IR 05000277/1987027
| ML20238C433 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/21/1987 |
| From: | Howe A, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20238C428 | List: |
| References | |
| 50-277-87-27OL, 50-278-87-27OL, NUDOCS 8712310011 | |
| Download: ML20238C433 (170) | |
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U.S.. NUCLEAR REGULATORY COMMISSION REGION I 1 OPERATOR LICENSING EXAMINATION REPORT q EXAMINATION REPORT NO. 87-27(0L) l
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FACILITY DOCKET N0. 50-277/278 FACILITY LICENSE NOS. DPR-44 and DPR-56 l
LICENSEE: Philadelphia Electric Company i 2301 Market Street Philadelphia, PA 19101 FACILITY: Peach Bottom Units 2 and 3 EXAMINATION DATES: October 5 to October 7, 1987 CHIEF EXAMINEP.: M V /.a .4/ - (7 Allen G. Howe, Senior Operations Engineer Date l APPROVED BY: / s ('O
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David'J. Lange, Chief, BWR'Section, IA /2 2[d) Date / Operations Branch, Division of Reactor Safety SUMMARY: Written examinations and operating tests were administered to four (4) senior reactor operator (SRO) and three (3) reactor operator (RO) candidates. All SRO's and two (2) RO's passed the<e examinations. One RO candidate failed. j
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DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS: I RO l SR0 l l Pass / Fail l Pass / Fail- l l l 1 1 I I I I l IWritten l 3/0 .I 4/0 l ! I _I I i l i I i 10perating l 2/1 1 4/0 l l l l l l l l l 10verall l 2/1 l 4/0 I I I I I I I I I ( 1. CHIEF EXAMINER AT SITE: Allen G. Howe, Senior Operations Engineer ' 2. OTHER EXAMINERS: D. Lange, Chief BWR Section, DRS - D. Florek, Senior Operations Engineer I L. Kolonauski, Operations Engineer : B. Turner, Operations Engineer 1 G. Robinson, Consultant USNRC C. Gratt<n, Reactor Engineer 3. The following is a summary of generic strengths or deficiencies noted on ; operating tests. This information is being provided to aid the licensee ' in upgrading license and requalification training programs. No licensee response is required.
STRENGTHS a. Knowledge of content of administrative procedures.
b. Ability to locate reference material in the control room.
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DEFICIENCIES a. Knowledge of alarms and trips caused by flow converter failures.
b. After entering the E0P's, the candidates were hesitant to use the procedures to respond to the plant conditions in a real time manner.
4. The following is a summary of generic strengths or deficiencies noted from the grading of written examinations. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is required.
R0 EXAMINATION STRENGTHS a. Knowledge of the power - flow diagram, net positive suction head, and the void coefficient.
b. Knowledge of diesel generator operation, yrywell chilled water system, and IRM trips and ranges.
c. Knowledge of the failed jet pump symptoms and actions and symptoms for a high Scram Discharge Volume level.
DEFICIENCIES a. Ability to perform reactor period and shutdown margin calculations.
b. Knowledge of the core spray system restart on a new initiation signal after being manually secured. Knowledge of EHC system runbacks.
c. Knowledge of shutdown and scram requirements when a fire exists in the Diesel Generator building. Knowledge of the High Radiation in the Reactor Building Exhaust procedure (E-6).
SRO EXAMINATION STRENGTHS ' a. Ability to calculate plant heatup rate. Power calculations when given initial power and period.
b. Knowledge of the RCIC system, the Diesel Generators, the Core Spray system, and Scram Discharge Volume high level functions.
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c. Knowledge of failed jet pump indications, Tagging per A-26, and NRC i notification procedure A-31.
DEFICIENCIES 1 l i l a. Knowledge of LOCA effects on level instrumentation, knowledge of plant chemistry, knowledge of post LOCA conditions which prevent a steam - cladding reaction, b. Knowledge of plant events which cause a rod block, half - scram, or scram.
c. Knowledge of operating limits when a recirculation pump trips and knowledge of dose rate limits in radiation areas. ' l GENERAL Although all candidates passed the SRO examination, there was a clear separation in the sectional and overall scores between the candidates who had the highest scores and those that had the lowest scores. The cause for this could not be determined from the grading.
5. Personnel Present at Exit Interview: NRC Personnel A. Howe, Chief Examiner D. Florek, Senior Operations Engineer T. Johnson, Senior Resident Inspector R. Urban, Resident Inspector Facility Personnel l Dickinson Smith, Manager
- Drew Smith, Superintendent of Operations
! Richard Andrews, Training Coordinator 6. Summary of NRC comments made at exit interview: The chief examiner thanked the training staff for their preliminary work to ensure that the examinations went smoothly.
He commented'that the location for the written examination was excellent. He also noted that the. written examination review went smoothly and only minor changes were initial'y noted.
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, l The generic strengths and weaknesses noted on the operating examinations were discussed.
The processing of the waivers was discussed and the facility was-informed that final licensing action would not be made until the waived requirements had been completed.
7. Summary of facility comments and commitments made at exit interview: The facility staff expressed their appreciation for the cooperation from the NRC prior to examination administration.
They also stated that the written examination was a fair test for the candidates, but they were also concerned that.the clarification given to individual candidates and not to the group may cause inconsistent answers from the candidates. i 8. Several eligibility requirements were waived to enable some of the candidates to take the examination. The facility-should review these requirements and ensure that they are completed in.a timely manner when conditions permit. Upon completion of the requirements, the facility , ' should notify the NRC so final licensing action can be made. 1 Attachments: 1. Written Examination and Answer Key (RO) q 2. Written Examination and Answer Key (SRO) i 3. Facility Comments on Written Examinations after Facility Review . 4. NRC Response to Facility Comments
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I U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION l l Facility: Peach Bottom 2 & 3 Reactor Type: BWR-GE4 , Date Administered: 87/10/05 Examiner: G. E. Robinson Candidate: M ASTEL
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INSTRUCTIONS TO CANDIDATE j Use separate paper for the answers. Wri te answers on one s ide only.
Staple question sneet on top of the answer sheets. Points for each question are indicated in parentheses af ter the questlon. The passing grade requi res 2 at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours af ter the examination starts. '
% of j Category % of Candidate's Catego ry )
Value Total Score Value Ca tegory 25.0 25.0 1. Principles of Nuclear Power Plant Operation, Thermodynamic 3, Heat Transfer and Fluid Flow 25.0 25.0 2. Plant Design including Safety and Emergency Systems 25.0 25.0 3 Instruments and Controls 25.0 25.0 4. Procedures - Normal, Abnormal, ; Emergency, and Radiological Control < i 100.0 TOTALS l Final Grade % j l
i All work done on this examination is my own. I have neither given nor I received aid. l l I Candidate's Signature j
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the admini stration of this examination the f ollowing rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties. - 2. Restroom trips are te be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3. Use black ink or dark pencil only to f acilitate legible reproductions.
4. Print your name in the blank provi ded on the cover sheet of the ex ami nat i on . 5. Fill in the date on the cover sheet of the examination (if necessary).
6. Use only the paper provided for answers.
7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
B. Consecutively number each answer sheet, writt "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
9. Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least thcee lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face , down on your desk or table. ! l
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12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO.NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has i bsen completed. )
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18. When you complete your examination, you shall: a. Assemble your examination as f ollows:
- (1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer, b. Turn in your copy of the examination and all pages used to answer the examination questions.
c. Turn in all scrap paper and the balance of the. paper that you did not use for answering the questions.
d. Leave the examin; tion area, as defined by the examiner. If after l eavi ng , you are found in this area while the examination is still in progress, your license may be denied or revoked.
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1. PRINCIPLES OF I:UCLEAR POWER P,LANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (2.0) . Using the enclosed Power-Flow diagram (T-LOT-0040-6), answer the following: l a. Insertion of control rods at full power with constant (1.0)~ pump speeds results in an increase in core flow.
Briefly explain why.
b. What damage could be expected if operation is.allowe'd (0.5) below the Minimum Flow Control Line? c. What Interlock precludes and prevents operation (0,5) below the Minimum Flow Control Line? l QUESTION 1.02 (3.0) For each of the following changes, will reactor thermal, power increase, decrease, or remain the same? Briefly explain each answer.
I a. Condenser vacuum goes from 29" Hg to 26" Hg (1.0) b. Isolation of extraction steam to the Feedwater (1.0) Heaters at 90% power, c. Sudden increase in reactor pressure (prior to (1.0) ;
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a reactor scram) QUEST 10N 1.03 (3.0) . a. Consider the Xenon transient following a power decrease from 100% to 50% af ter several months of operation at full power.
