLUDING SAFETY AND EMERGENCY SYSTEMS (25)
- ANSWERS The vacuum breakers (check valves) will open after the EMRVs close (0.75)
when the discharge pipe cools and the pressure inside the pipe falls below that in the drywell (torus). (0.75) If the vacuum breakers did not function properly, the low pressure condition would cause a long water slug to rise in the discharge pip (0.75) This condition must be avoided because of the steam vent clearing phenomen If the water slug is too long, structural damage may eccur. (0.75) REFERENCE: Lesson Plan 9 5, ppg. 8 and 1 Exam Bank Qs # AD 7 and .2 Normally RSCCW is ops ~ated with one pump running and two heat exchangers in servic (0.5) For Shutdown Cooling operation, a second pump must be starte (0.5) f ,, j The discharge valve must be shut (0.5) or the pump must be running (0.5) to prevent short circuiting the vesse (0.5) REFERENCE: a. Procedure 309.2. Section 3.O. Procedure 305, Section 3.3.15 . d'
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RO Examination Answers Page 5A of 18A
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Apett, 1985 Oyster Creek (1) On a loss of solenoid power, the inlete#n G outlet. --d _ _ -_ _ :' ' -- val ves wi l l fait close (C VP) and the inlet purgw avalves will fail ope (0.214
.Ad cNSM (2) On a loss of instrument air, all valves will fail as i (0.3)
l If instrument air is lost, the inlet and l outlet valves only may be operated Cthrough five cycles 3 with air in air accumulators.
l REFERENCE: Lesson Plan # 50, Secondary Containment and SBGT, opg. 11 - 1 . (1) Cooling water to the TBCCW when Circ Water is s shutdown ( 0 ~) 7 )
-42i neanr.atnw E L cidr r' th: C w. i i. . i . . .. ._ , n. L. - ,
k--t _,m .. .v . _ f_i! ;nd . :- t rd du-i ; .;-- 21 _ : : _ . - u w. . . --
(3) Provides a manual backup of sealing water to the Circ Water pumps when normal Fire g Protection Water System Supply is unavailabl )
REFERENCE: Lesson Plan # 51, Service Water System, pg. *ry 7, . ?2 To preclude precipitation of the sodium pentaborat *
j b.A k&Water level in the tank above the heater _ 6"_ I REFERENCE: OC Exam Bank, Question # LP-5.
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RO Examination Answers Page 64 of 10A
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April, 1995 Oyster Creek
! a. The reserve pumo will start on low vacuum C18 l inches] in the vacuum tan j If a float valve should fait closed, air f i
entrained in the circ water will collect in the j high points of the system - primarily at the j top of the af f ected water bo (0.5) t I l i This will result in circ water not entering the i upper tubes, (0.5) i Resulting in decreased condensation of the ! turbine exhaust and reduced condenser efficienc (0.5) i
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REFERENCE: Lesson Plan # 14. Main Condenser Operation, ppg. 9 - 1 ' i l a. Low Low reactor water level C7' 2"3 (1 of 4) or High Drywell pressure C2 psig3 (1 of 4) b. If the priority pump discharge pressure doesn't - reach 100 psig within 10 seconds l g'V., c. After 230 psig system pressure is achtsved [ l
.j! , y to 9 Ad scuss 4 p4 ($h A)es. medQdo'rb When reactor pressure falls to 295 psig i
i e. Must have cleared or overridden all initiating l and auto start signals (0.5)
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Control by opening and closing the parallel valve (0.5) * REFERENCE: Lesson Plan # 10, Core Spray System. Figure - Core Spray Logi , O.P. 309, Section 5. , ;
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l f RO Examination Answers Page 7A of 10A i
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April, 1885 Ovster Creek 2.8 (1) Type I Low flow or outlet valve shut in service filter C80 gam 3 Aux pump cooling water high outlet tema C130 F3 . Non-regen Hx cutlet temp high C140 F3 l High system pressure C140 psig) , l SDLC on C15 gam 3 J 4, g,y gj %Q l Type II "'[ g '. pn A* l 1. High Drywell pressure C2 pote-3 . l Lolo Rx water level C7 W (2) Type 1 Isolates V-16-1, 7, and 14 Trtos system recirc pump Aux pump trips on two valves clostng Type II Isolates V-16-1, 2, 14, and 61 Trips systen recirc pumo Aux pump trips on two valves closing (3) The isolation must be reset using the Drywell Isolation reset switch for both types since the
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signal seats i (0.4)
* (0.2) per item in parts (1) and (2). Setpoints not require REFERENCE: OC Exam Bank, Duestion # CU-6 and Lesson Plan # 43, Reactor Cleanup .