I. Will the peripheral control rod worths . (2.0) i increase, decrease, or remain about the-same during the Xenon peak? Briefly explain ,
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your answer, 1 II. If the decrease'in reactor power was from (0.5) 100% to 50%, would the reactivity due to l equilibrium Xenon at 50% power be more than, less than, or about equal to one half of
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j the 100% equilibrium value? b. TRUE or FALSE? When starting with a Xenon free core, (0.5) It takes longer to reach equilibrium Xenon at 100% power than it does at 50% power. ;
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l PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT l 1.
TRANSFER AND FLUID FLOW l QUESTION 1.04 '(2.0) .
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a. Your reactor has just been declared critical af ter a (1.5) refueling outage. Because of a rod drive problem', one rod is fully inserted and disarmed and the reactor is now
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subcritical. The countrate does not fall to the original countrate prior to startup (when all rods were inserted).
' Briefly explain why.
b. TRUE or FALSE? For a given amount of positive reactivity (0.5) insertion while the reactor is subcritical, the closer the reactor is to critical the larger the change in countrate.
j QUESTION 1.05 (2.0) l a. Explain what happens within a pump when the "available" (0. 5) NPSH (net positive suction head) drops below the value of the minimum "requi red" NPSH.
A b. Indicate whether available NPSH will increase, decrease, or remain about the same for the following. Consider each I change separately.
I. For the Reactor Recirculation Pumps - reactor (0.5) water level is increased.
II. For the Reactor Recirculation Pumps - reactor (0,5) pressure is increased.
Ill. For the CRD pump - system output flow is (0,5) ) significantly increased.
QUESTION 1.06 (2.0) a. Define Critical Power (0.5) i b. Will the Critical Power initially increase, decrease, (1.5) ! or remain the same if a Feedwater Heater is bypassed while at full Power? Briefly justify your answer.
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l 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUE; TION 1.07 ( 1. 5) Your reactor has just reached cri ticality af ter a refueling outage (1.5) and is placed on a ninety (90) second period. Thereafter no rod movement or recirculation flow changes occur. Af ter some titre has elapsed, you find that the power increases by a factor of 10 in 300 seconds. Have you reached the heating range? Justi fy your answer and show all work.
QUEST 10N 1.08 (2.0) During heatup the following data was obtained (2.0) TIME REACTOR PRESSURE 1: 30 52 psig 2:00 102 psig 2: 30 180 psig From 1: 30 to 2: 30 what was the heatup rate? Show all work.
QUESTION 1.09 (2.5) a. You have been informed that the Shutdown Margin (SDM) for (2*0) your reactor is 3%. You have an indication of 300 CPS on the SRM instrumentation. Control rods are wi thdrawn and your countrate on your SRM's increases to 800 CPS. What is your new shutdown margin? b. While subcri tical, your reactor coolant temperature increases. (0.5) Would your Shutdown Margin increase, decrease, or remain the same? i i
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- - _ - . -- - , _ , _ _ _ _ _ - _ _ _ _ _ _ _ - _ . _ _ _ - _ _ . - _ - _ _ - _ _ - _ _ - _ _ _ . _ _ _ _ _ - 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTidN 1.10 (2.0) If one recirculation pump trips while at power, OT-112 requires the operator to reduce power by driving in deep rods. Assume that a scram has been avoided and a restart of the pump is planned a. Why does the. procedure require 6he operator to continus (1.0) to insert rods prior to pump restart? b. Explain the problem that. could occur if shallow (1.0) rods were inserted instead of deep rods.
QUESTION 1.11 (1.5) Consider the inadvertent closure of one MSIV while. operating (1.5) at 70% power. Once the transient has stabilized, will the reactor pressure be greater, less, or the same as before the MSlV closed? Briefly justify your answer.
QUESTION 1.12 (1.5) Indicate whether the change -.in condi tions given below would cause the VOID COEFFICIENT to become more negative, less negative or have no effect. Consider each change independently, a. An increase in reactor pressure (0.75) b. Removal of several control rods while. power (0.75) remains constant (due to burnup).
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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QUESTION 2.01 (3 0)
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Consider the Reactor Core isolation Cooling System-a. Indicate the conditions which would cause an' (0. 5) - automatic swapover. of the RCIC water supply from the condensate storage tank to the torus. i
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indicate appropriate set points.
b. Briefly explain how RCIC System response differs (1.5) from a high reactor water level trip as compared to a RCIC turbine trip and indicate the reason .; for this difference. ! c. What is the minimum RCIC turbine speed allowed and (1.0).
why is this limitation required? QUESTION 2.32 (3.0) a. Identify the below IIsted components by matching' them wi th the ap;,ropriate letter from Figure' 1.
Drive Piston. (0.25) Exhaust Scram' Valve (0.25)- l Index Tube .(0.25) Col.l et : Fi ngers (0.-25) b. Using Figure I as a guide, describe the flow path (2.0)
(or sketch it on Figure 1) through the' drive mechanism on an Insert signal.
QUESTION 2.03 (3.0) l Consider the Standby Liquid Control System (SBLC)' a. Why must the solution in the SBLC storage tank for (0.5)
' Unit 3 be heated? -
, l b. What three automatic actions.will occur when this (0.9) ! system is placed in operation manually with the Key Locked Switch? i c. Excluding alarms, IIst'four Indications ~ that the control -(1.6) room operator could use to veri fy that SBLCis injecting' , into the vessel. (assume that It is too early . to. detect-any changes- in power)
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i l 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS l QUESTION 2.04 (3 0) Consider the Condensate System a. What three (3) valves automatically : lose on low (1.0) pressure in the condensate pump discharge header? b. TRUE or FALSE 7 The plant can operate within (0 5) { administrative 1Imits at 75% power wIth one of the
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three heater st rings valved out of service.
c. What two (2) automatic actions occur if one condensate (1.5) pump trips while the reactor is at full power? ! QUESTION 2.05 (2.5) a. An unexpected closure of the inboard MSIV on "A" steamline occurs while the reactor is operating at 70% power. For tr.e equipment or systems listed below, Indicate whether or not their steam supply would be interrupted.
i. High Pressure Coolant injection Turbine (0.25) 11. Reactor Feed Pump Turbines (0.25) til. Steam Seals (0.25) iv. Steam Jet Air Ejectors (0.25) b. TRUE or FALSE 7 If FALSE change the statement so that it is true.
1. For all modes of operation, the' MSlV (0 75) Isolation for high radiation level in the main steam line :s set at 30R/HR II. Isolation of any two main s team lines by (0.75) MSlV closure when the mode switch is in run alway; produces a scram.
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (3 0) ) QUESTION 2.06 Consider the Diesel Generators a. Briefly explain the differences in system equ'ipment (1.0) response between a manual start using the manual start ; switch and a manual start using the QUICK { START Button when starting the diesel generator j f rom the control room. ] i' b. On a Maximum Credible Accident start, list four (4) Diesel Generator trips that are not bypassed. (2.0) QUESTION 2.07 (2.5) Consider the Drywell Chilled Water System (DCWS) q a. Assume that complete loss of Drywell Chilled water (1,0) has occurred while at full power. What three (3) heat loads are affected? b. While Unit 2 is operating at full power, the #2 (1.5) 13 K/ bus is lost. List the automstic actions that should occur in order to maintain cooling to the drywell .
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (2.5) j QUESTION 2.08 Consider the Core Spray System a. The reactor pressure is. 350 psig and the core, spray -(1.0) system auto initiated because of low water level in the reactor vessel. , When water level: recovered, all four pumps were taken to STOP by the operator.- If the water level again falls below the initiation signal, will the pumps auto start? Briefly explain why. they will or will not auto start. Assume no operator action.
b. TRUE or FALSE? 1. Each of the' four. Core Spray pump motors - (0. 5) : and associated automatic' motor . operated valves receive power from di fferent - emergency auxiliary buses.
11. It is 'possible for the operator, from the (0.5) control room, to manually line up the. core spray system so that pump suction is taken from the Condensate Storage Tank instead of the torus.
Ill. It is possible to manually open the out- (0.5) board motor operated injection valve from the control room if the inboard valve. ls.
closed and the reactor pressure is above 450 psig.
QUESTION 2.09 (2.5) Consider the Shutdown Cooling Subsystem a. Briefly explain how the design of the head spray (1.0) subsystem in Unit 3 differs from that in Unit 2.
b. For shutdown cooling operation, why must power (0.5)
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be. secured to the minimum flow valve? . c. List the valves that are automatically shut on a (1.0) Group iib isolation. ,
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3 INSTRUMENTS AND CONTROLS QUESTION 3 01 (1.0) While suberltical during a cold startup, a rod block (1.0) occurs while attempting to wi thdraw the first rod.
All neutron monitoring mode switches are positioned in
" operate" and all power supplies are normal . Give two possible reasons for the rod block. Do not include rod .
blocks generated by RSCS or RWM.
QUESTION 3.02 (3 0) i A reactor startup is in progress. The Rod Worth Minimizer has been bypassed due to a malfunction. The RSCS is controlling.
a. All rods in Groups Al and A2 are full out. The Rod (0. 5) Sequence Selector Switch (RSS) is in the A34 position.