System, pg. S.* ' l
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Is i l April, 1995 Oyster Creek l ! ! 2.9 (1) Vent valves C1, 19 and 5, 203 are normally ! open (O.25) l (2) Condensate return valves C 34 and 353 l normally closed (0.25) l (3) Condensate AC valves and Steam valves ' C30,31,36, an J 32, 33, 373 normally open (0.25) * The valve lineup provides for the venting of non-condensibles to the main steam lines (0.25) and for keeping the ICs war (0.25) The AC condensate return valve is kept open for IC reliabilit (0.25)
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b. In both cases, the suction and main discharge i valves of at least two loops must be kept open (0.5) , to assure a flowpath for natural circulatio (0.25) If two or more rectre pumps are operating, the i main discharge valves for the non-operating 7 loops must be closed, and the suction and ' discharge oypass valves must be ope (0.5) ! This is to assure forced circulation through the cor (0.25) REFERENCE: 0.P. 307, Sections 2.2 and i Lesson Plan # 21, Isolation , l tj'/ Condensors, ppg. 6 and 1 ,
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2.10 a. The normal power source in the main generator through the aux transforme (0.5) ; 1C and 1D are fed from 1A and 15, respectively.(0.5) ' I b. If power is lost from the aux transformer, all four buses will be powered from the startup tranaformer if 1t ie availabl (0.5) I If the startup transformer is not available, 1C and 1D will be powered from their respective diesel No power will be available for 1A and 1 (1.0) REFERENCE: Lesson Plan # 39 Plant Electrical . Distribution, ppg. 5 - 6, and P & ids 3001 and 3002.
l l , RO Examination Answers Page 9A of 19A
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April, 1985 Oyster Cronk l 2.11 High drywell pressure low low water level "b= _ c: ,.-., un : -'c- Tvir blod ink & v (00 , A 0;. i -ii.. L,ck e CC ;- r-d I e backup l , pump 4 dd k c)paci b .th., ESW / ( , REFERENCE: O. P. 310, Sections 3.3 and i End (d Category ' -
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:.u ,. .,1,e a M N L l}r;Jy, 1 di. Y Tr'da #v k
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( , , 0.5 $nw f ({rsesd 4% p sA r ep. h apped dt O M 4 Ye4 ESP) pw.. / F5 l
! ! RO Examination Answers Page 10A of 18A
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s + l, * l April, 1995 Oyster Creek t l
" Trip Unit Rod Block Scram Rosetable Inop Y Y l Downscale Y *
Int. Slope Y Y Normal Slope Y Y
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High Y High-High Y
* Get a reactor scram with the companion IRN upscale at High-High
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0.3 per named trip unit O.12 per correct Y or * I Lesson Plan # 37, Nuclear REFERENCE: Instrumentation, ppg. 21 - 2 .4 (1) The load reject scram will clear (be bypassed) when steam flow fails to less than 40 percent
, "*g of rated flo (0.5) ' ' ' (2) The mode switch must be taken to shutdown by procedure 3200.01 to permit SDV Hi bypas (0.5)
CTwoI.4eJe a es. % h t h p gsJ ,,H w a,o m :
(3) The high discharge volume trip may then be and must be bypassed to. vent and drain the SD (0.5)
e This can only be done after a 10 second time delay to give the scram time to complete the rod insertio (0.5)
(4) The scram reset switches may then be depressed to reset the scram Creenergizing the scram relays and pilot solenoid valves, and to supply air to the scram valve (0.5)
REFERENCE: Lesson Plan # 46, Reactor l Protection System, ppg. 22 - 25.