How far can the first rod in Group 3 be withdrawn? b. All rods in Groups Al, A2, A3, and A4 are full cut (1.0) and the Rod Sequence Selector Switch and Sequence Mode Selector Switch are in the NORMAL position and the reactor is at 10% power. All rods in Group 5 have been pulled to position 4 and one group 5 rod has been pulled to position 6. If the reactor power is increasing faster than desired, which rod (s) can be inserted? Briefly explain, c. Briefly explain the LOW POWER SET POINT. I ncl ude (1.5) how i t af fects the RSCS system operation. Indicate its sensing point and associated power level.
QUESTION 3.03 (3 0) Consider the Reactor Recirculation Flow Control System a. Indicate which of the conditions given below would (1.0)
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inhibit a recirculation pump start 1. Discharge valve open 11. Lube oil pressure 35 psi 111. Generator Field Breaker Open-Iv. M-G Set Supply Air Isolation Gates inserted.
b. What are the two actions utilized to reset a Scoop (1.0) Tube Lock once the initiating condition has been cleared? c. Briefly explain why the reci rculation pump must be (1.0) star,ted with a minimum of a 10% speed control signal .
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3 INSTRUMENTS AND CONTROLS QUESTION 3.04 (3 0) Consider the Wide Range Yarways a. How far above the top of active . fuel (TAF) is '
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instrument zero? b. Following a surveillance, an equalizing valve for 1(1.0)- reactor level instrumentation is lef t open. Would' the Indicated water level increase, decrease, or remain the same? Briefly explain your answer.
c. At a reactor pressure of 200 psia, does water level (1.0)- read higher or lower than the actual water level? Briefly explain your answer.
d. TRUE or FALSE 7 With a higher than normal drywell (0.5) temperature caused by a small primary coolant break, Indicated water level will read lower than actual water level.
QUESTION 3.05 (3.0) Consider the Automatic Depressurization System (ADS) .(1.0) a. With the following conditions present, triple. low . water level, high drywell pressure and confirmatory low water level, list the other requirements that must be met before reactor depressurization will take place.
b. For UNIT TWO, what four means are available for (2.0) reclosing the ADS Valves once initiation ,has occurred?
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3 INSTRUMENTS AND CONTROLS _
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QUESTION 3 06 ' For each of the lettered conditions given below, Indicate which cf the following will AUTOMATICALLY occur: (if i.e., more than one half-scram 1s action occurs, state the most severe action, Assume no operator actio'n.
more severe than a rod block) . I Scram 11 Hal f-s cram Iii Rod block IV No action (0.5) a. During startup and operating on Range 3, low voltage to IRM's "A" and "H" occur.
(0.5) b. At 20% pcwer, all four turbine stop valves close.
(0.5) c. At 10% power, foss of voitage to "C" APRM occurs.
(0.5) d. While operating at 60% recirculation flow, power is increased to 85%.
(0.5)
e. At 60% power, the APRM flow converter fails down-scale.
(2.0) QUESTION 3 07 (1.0) a. The control room high radiation alarm occurs. Indicate the change in fan configuration which automatically occurs. Assume the Control Room fan configuration was normal prior to the alarm.
l (1.0) b. Which of the combinations given below of the Control Room Ventilation Supply RAD Monitors would cause an isolation of the control room vent?
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I Hi A and Falled B 11 Hi H1 A and Hi B
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111 Falled A and Hi HI B (3.0) QUESTION 3 08
" Synchronous Speed _NotList Selected" is an automatic the remaining runback associated two runbacks wi th. the EHC- Logic System.
associated with the EHC 'and for each:
(2.0)
a. Give the associated initiation signal and (1.0)' b. State the final turbine load condition for each ! runback. i
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3 INSTRUMENTS AND CONTROLS QUESTION 3 09 (1.5) During operation at rated conditions, it is discovered (1.5) that the scram discharge volume high water level bypass switch has been in the bypass position since startup.
Could this have prevented a VALID high Instrument volume scram from occurring? Briefly explain your answer.
QUESTION 3.10 (1.5) While operating at full power, a slow leak develops in the (1.5) ins trument air line. Assume this leak is beyond the make-up capacity of the air compressors even with service air isolation. Assuming no operator action, how will reacter poaer be af fected. Briefly explain your answer.
QUESTION 3.11 (1.5) With the mode switch in STARTUP and the IRM "C" reading 11 (1.5) on Range 7, what trip (s), i f any, would occur if IRM "C" was downranged to range 67 Briefly justify your answer.
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- _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL QUESTION 4.01 (3.0) Indicate whether the following conditions require an entry to: 1. RPV control (T-101) 11. Containment Control (T-102) lii. both T-101 and T-102 Iv. Not an entry condition a. Drywell temperature 150*F (0.5) b. Drywell pressure 2.8 psig (0.5) c. RPV level at -39 inches (0.5) d. Torus Temperature 92*F (0.5) e. A Group 1 Isolation occurs (0.5) f. Power is 6% af ter a turbine trip f rom full power (0.5) QUESTION 4.02 (3 0) a. Indicate the PBAPS employee administrative whole body (0.5) limi ts for radiation exposure received per quarter, both with and without a NRC Form 4.
b. An unlocked HIGH RADI ATION AREA is encountered: i. What is the minimum and maximum radiation (0.5) level you would expect to find in this area? II. List the three (3) entry requirements which (1.5) must be met to enter this area.
c. State the recommended upper dose limi t (whole body) for emergency exposure in accordance with EP-207 1. to save a life (0.25) 1 1.. to operate equipment to mitigate an emergency (0.25) !
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4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL QUESTION 4.03 (2.0) In accordance with ON-106, Stuck Control Rod a. What restrictions are placed on attempted (1.0) rod movement of the stuck rod? b. While attempting to . eve the stuck rod, what (1.0) three (3) CRD System parameters should be monitored? QUESTION 4.04 (3.0) A fire is reported in the Diesei Generator Building
; (ON-114) i a. Under what circumstances would a controlled shutdown (1.0)
be required in accordance with ON-il47 b. Under what circumstances would ON-Il4 requi re the (1.0) operator to initiate a scram? c. ON-ll4 Indicates that it is important that the (1,0) recirculation pumps do not trip when a scram occurs.
Briefly explain why.
l QUESTION 4.05 (1.5) l In accordance with the Procedure for Corrective and Preventive Maintenance Using Champs (A-26A) l a. What does an Operation Verification Tag hung on (1.0) ' a component indicate? b. TRUE or FALSE? It is not mandatory to place (0 5) l Information Tags on equipment located in Radiation Areas.
QUESTION 4.06' (1.5) q
'1 in Procedure T-101, Reactor Control, NOTE #20 states (1.5) 1 i " Don't disable ECCS auto initiation unless misoperation is confirmed or adequate core cooling is assured."
What are the two (2) different conditions that define misoperation of ECCS? l l
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4. PROCEDURES - NORMAL, ABNORMAL,. EMERGENCY AND RADIOLOGICAL CONTROL' )
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QUESTION 4.07 (2.0) PBAPS ON-100 lists symptoms of a failed Jet pump. For each of .] the parameters listed below, state whether en increase, decrease ] or no change would be seen if a failed jet pump occurr'ed. 'j a. Reactor power (0. 5) b. Core Plate delta pressure indication (0.5) c. Drive flow in the defective Jet pump loop (0 5) d. Delta Pressure on the Jet pump sharing a riser.with the failed jet pump (0.5) I QUESTION 4.08 (2.5) .q According to SE-1, Plant Shutdown from the. Emergency Shutdown Panel, it is required that the operator place the drywell instrument air in service prior to leaving the control' room, a. Explain the manipulations required by.the operator in order to place the instrument air in service. (1.5) b. Briefly explain WHY this step is required. (1.0) QUESTION 4.09 (2.0) The first indication of a high level in the Scram Discharge Vol' me u (SDV) is likely to be the "SDV NOT DRAINED" alarm. This requires-entry into OP-105, SDV High Level a. What two (2) immediate actions are required? (1.0).
b. In addition to the "SDV NOT DRAINED" alarm, what two (2)- automatic actions should be verified if the level continues to rise? Include setpoints. (1.0) QUESTION 4.10' (2.0) While operating at 60% power, alarms annunciate in the control room that Indicate High Radiation in the Reactor Building' Exhaust (Procedure E-6) a. In addition to the alarms, what two (2) automatic actions occur? (v.0) b. List the immediate operator actions. (1.0)
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QUESTION 4.11 (2.5) in accordance wi th the Normal Startup Procedure, GP.-2 I a. Under what condi tion should the " Emergency Rod In" (0.5) switch be used? i f b. Explain the reason why the use of the " Emergency Rod (1.0) l in" switch should be minimized. I c. Why must the EHC be in service prior to establishing (1.0) criticality?
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PROPERTIES OF SATURATED STEA.M AND SA'"JRATED WATER (TE.MPERA*"JRE ) Volume, ft'/lb Enthalpy, 8tw/lo Entropy. Etw/le a r Te o rs. Water Evan Ste am water Evan Steam water Evao Steam '['
*t 'ry "g hg h, q h, Sq 3,y S g
32 0.08859 0.01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 1 35 0.09991 0.01602 2946 2948 3.00 1073.8 1076 8 0.0061 2.1706 2.1767
35 l 0.12163 0.01602 2446 2446 8.03 1071.0 1079.0 0.0162 2.1432 2.1594 40 ! 45 0.14744 0.01602 2037.7 2037.8 13.04 1068.1 1081.2 0.0262 2.1164 2.1426 45 ! 50 0.17795 0.01602 1704.S 1704.8 18.05 1065.3 1083 4 0.0351 2.0901 2.1262 50 j 60 0.2561 0.01603 1207.6 1207.6 28.06 1059.7 1CS7.7 0.0555 2.0391 2.0946 60 1 70 0.3629 0.01605 BEE.3 865 4 38 05 10540 1092 1 0.0745 1.9900 2.0645 70 EO OSMI 0.01EC7 633.3 633 3 48 04 104E 4 10964 0.0932 1.94~5 2.0359 EO I 90 0EH1 a01510 465.1 465.1 55.02 1042.7 1100.6 0.1115 1.8970 2.0'.56 90 l 100 0.9492 0 01613 350 4 350.4 68.00 IC3 7.1 1105.1 0.1295 1.8530 1.9825 100 { 110 1.2750 0.01617 255.4 255.4 77.98 1031.4 1109.3 0.1472 1.8105 1.9577 110 120 1.6927 0.01620 203.25 203.26 87.97 10256 1113.6 0.1646 1.76e3 1.9339 120 130 2.2230 0.01625 157.22 157.33 97.96 1019.8 1117.8 0.1817 1.7295 1.9112 130 140 2.SS92 0.01629 122.98 122.00 107.95 1014.0 1122.0 0.1985 1.6910 1.8E95 140 150 3.718 0.01634 97.05 97.07 117.95 1008.2 1126.1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 170 5.993 0.01645 62.04 62.06 137.97 9962 1134.2 0.2473 1.5822 1.8295 170 180 7.511 0.01651 50.21 50.22 148 00 990.2 1138.2 0.2631 1.54E0 1.8111 180 190 9.340 0.01657 40.94 40.96 15S.04 984.1 1142.1 0.2787 1.5145 1.7934 190 * 200 11.526 0.01664 33.62 33.64 168.09 977.9 114 E.0 0.2940 1.4824 1.7764 200 210 14.123 0.01671 27.80 27.E2 17E.15 971.6 1149.7 0.3091 1.4509 1.7600 210
. 212 14.696 0.01672 26.78 26.50 *80.17 970.3 1150.5 0.3121 1.4447 1.7568 212 220 17.186 0.01678 23.13 23.15 1B8.23 965.2 1153.4 0.3241 1.4201 1.7442 220 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3902 1.7290 230 240 24.968 0.01693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13.802 13.819 ~ 218.59 945.4 1164.0 0.3677 1.3323 1.7000 250 260 , 35 427 :.01709 11.74! 11.762 225.75 9356 11E7 4 0.3519 1.30*1 1.5552 260 .'
270 41.E56 0.01718 10.042 10.060 238.95 931.7 1170 6 0.3960 1.2759 1.6729 270 280 49200 0.01726 8.627 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 280 290 57.550 0.01736 7.443 7.460 .259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7- 0.4372 1.1979 1.6351 300 310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0.4506 1.1726 1.6232 310 320 89.64 0.01766 4.896 4.914 290.4 894.8 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0.01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 360 153.01 0.01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 380 195.73 0.01536 2.317 2.335 353.6 844.5 1198.0 0.5416 1.0057 1.5473 380 400 247.26 0.01864 1.8444 1.8530 375.1 825.9 1201.0 0.5667 0.9607 1.5274 400 420 308.78 0.01894 1.4808 1,4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 440 381.54 0.01926 1.1976 12169 419.0 785.4 1204.4 0.6161 0.8729 1.4890 440 460 466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 480 566.2 0.0200 *- 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4518 480 500 680.9 0.0204 0.6545 0.6749 487.9 714.3 12022 0.6890 0.7443 1.4333 500 520 812.5 0.0209 0.5356 0.5596 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 540 962.8 0.0215 0.4437 0.4651 536.8 657.5 1194.3 0.7378 0.6577 1.3954 540 560 1133.4 0.0221 0.3651 0.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 560 580 1326.2 0.0228 0.2994 0.3222 589.1
;
589.9 1179.0 0.7876 0.5673 1.3550 580 l 600 1543.2 0.0236 0.2438 02675 617.1 550.6 1167.7 0.8134 0.5196 1.3330 600 620 1786.9 0.0247 0.1962 02208 646.9 506.3 1153.2 0.8403 0.4689 1.3092 620 640 2059.9 0.0260 0.1543 0.1602 679.1 454.6 I 1133.7 0.8666 0.4134 1.2821 640 660 2365.7 0.0277 0.1166 0.1443 714.9
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392.1 1107.0 0.8995 0.3502 1.2498 660 i 680 2708.6 0.0304 0.0808 0.1112 758.5 310.1 1068.5 0.9365 0.2720 1.2086 680 i { 700 3094.3 0.0366 0.0386 0.0752 822.4 172.7 995.2 0.9901 0.1490 1.1390 700 705.5 32082 0.0508 0 0.0508 906.0 0 906.0 1.0612 0, 1.0612 705.5
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o EQUATION SHEET f = ma v = s/t Cycle efficiency = (Network out)/(Energy in) w = mg s = V t + 1/2 at 2 E = mc2 KE = 1/2 my A = IN a = (Vf.- Vg )/t . A'= A e4D PE = mgh V =V + at w = e/t 1 = in2/t f =0.693/tg
*
NPSH = P in - Esat 1/2 1/2' b))
,
m a oAV
[(t1/2}'* ( b)]
.
AE = 931 am
, I=Ie'* g 0 = mCpat-Q = UAah !'= 1,e~"*
Pwr = W ah f I = I, 10~*/IY' - TYL = 1.3/u-P = P,10 sur(t) HYL = -0.693/v P = Po et /T SUR = 26.06/T SCR = S/(1 - K,ff).
CR x = S/(1 - K,ffx) SUR = 26o / t* + (8 - c )T CR)(1 - K,ff)) = CR2( -k eff2)-
'
T = (t*/o ) + [(a - o )/lo ) ' M = 1/(1 - K,ff} = CR)/CR g T = t/(o - a) M = (1 - K,ff,)/(1 - K,ff),' f T = (a - o)/(ao) SDM = (1 , .X,ff)/K ,ff 1 o = (X,ff-1)/K,ff = AK,ff/K,ff t' = 10** seconds-
.
1 = 0.1 seconds-I o = [(t*/(T K,f f)] + [s,f f/ (1 + b))
=1d P = (r6V)/(3 x 1010) .I)d)
I d) 2 =2 1d2 2-2 i r = eN 2 R/hr = (0.5 CE)/c (meters) NPSH = Static head h)-- P sat R/hr = 6 CE/d2(feet) I Water Parameters Miscellaneous Conversions l 1 gal. = 8.345 lbm.
1gaj.=3.78 liters I curie = 3.7'x 10l0 os I kg = 2.21 Tom . I ft- = 7.48 gal. . I ap = 2.54 x 0" Staine.
Density = 62.4 lom/ft4 i' Density = 1 gm/cm* '
.
I mw'= 3.41 x 106 Stu/hr Heat of vapor.ization =.970 Btu /lbm lin = 2.54 cm
*F's g/5'r + 32 Heat of fusien =: 144 Stu/lbm.
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. I Atm = - 14.7 psi = 29.9 in. hg. * C = 5/o {'F-32) ;
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Wt fr w 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS.
HEAT TRANSFER AND FLUID FLOW ANSWERS -- PEACHBOTTOM 263 - 87/10/05 - G. E. ROBINSON ANSWER 1 01 (2.0) a. Reduced power lowers the core and separator pressure drop (1.0) thereby reducing resistance to flow. Thus core flow increases. or ce~e 4 P crops ca a recuit er rea'uchm o c qui e p e are. ola n-s/ //ow . } b. Recirculation and/or Jet Pump Suction Cavitation (0.5) l c. Recirculation pumps cannot be operated above minimum (0.5) spe'e d, 30%, until total feedwater flow is 20% (20% feedwater flow Interlock) . REFERENCE LOT 40, Reci rculation Flow Control , LO 11, pgs. 18, 19 K/A 202002 Reci rculation Flow Control System l Kl. Knowledge of cause-effect b'etween RFCS and l Kl.02 Reactor Power (4. 2) Kl .0 3 Reactor Core Flow (3 7) K4.06 Knowledge of RFCS interlocks which prevent NPSH (3.1) , e I
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW G. E. ROBINSON ANSWERS -- PEACHBOTTOM 2s3 - 87/10/05 - ANSWER 1.02 (3 0) f (0.25) a. Power remains the same (0.75) Increase in condenser pressure causes a decrease in plant efficiency but does not cause a change in thermal power power increases * aN-P"< W d~~%d b. kh.