I OC Exam Bank, Question # RP-0.
l i - l RO Examination Answers Page 12A of 10A ,
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April, 1905 Oyster Creek 1- High pressure C1060 psigJ (0.25) sensed by the isolation condenser pressure l*/ . sensors (0.5) '.' Low low reactor water level (0.25)
[ sensed by the core spray level sensors (0.5)
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, vu ss ye,stal There i.,4 tw are 40ur usi }$hannel sensor " **$*l*YL*
s , and it takes Q l, two channels to cause a tri (0.5) The trip must be caused either by Channelu A
, and B together, or by Channels C and D togethe (0.5)
125 VDC power is thus provided to Channels A and C from one source CPower Panel DJ, and to ( Channels B and D from a sucond source CPower , Panel F (0.5) l REFERENCE: Lesson Plan # 40, Recirc Flow Control, Attachment 2.
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- (1) Isolation Condenser pressure 1 (2) RCP # 1 seal pressure l (3) Claanup System pressure
"" Any two of thre **
REFERENCE: OC Exam Bank, Question VI-6.
l (1) Change in flux !
(2) Rod Overtravel Alarm (3) The control rod position display goes l dark.
l (1) Drive the rod in until a power response is observed, or until it reaches "00".
(2) Withdraw the rod to its procrammed position, observing power respons (3) If "4G" is reached, perform a coupling ! check.
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(4) If it does not recouple, insert to "00" and isolate the HC REFERENCE: Lesson Plan # 45, Reactor Manual Controls, pg. 1 OC Exam Bank, Question # RN-5 ABN-32OO.06 Abnormal Control Sad Nation, Section ,
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s April, 1905 Oyster Creek Fuel Pool low range (C-9)
)flesah Aa-Reactor (D-9) Operating floor woutpment hatch areaf A )p .jj[[j
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VM H&ou Y lwd $t Two min timer starts 7, If the timer times out Isolates the reactor building normal ventilationf0f) (o.7f) Initiates the SDGT System (o,75 ) REFERENCE: Lesson Plan # 4, pg. .9 This switch will determine the response of the pumps during the sequential loading of the DGs , after a loss of electrical powe (0.75) l It is normally in the Normal positio (0.25) . l When in Normal, if a loss of power occurs with ! no LOCA signal, the pumps will auto start in sequence C166 sec If a LOCA signal is present, the pumps will not start Cannce thov aren't needed). (1.0) When in Dypass, the pumps will auto start in
sequence whether a LOCA signal is present or t f:
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not on a loss of powe (0.5) k If the switch was initially in Normal when a loss of power occured with a LOCA signal, it nust be taken to Dypass by the operator to start a pump if the operator has determined the need for a pum (0.5) REFERENCE: Lesson Plan # 41, RDCCW, pg. .10 Dy adjusting the bias adjust knobs on each of j the individual M/A station The feed reg valve lockup would have occured due to a loss of instrument air pressure C(70 psig3 ,k f gg) /r-or due to a 1cas pf electri al signa G."t(74) ? 3gg e 3
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RO Examination Answers Page 14A of 10A
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s * April, 1995 Oyster Creek l 4 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY, AND RADIOLOGICAL t CONTROL (25) - ANSWERS If a fast start is initiated during this I period, a potential failure of the turbocharger bearing may result. This may cause the DO to fail to start when it is neede The testing must have have been completed at least three hours prior to releasing the second l DG for maintenanc (0.75) ' The second DG may be started and operated as long as it is not started or cycled in the fast start mod (0.75) l Yes ! REFERENCE: 0.P.341, Sections 2.2.4, 2.2.11, and 2.2.12, and Standing Order # 2 .2 (1) MW electric increase (2) Reactor thermal power increases (3) LPRM readings increase (4) Steam flow increases
{(y;,p (5) Feed flow increases (6) Other reasonable parameter changes not i listed in the procedure may be accepted (1) CRD for abnormal control rod motion (2) Recirc for a flow abnoreality (3) Turbine Generator for a pressure regulator ea1 function .
Mag As. o h erasm Uc a sw wa e.a. less 4 F W k 4s REFERENCE: ABN-3200.07. Sections 2.2 and 3. '
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l C * is April, 1985 Oyster Creek ! , To insure the pumos will trip whenever a scram occurs.