Due to cooler water entering reactor, there is an increase In water density (shorter slowing down and diffusion lengths) less ' neutron leakage (moderator temperature coef fi ci ent)
(0.25)
W c. power increases (0. 75) due to collapse of vol ds , increase in water densi ty, less neutron leakage (Void Coefficient) Also cowk. reaglWy efech # {kkag & REFERENCE LOT 1230 Rankine Cycle, LO3, pg. 6 LOT 1250 Plant Efficiency , LO2 LOT 1440 Reacti vi ty Coeffi cients and Defects, LO 3, 4, pg. 3 K/A 259001 Reactor Feedwater System Know.of the effect that malfunction of Teed Water System K3 12 has on Reactor Power (3.8) EPE 295025 High Reactor Pressure EK 1.01 Know.of the operational Implication of press. effects on reactor power (3 9)
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLulD FLOW ANSWERS -- PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON ANSWER 1.03 (3 0) a. 1. Peripheral rod worth will increase (0.5) Xenon concentration will be highest.in the (0. 75) center of the core where the highest flux previously existed This will suppress the flux in the core center (0.75) thus increasing the flux near the peripheral rods , therefore increasing thei r worth.
11. More than one half the value at 100% (0.5) b. FALSE (0,5) REFERENCES LOT 1490 Control Rod Worth, LO 4, pg. 4 LOT 1510 Xenon , LO 3,4, pg. 5,' T-LOT-1510-4 K/A REACTOR THEORY 292006 Fission Product Poisons Kl.05 State the effect on reactor operations of equilibrium Xenon (2.9) Kl.06 State the effect on reactor operations of Maneuvering Xenon (2. 7) K/A REACTOR THEORY 292005 Control Rods Kl.09 Explain direction of change in the magnitude of CRW for a change in Xenon (2.5)
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I l i l 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, f HEAT TRANSFER AND FLUID FLOW ANSWERS -- PEACHBOTTOM 2t,3 - 87/10/05 - G. E. ROBINSON
.
ANSWER 1.04 (2.0) f (0.75) a. K-effective is less than one but not as low as i t was when all rods were in. i the subcritical multiplication factor is higher (0. 75) Therefore, Th us (1/1-K-effective) for the one rod inserted case.
the countrate is higher.o v /en fle,,.d r/wkm cbr 4ces h'"f wscs whok 6 4 ine wos ed Ass k . j P W m r L L t.
(0. 5) b. TRUE i
REFERENCES LOT - 0970 Subcritical Multiplication, LO 2,3, pgs. 5, 8 K/A REACTOR THEORY 292008 Reactor Operation Physics Kl.03 Describe count rate and period response which should f be observed for rod withdrawal during approach to cri ti cali ty (4.1)
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I 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, ; HEAT TRANSFER AND FLUID FLOW ANSWERS -- PEACHBOTTOM 2&3 - 87/10/05 - G. E. ROBINSON i l ANSWER 1.05 (2.0)
(0. 5) '
a. Wi thout adequate NPSH the pump would cavi tate (vapor bubbles would form at the inlet and collapse within the pump)
(0.5)
b. 1. Increases (0 5) 11. Increases (0. 5) lii. decreases
)
I l l REFERENCES PBABS Fluid Flow Handout, pg. 4-4 LOT 1290 Pumps, LO 4 K/A COMPONENTS 291004 Pumps Kl.06 Need for NPSH (3.3) Kl.14 Relation between flow and suction head (2.5)
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PEACHBOTTOM 2&3 - 87/10/05 - G. E. ROBINSON ANSWER 1.06 (2.0) j i a. Critical power is deff ned as that bundle power which (0,5) causes critical quali ty to exist at some point in the j bundle or causes transi tion boiling to occur wi thin the a bundle (ei ther is acceptable) i
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b. Critical power increases as subcooling increases. (0 5) A greater enthalpy rise is required to bring the coolant (1.0) ) to saturation therefore the power required to achieve q OTB is greater, j i l l i REFERENCE LOT 1370 GEXL and Cri tical Power, L01, pg. 3 LOT 1360 Transition Boiling and ATLMS Testing LO 3, 4, pgs. 3 & 4 , K/A THERMO: 293009 Core Thernal Limi ts K1.17 Oefine Critical Power (3 3) K1.22 Describe ths effects of subcooling on critical power (2.9) l l l
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1. PRINCloLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON ANSWER 1.07 (1.5) ,
l P(t) = P(0) exp(t/ tau); in 10 = 300 sec/ tau (1,0) tau = 300 sec/2.3 = 130.2 seconds period has become longer, therefore heating range has been (0 5) reached REFERENCE LOT 1430 Reactor Period, LO 1, pg. 2 K/A REACTOR THEORY: 292003 Reactor Kinetics Kl.08 Solve problems for power changes and period (2 7) K/A REACTOR THEORY: 292008 Reactor Operational Physics K1,ll Describe reactor power and period response prior to reaching POAH (3.6) Kl.13 Explain characteristics to lock for when POAH j is reached ( 3. 8) l
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l 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS,
. HEAT TRANSFER AND FLUID FLOW l
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ANSWERS -- PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON l
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l ANSWER 1.08 (2.0) l From Steam Tables obtain reactor temperature by fi rst converting PSIG to PSI A then obtaining the saturation temperature Tine P res s ure Tempe ra ture 1: 30 67 psia 300*F (0.75) 2: 30 195 psia 380*FL (0 75)-- Heat up rate is 80F per hour (0.5) Declud pa daI c<eJ:/ fu Stllvre b c.en ve,.t fs b J P ' h -
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REFERENCE LOT 1160 Steam Tables LO 2, pg. 5 !
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K/A THERMO: 293003 Steam Kl.23 Use of saturated steam tables (2.8)
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS, HEAT TRANSFER AND FLulD FLOW . ANSWERS -- PEACHBOTTOM 2s3 - 87/10/05 - G. E. ROBINSON
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ANSWER 1.09 (2.5) a. SDM = l-K-eff; CRI (1-K-effl) = CR2(1-K-eff2) (1.0)
(300 cps) ( 03) = -(800 CPS) SDH2 SDM2 = .01125 = 1.125% (1.0)
b. an increase in moderator temperature causes an (0. 5) increase in SDM. (Due to increase in resonance escape probability) REFERENCE LOT 950 Excess Reactivity and SDM, LO 6, 7, pg. 4 LOT 980 Count Rate Comparison, LO 2, pg. 2 K/A REACTOR THEORY: 292008 _ Reactor Operational . Physics K1.03 Describe count rate which should be observed for rod wi thdrawal during approach' to cri ti cal i ty (4.1) K/A REACTOR THEORY: 292002 Neutron Life Cycle K1.14 Evaluate change in SDM due to changes in plant pa rameters (2.6) Kl.11 Define SDM ( 3. 2)
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_ _ - _ _ _ _ - - _ - _ . . ___- . _ ODYNAMICS,_ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. TI llEAT TRANSFER AND FLUID FLOW _ _
-- PEACHBOTTOM 2&3
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87/10/05 - G. E. ROBINSON ANSWERS (2.0) ANSWER 1.10 (0.5) a.
An adequate scram margin must bei established in order - lation pump.
to prevent a scram upon' restart of the rec rcu h (0 5) A flux spike (of approximately 10%)can be expected f rom t e first opening Jog of the recirculation discharge valve. (1.0)
,;;3h: 'xcc::e. ;
o
,oc vv. ,
m T Tm wum;J mou,c vo i d f i au iv , -; n .! b. T h . in.ree;;e, evo N causing tne. net O0*9 l
;;; : J; crc ::.g p.rsibf lausmy fe es/Iy local *mJ a. Pin-shapin .
exc ess of eccep 44/c. li e h '
- .D re7u/an y 4 fixes in-i REFERENCE :
OT-112, Recirc. Pump Trlp Bases, pg.1-3 i LOT 1860, Power to Flow Map, LO 4, pg. 5, 6 l APE 2 5001 Partial or Complete loss of Forced Core Flow K/A Circulation
'Know of operational impilcations of partial loss AK1/02 of forced circulation (3 and Power / Flow 3)
distribution I f
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERA,TI0ll, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PEACHBOTTOM 2&3 - 87/10/05 - G. E. ROBINSON
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l ' ANSWER 1.11 (1.5) ;
-1 Reactor Pressure will be greater (0. 5)
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i. Flow increases In' the remaining three lines which increases (1.0) l the requi red dri ving head. Thus requi ring a higher reactor pres s ure .
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l y J REFERENCE l LOT 120 Main Steam and Pressure Relief System, pg. 16 K/A 239001 Main and Reheat Steam System A 1.08 Ability to predict changes in Reactor Pressure (3.8)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON
ANSWER 1.12 (1.5) a. .less negative fD939)(o.75) a darr-::: ', :S >cid 'r ::!:r~g:fe : ! es
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b. less negative @NEBE)(o.7s> ' l a inrnor'" rr--+t " re e g ,er : er t 1
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l LOT 1460, Void Coefficient, LO 3, 4, pg. 3, 4, 5 i
'l K/A REACTOR THEORY: 292004 Reactivity Coefficients describe the
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effect on the magnitude of the Vold Coeff. from changes in
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Kl.Il Core Void Fraction (2.5) Kl.13 Core age - (2,1)
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ANSWERS. -- PEACHBOTTOM 2&3 - 87/10/05 . G. E. ROBINSON ANSWER 2.01 (3 0) o a. Low level in CST Tank (0 3), 10,000 gal or 5 ft. 7 inches above bottom of tank (0.2) b. High Water Leve! Lioses M0 '131, not the trip. throttle valve. (0. 5 ) .. Trip Throf tle valve must be reset. (0.5) Closing MO 131 Instead of the trip throttle valve allows ( 0.5) l auto restart at -48 inches.
c. 2200 rpm- (0.5) Any of the following: prevent equi pment . damage, prevent water hammer, (0.5)' or prevent demage-to exhaust check valve
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REFERENCE LOT 380, RCIC, LO 2, 3, pgs. 4, 14, 20 K/A 217000 RCIC , K4 Knowledge of RCIC design features and or interlocks which ! provide for: K 4.02 preventing over filling reactor vessel (3 3) J K 4.04 preventing turbine damage (3 0) K 4.07 alternate supplies of water (3.6)
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_ ' l 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY-' SYSTEMS ANSWERS - -PEACHBOTTOM 263 - 87/10/05
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G. E. ROBINSON
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ANSWER 2.02 (3.0)
.a. N Drive Plston -(0.25) -
0 Exhaust Scram Valve (0.25)L L Index Tube - (0.25) J Collet Fingers . (0. 25) : b. See Enclosed Figure -(2.0)
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REFERENCE LOT 70 CRD Hydraulic System LO 1, 3 pg. 5, T-LOT-0070-2 and 3 K/A 201001 CRD Hydraulic System K4.10 Know. of CRD Hydraulle system design feature for.. controlling rod movement (3.1)
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_ _ _ _ _ _ - _ _ _ _ _. _ - . - . _ - _ _ . _ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ANSWERS -- PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON l ANSWER 2.03 (3 0) a. to prevent precip!tation of the sodlura pentaborate (0.5) i solution b. Selected pump starts (0 3) both squit valves fire (0.3) Isolation of RWCU system (0.3) c. Any four (0.4 each) Solution tank level decreasing injection pump discharge header pressure injection pump run Indicator lights Squib valve loss of continuity lights Maintenance valve position indicator REFERENCE LOT 310, SBLC, LO 7, B,'pg. 5.9 K/A 211000 SBLC
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K4.02 Knowledge of design for keeping sodium.pentaborate in q sol ution - (3.8) K4.08 Knowledge of System initiation (4.2)
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A4 Ability to predict changes' in parameters -l j-A1. 01 Tank level (3.6) A1.02 Explosive valve Indication (3.8)
A1.03 Pump-Discharge press
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(3.6)
A1.04 Valve operations (3.6) A1.10 Lights (3 7)- . i
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS : l ANSWERS -- PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON ANSWER 2.04 (3 0) a. reci rculation valve ( bc/u sa4 rccere, Sl /p{'M *" b M 3* kpot m(0accepkkle) 34) Sue condensate reject control valves (0.33) CRD pump suction valves (0 33) b. TRUE (0 5) c. Reactor recirculation pumps run back to 60% speed. (0.75) A 90% maximum speed signal is inserted into the (0.75) feedwater control system.
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[2 & % W.\ 1 REFERENCE LOT 520, Conder. sate system, LO 6, 11, pg. 16, 17 K/A 256000 Reactor Condensate System A 2.05 Ability to predict impact of inadequate system flow (2.9)
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, _ _ _ _ _ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS l ANSWERS -- PEACHBOTTOM 2c3 - 87/10/05 - G. E. ROBINSON
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ANSWER 2.05 (2.5) l a. 1. not interrupted (0.25)
l 11. not interrupted (0.25) 111. not interrupted (0.25) iv. AMinterrupted 3 (0.25) <
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b. i. FALSE (0.25) It is set at 3 x NFPB (0.5) 11. FALSE (0.25) any three steam lines (0.5)
l REFERENCE l LOT 120 Main Steam and Pressure Relief System, LO 2,15, pgs. 4,15 K/A 239001 Main and Reheat Steam Systems I K1 Knowledge of physical connection between Main Steam and: )
Kl.08 Condenser Ai r Rereval System (2 9)~ Kl.09 Steam Seal (2.7) 1 Kl.18 HPCI (3 5) Kl.22 Feedwater System (3.1)
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K4.01 Knowledge of Main Steam System interlocks which provide for auto isolation of steam lines (3.8) l
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 3 ANSWERS -- PEACHBOTTOM 2&3
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87/10/05 - G. E. ROBINSON
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ANSWER 2.06 (3.0) a. Manual start from control room wi thout QUICK START (0 5)' requires the prelube pump to start ' and delays . the diesel generator start for 3 minutes (0. 5) b. Any four (0 5 each) W Engine Overspeed Trip (Electrical or Mechanical) Generator Phase Dif ferential Overcurrent Neutral Overcurrent Manual Cardox Injection e_____ -- _;_ REFERENCE LOT 670, Diesel Generator and Auxiliaries, LO 3, 4, pgs. 32, 39 K/A 264000 Emergency Diesels K4.02 Knowledge of emergency generator interlocks for emergency generator trips (LOCA) (4.0) Kl.06 Knaaledge of cause-effect for starting system (3.2)
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY' SYSTEMS ANSWERS -- PEACHBOTTOM 2&3
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87/10/05 - G. E. ROBINSON ANSWER 2.07 (2.5) DW area' cooling units (0. 33) a.
Recirculation pump motor coolers (0 34) Drywell equipment drain sump cooler (0. 33) b. The motor operated transfer (swap) valves and the air operated Isolation valves will open to isolate DW chilled water and line up RBCCW to both the "A" and "B" headers (1.0) The non-essential loads would be isolated by an air operated valve within the-RBCCW system '(0.5) REFERENCE LOT 150, DCVS, LO 2,' 3, 5, pg. 3, 9 K/A 223001 - Prim. Cont. and Aux.
K1.04 Knowledge of cause-effect wi th Drywell Equipment Float l Drain System (3 1) K6.01 Knaaledge that a loss has on Drywell' Cooling (3.6)
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_ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ _ - . 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ANSWERS -- PEACHBOTTOM-2t,3 - 87/10/05 - G. E. ROBINSON
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ANSWER 2.08 (2.5)-
)/G g ,.a-rr ,v4 n a r> h<b a. JKI they will ocr(suto ste% -{G.5) N #N
- %. . ivete. , auto pump st:rt r ::t pt9 1 utter p*f N.5'M *
fer :: h pw.op must b puched. s
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11. FALSE (0.5) iii. TRIJE (0.5) REFERENCE LOT 305, Core Spray System, LO-3, 4, pg. 4,'9, 10 K/A 209001 L w Pressure Core Spray System K1.08 Knowledge of physical connections with A.C. electrical power (3.2) Kl.02 Knowledge of physical connections with CST. (3 1) K4.01 Knowledge of design features for prevention of overpressurization (3.2) K4.08 Knowledge of interlocks for auto system i ni tiation _( 3 8) . l
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS , ANSWERS -- PEACHBOTTOM 263 - 87/10/05 - G. E. ROBINSON
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ANSWER 2.09 (2.5) a. Part of the flow f rom RHR Loop "A" may be diverted (1.0) to a spray nozzle in UNIT 2, while for UNIT 3 it is LOOP "B".
b. To prevent diverting coolant to the torus (0. 5) c. Inboard S/D cooling suction isolation valve (0.25) Outboard S/D cooling suction isolation valve (0.25) l Rx vessel head spray isolation valves (0.25) LPCI Shutdown Cooling injection valve (M0 25) (0.25) REFERENCE LOT 370, RHR, LO 2, 3, 6, pg. 7,18 K/A 205000 Shutdown Cooling K4.05 Knowledge of design which provides reactor cooldown rate (3 6) K6. 04 Knowledge of effect malfunction will have on Reactor Water level (3.6) A3.01 Ability to monitor auto operation of valve operation (3 2) System Gen. 10. Ability to explain all system precautions (3 2)
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3 INSTRUMENTS AND CONTROLS ANSWERS -- PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON ANSWER 3:01 (1.0) SRM's indicate below 100 cps and not full inserted (0 5) count rate less than 3 CPM (0,5)
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REFERENCE LOT 240, SRM System, LO 2, 4, pg. 4, 11 K/A 215004 SRM's A2.05 Ability to predict impact of faulty operation of detector on SRM System (3 3) K4.01 Knowledge of rod withdrawal blocks (3.7)
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-ANSWER 3.02 (3 0) .