! (1) When the intake temperature is above 60 I degrees to insure di ution in the discharge canal to preclude thermal shock during a scra (2) When thw weekly CRD scram brush recorder assurance check is being performed to preclude j a pump trip when the recorder is teste (3) When the pump is required to provide dilution flow during flow during an overboard (radioactive discharge) releas CIf a scram I occurs during this time, the release should be terminated and the pump trippe (4) When the reactor in, shut down because they aren't neede Also f ol lowing a scram, they , should not operate to minimize the rate of temperature decrease in the discharge cana .5 for condition and 0.5 for reaso REFERENCE: O.P. 324, Section w
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1 (1) RPV 1evel below the low level scram setpoint C135 inches]
(2) Drywe11 pressure above scram setpoint C psig3 (3) A condition that requires a reactor scram AND -
power is above 2 percent (APRM downscale trip)
' (4) RPV pressure above high reactor pressure scram setpoint C1050 peig3 (5) MSIV closure (6) A condition which requires a reactor scram in the judgement of the operator to etther conserve RPV inventory or to reduce the release of radioactivity into the environmen REFERENCE: EMG-3200.01. pg. 3.
l ' After interrupting the DC control power. the f. 6,o.Lc4 e4. eda breakers will continue to function as require gl.u k However, when re-applying CExamples certain breakers may tri DC, the DC trip coils on;{ g g g ,, _ are the ,4 ,Q feedpump, recirc pump, and RWCU pump breakers.] i w_ gg __- REFERENCE: O. P. 340.1, Section 2.2.5. and
, OC Exam Bank, Question # DC- ' O*Ph bO# %
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m %.M* w edrt u e !
RO Examination Answers Page 16A of 10A
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\ \ . April, 1985 Ovster Creek Alto dew.Tal1 1 (1) 230 AV breakers open " * ' ""
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(2) Turbine steam valves close *"****#**-N'^'
(3) Bypass valves open b ' '* M '
(4) Loads transfer to startup transformer L' 'O** .
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(5) DGs start and idle d ' I *** !** * * W (6) Main turbine aux oil pump starts 'b W ( (7) Turning gear att pump starts 4 L*i' O 4 '"
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b. Reactivity will increase because of a loss off ,, [at.*{ ,j M,
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4eedwater heatin I 4
% w. bed k. ABN- %EFERE *%
8.,.huaw 4-p q nom 00.10, Sectfons 6o3 J 3. 3.2, 3.6, 3.7.
l (1) Monitor the plant computer which provides a readcut in Ibm /h (2) Use curve in procedure to convert ope reading as a function of reactor pressure to minimum required flo b. Prior to reaching 125 peig CBearings and mechanical seats designed for only 150 pen c. By setting the NFR 100 psig above reactor 2 X.er d - ! pressur J 4 f., g,s .,,,6
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REFERENCE: 0.P. 201.2, Gections 4.t, 4.9, 4
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and 4.1 .0 a. Direction will be to EMG-3000.02, Containment Control, and specifically into TOR / (0.5) The specific entry condition is Torus water . l temperature above 90 degrees [possibly . lust I rising and approaching 90 degrees - CAF3 (0.5) b. (1) EHecute TOR /T. DW/T, PC/P, and TOR /L Me ' concurrentl Note that no action will be required in JGf'579 in the latter thre (2) Attempt to close the valve by 4 (a) cycling it ( (b) removing control power to it by pulling the fuses (O 4 (3) If the valve cannot be reumated, scram the reacto ( REFERENCE: EMG-3000.02, Containment Control and above indicated sub parts.
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RO Examination Answere page 17A of 1sA
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April, 1985 Oyster Creek ,
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r . "Is fully qualified, and/or licensed. to assume j the shift position." "Is fully aware of existing plant / equipment I ! conditione." ' Exact wardan, not required. - Ah, - ifM 'wJ..df 4J k s's 4*'Is W 4f f **/* A * ' i c REFERENCE: A.P. 104, Section 4.3.1.1.; A M He'I*I h '
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i ' 4.10 (1) Isolation Condensers Must trip A and E recirc pumps -
(2) EMRVs only if Torus Water Level t u hbove 90 l inches] Qc,ht I (3) Cleanuo only if boron injection is not reautred t (4) Main Steam Line Drains only 1i the Main Condenser is availabl [
0.25 for each system 0.5 for each restriction REFERENCE: ADN-3200.01, Section 3.7 and Table 2.
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