l L a. Can be completely withdrawn (0,5) b. only the rod at position 6 (0 5) once a rod in a latched group is moved out one notch, (0,5) l it removes' the permissive to move any other rods in the group in.
c. LPSP is that point at which RSCS interlocks are bypassed (0.5) or inittsted set at about 21% power '(0. 5) sensed by Main Turbine 1st stage shell pressure (0.5)
(NOTE: Actual set point is higher,102 psig; 85 psig - 21%)
REFERENCE LOT 100 RSCS, LO 2,3, pg. 6, 7, 8 l K/A 201004, RSCS - l-Kl.01 Knowledge of cause-effect with manual control system (3.2) K4.02 Knowledge of interlocks which provide insert rod blocks (3.1) K5.03 Knowledge of operational appiteations of group notch control 1Imits (3.3) "
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j 3. INSTRUMENTS AND CONTROLS ANSWERS -- PEACHBOTTOM 2r,3 - 87/10/05 - G. E. ROBINSON
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ANSWER 3 03 (3 0) ,
-1 a. l. Inhibit (Discharge valve ~must be closed) .(0.25)
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II. (Would not inhibit) (0.25) 111. (Would not i nhibi t) (0.25) . iv, inhibit (Gates must be wi thdrawn) (0.25) b. Balance actual . generator speed with speed demand (0,5) i Press Reset Button (0 5) c. To insure MG does not receive a scoop tube lockout (1.0)
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REFERENCE LOT 30, LO 5, pgs. 20, 21 LOT 40, LO 7, 9, pgs. 8,10 " K/A 202001 Reci rculation System A4.01 Ability te manually operate recirculation pumps (3.7) i K4.10 Knowledge of pump start permissives (3.3)
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- _ _ - _ - _ _ _ _ - _ _ _ l t I 3 INSTRUMENTS AND CONTROLS ANSWERS -- PEACHBOTTOM 2&3 - 87/10/05 - G. E. ROBINSON ANSWER 3.04 (3.0) , a. 172 inches (or'178 inches) (0 5) b. higher (0.25) An equalizing valve open causes the. reference and variable - (0. 75) legs to equalize. The D-P cell senses zero D-P and - yields a high level. signal.
c. higher (0.25) The water temperature in the variable leg is less than (0.75) the actual water temperature. Therefore the density of the water in .the variable leg is greater than the actual water density.
_ d. FALSE P'M M h l REFERENCE l LOT 50, Reactor Vessel Instrumentation, LO 6, pg. 7, 8,17,18 K/A 216000 Nuclear Reactor Instrumentation K5.01 Knowledge of concepts of vessel level' Instrumentation (3 1) A2.01 Ability to predict impact of detector equalizing . valve leaks (2 9) A2.08 Ability to predict impact of elevated con'tainment temp. (3.0)
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3 INSTRUMENTS AND CONTROLS l ANSWERS -- PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON ANSWER 3 05 (3.0) a. At least one RHR pump or two Core Spray pumps operating; (0.5) 105 second tiner-tines out (0. 5) b. Allow reactor pressure to decay below 50 psig; (0.5) Depress the "A" and "B" Timer Reset pushbuttons (0. 5)
(to deenergize logic relays and break seal-in)
Shutdown the RHR or Core Spray Pumps (0.5)
, indidat Place Keylock swi tches "A" and "B"
\ l in . :' te (0.5) REFERENCE LOT 330 LO 2, 3, pg. 8 K/A 218000 ADS K4.03 Knowledge of ADS logic control (3.8)
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A2.06 Abili ty to predict impact of ADS initiation signal present (4.2)
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3 INSTRUMENTS AND CONTROLS j ANSWERS -- PEACHBOTTOM 253 - 87/10/05 '- G. E. ROBINSON
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ANSWER 3.06 (2.5) - l a. 1. - Scram (0.5) . :I
' k b. iv. - no action (0. 5) - c. 11. - half scram . (0. 5) - d. iii. - Rod Block (0 5).
e. . ins - N h a lE- S C#"A ~(0.5) REFERENCE , LOT 250 IRM, Lo 5, pg. 10 LOT 270 APRM, LO 2, pg. 7, 8 LOT 300 RPS , LO 8, pg. 11, 16 K/A 215003, IRM's K4.02 Knowledge of interlocks for reactor scram . (4.0) K/A 212000 RPS K1.01 Knowledge of cause-effect with nuclear inst. (3 7) Kl.10 Knowledge of cause-effect wi th main turbine (3. 2) ' K6.02 Knowledge of effect of loss. of. nuclear inst. (3.7)
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3 INSTRUMENTS AND CONTROLS ! ANSWERS -- PEACHBOTTOM 263 - 87/10/05 - G. E. ROBINSON ANSWER 3.07 (2.0) a. Fresh air supply fans trip (0.25) ! AC fans trip (0.25) Return fans trip (0.25) l Emergency vent fans start (0.25) . b. (i no isolation) (0 33)
(il no isolation) (0.33)
lii isolation (0 34) l REFERENCE LOT 720 Process Radiation Monitoring, pg. 11 K/A 290003 Control Room HVAC ! i Kl.01 Knowledge of cause-effect of Rad. nonitors (3.4) K4.01 Knowledge of design which provide system reconfiguration (3.1) A3 01 Abillty to monitor automatic operation of
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reconfiguration (3.3)
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3. INSTRUMENTS AND CONTROLS ANSWERS -- PEACHBOTTOM 263 - 87/10/05 - G. E. ROBINSON ANSWER 3.08 (3.0)- l a. Load Rejection, initiated-at 40% mismatch between HP (1.0) exhaust pressure and generator amps Loss of Stator Cooling,' initiated upon a loss of stator (1.0)' cooling c, , 9,9 T p . ev L.= P've sssvg b. Load Rejection - Runs back to zero load (0.5) Loss of Stator Cooling - Runs back to 23% (0.5) REFERENCE LOT 590, EHC Logic LO 5, 6, pg. 37 VJA 241000 Reactor Turbine Press Regulating System K4.07' Knowledge of design features pro,ilding Generator Runback (3.2)- K6.16 Knowledge of effect of loss of stator water cooling on system (2 9) A2.09 Ability to predict impact of Loss of Generator Load (3.4)
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A2.10 Ability to predict impact of Loss of Stater - Water Cooling (3.1)
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i 3 INSTRUMENTS AND CONTROLS ANSWERS -- PEACHBOTTOM 263 - 87/10/05 - G. E. ROBINSON ANSWER 3.09 (1.5) NO (0.5) This switch is only active in the shutdown or (1.0) refuel swi tch pos i tion i REFERENCE LOT 300, RPS, LO 8, T-LOT-0300-14 l K/A 201001 CRD Hydraulic System l K4.ll Knowledge of the effect of malfunction to protect J against filling the SDV during non-scram condition (3.6) l
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I 3 INSTRUMENTS AND CONTROLS l l l ANSWERS -- PEACHBOTTOM 27,3 - 87/10/05 - G. E. ROBINSON l
# 1 ANSWER 3 10 (1.5)
A Reactor power wi11 decrease (0.5) l As scram air header pressure decreases control rods will (1.0) begin drif ting in at random l l
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REFERENCE ON 108 l K/A 201001 CRD Hydraulic System K6.03 Knculedge of effect of loss 'of plant ai r systems (3 0) i l
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3 INSTRUMENTS AND CONTROLS ANSWERS -- PEACHBOTTOM 2&3 - 87/10/05 - G. E. ROB illSON ANSWER 3.11 (1.5) An IRM UPSCALE TRIP (Rod Block) would occur (0.75) IRM would read 110 on the 0-125 scale (0.75) REFERENCE LOT 250, IRM System LO 5, pg. 5, 9 K/A Components: 291002 Sensors , Detectors K1.20 Neutron Monttoring Units (3.2)
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4. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
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ANSWERS - PEACHBOTTOM 253 - - . 87/10/05 - G. E. ROBINSON
ANSWER.4.01 (3 0) I a. 11 (entry condi tion to T-102) -(0.5) ) b. 111 (entry condition to both T-101 and T-102) (0.5) c. Iv (not .an entry condition) ' (0. 5) d. Iv (not an entry condition) (0,5): e. 1 (entry condition to T-101) .(0,5) f. 1 (entry condition to T-101) (0.5) REFERENCE LOT 1560 INTRODUCTION TO TRANSIENT RESPONSE, LO 9, pg. 6 K/A APE 295010 High Drywell Pressure System Gen. II, Recognize entry. condition (4.2) l
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4. PROCEDURE - NORHAL. ASNORMAL, EMERGENCY AND RADIO _0GICAL CONTROL-ANSWERS - PEACHBOTTOM 2&3 - 87/10/05 - G. E.-ROBINSON
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ANSWER 4.02- (3 0) a. with Form 4 2500 mrem /qtr. (0. 25) w/o Form 4 1000 mrem /qtr. (0.25) b. 1. between 100 mr/hr and 1000 mr/hr . (0. 5) tL. All ' dos imetry devices (0.5) ' RWP -(0.5)
<d . . (Any one of. the followingL is acceptable) '
(0.5)
Constantly indicating' dose rate inst.
Integrating alarming dosimeter HP Escort j i c. I 75 rem (0.25) !
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11 25 rem (0.25)-
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REFERENCE LOT 1730. Rad. Exposure Limits, LO 1, 2,15, pgs. 4, 6, 8 K/A PW Gen. 294001 Kl.04 Knowledge of facility . radcon: '(3.3) K1.05 Knavledge of fact 11 ty req. for controlling access. (3.2) _
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I 4. PROCEDURE - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS - PEACHBOTTOM 253 - 87/10/05 - G.-E. ROBINSON ANSWER 4.03 (2.0) a. Rod Motion restricted to one notch in either direction (1.0) b. Dri ve fl ow (0.33) Drive Pressure (0.33) settle, drive in, and drive out lights (0.34)
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REFERENCE , LOT 1550, ON Procedures LO 2, pg. 3, 9 K/A System 2010013 CR and Drive Mechanises A2.01 Ability to predict impact of stuck rod (3.4) System Gen.10 Ability to explain and apply system-precautions (3.2)
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4. PROCEDURE - NORMAL, ABNORMAL EMERGENCY AND RADIOLOGICAL CONTROL l ANSWERS - PEACHBOTTOM 2&3 - 87/10/05 G. E. ROBINSONL
. ANSWER 4.04 (3 0)
a. If the fi re brigade cannot extinguish the. fire and (1.0): . I off-si te ass is tance is requi red b. If the fi re jeopardizes normal ' plant shutdown or (1.'0) . ) ECCS capability c. continuous recirculation flow will prevent stratification . ( 1. 0 ) ~
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I REFERENCE LOT 1550 ON Procedures, LO 2, pg. 19, 20 i K/A System 286000 Fire Protection System- j
Sys. Gen. 14 Abili ty to perform wi thout' reference-1.0. octions (3.8) i K/A System 202002 Reci rculation Flow Control System '{ Sys. Gen. 10 Ability to. explain system precautions (3 3)
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#EACHBOTTOM 2&3 -
87/10/05 - G. E. ROBINSON ANSWER 4.05 (1.5)
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a. Used on nuclear safety-related indicators on controls (1,0) for which operabili ty testi ng i s requi red and must be deferred due to plant condi t ions b. TRUE (0. 5) -
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REFERENCE LOT 1570 Admin. Procedures LO 3g Procedure A-26A pg. 15, 16 K/A 294001 Plant Vide Generics i Kl.02 Knowledge of tagging procedures (3 9) I l
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4. PROCEDURE - NORMAL, ABNORMAL, EMERGENC AND RADIOLOGICAL CONTROL ANSWERS - PEACHBOTTOM 2&3 - 87/10/05 -~ G. E.' ROBINSON lj i ANSWER 4.06 (1.5) . u Spurious initiations. (0.75) continuel operation beyond automatic trip points (0.75) i
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i l REFERENCE LOT 1560 intorduction' to Transient Response. Proc. , LO 3 Basis for Trip Procedure Notes, pg. 17 K/A EPG 295031 Reactor Low Water Level Sys. Gen. 7 Ability to explain precautions .(3.7)
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b. decrease in core plate DP (0.5) c. increase in defective : loop drive flow (0.5) . d. decrease in DP on jet pump sharing riser (0,5) REFERENCES LOT 1550, off Normal Proc. , LO 1 ON-100 Failure of a Jet Pump pg. 7 K/A System 202002 Reci rc. Flow . Control System K3 Knowledge of effect malfunction has on K3.01 Core Flow (3.5) K3.02 Reactor Power- (4.0) Sy. Gen. 15 Ability to recognize abnormal Indications
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which are entry conditions to abnormal proc. (3.8)
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ANSWERS - PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON ANSWER 4.08 (2.5) a. Place both drywell inst. N2 valve bypass switches in the BYPASS position; (0.5) Place drywell inst. N2 valves A 0-2(3) 969 A&B in the CLOSE position; (0 5) Place drywell inst. N2 valves A 0-2(3) 969 A&B in the AUT0/ (0.5) OPEN position b. Drywell instrument air is essential for operator control of the E, H and L relief valves at the ESP (1.0) REFERENCE
SE-1 Plant Shutdown from ESP-Basis, pg. 2 l ) K/A APE 295016 Control Room Abandonment l AK 2.01 Knowledge of Interrelationship with RSP (4.4) Sys. Gen. 10 Ability to perform 10A (3.8) i i e
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ANSWERS - PEACHBOTTOM 2&3 - 87/10/05 - G. E.'R0dlNSON l' ANSWER 4.09 (2.0) a. Verify that all scran discharge vent and. drain valves are open -(0.5) i i If a scram occurs, enter procedure T-100 (0,5)~ b. Rod block 25 gal. (0 5) Scram 50 gal '(0.5) l
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REFERENCE
l LOT 1540 Operation Trans. Proc. L . O. 1,2 OT-105 - SDV High Level pg. 1, 2 K/A System CRD Hydraulle System l. K4.ll Knowledge of interlock for protection against filling l the 50V during non-SCRAM conditions (3.6)' Sys. Gen. 14 Ability to perform 10A (3 7)
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L ANSWER 4.10 (2.0) a. Group til isolation will occur (0.5) SBGT system will s tart (0.5) b.
Evacuate the reactor buildin9 (0 5) Vert fy group iii isolatlon (0,5) REFERENCE E-6, High-Radiation-Reactor Building Exhaust, pg.1 . K/A System 288000 Plant Ventilation System K4.01 Knowledge of interlocks which provide for auto initlation of SBGT ' (3 7)
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i 4. PROCEDURE - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTRO1 ANSWERS - PEACHBOTTOM 253 - 87/10/05 - G. E. ROBINSON ANSWER 4.11 (2.5) _ a. used if a timer card malfunction occurs ov whew M5M (0. 5) b. This switch bypasses the timer card and therefore the (1.0) settle function of the P.MCS. Settling occurs due to seal leakage through the CRD. This damages the seals.
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av 6,gosues 9 roup ne /c & M / mude c. EHC is the primary means of controlling pressure and hence (1.0) h ea t- up E
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REFERENCE LOT 1530, GP-2, Normal Plant Startup pgs. 10, 11 K/A 201002 Reactor Manual Control System L.06 Knowledge of RMCS design features which provide
" Emergency IN' rod insertion (3.5)
K/A 241000 Reactor / Turbine Press. Regulating System Sys. Gen: 4 Knowledge of system function (3.4)
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Q fj]gj ' t1aa mrt;TE7 a 4.n j U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _ PEACH _@OTTgM_gh3,_______ Stare - 06(0 REACTOR TYPE: _gWR-GE4_________________ 66 -1430 DATE ADMINISTERED: _gZf1@l@D________________ EXAMINER: _KQLgNAU@K11_L.__________ CANDIDATE: _________________________ IN@lSUCIlgN@_JQ_C@NDID@lE Use separate paper for the answers. Write answers on one side onl y.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grace 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 j the examination starts.
l l l ! % OF . ' CATEGORY % OF CANDIDATE'S CATEGORY _ _ V_ A_ L_ U_ E_ _ _ _T O_ _T_ A >
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_ A_ L_ U_ E _ _ - _ _ _ _ _ - - _ _ _ _ - _ C_ A_ T_ E G_ O R_ Y_ _ _ _ _ _ --- _ _ _ _ -
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1 y dsn4n ' ' 21_____ _=ereg ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT 1 A47 OPERATION, FLUIDS, AND ; 25,7 THERMODYNAMICS j 27.7 ! -et#447 ! _291@@__ _AEssa ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, j g g, f AND INSTRUMENTATION i A>FdbY+<!* M19 ; _2Et@@__ _A5 55 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, l , ' EMERGENCY AND RADIOLOGICAL 23,7 CONTROL ! 23,00 7 3 p,7 _;, , w,, _, _SEsss__ _SE ?s ___________ ________ 8. ADMINISTRATIVE PROCEDUREF; 4 CONDITIONS, AND LIMITATILWS 6184-17 INDI??'__ _________-- _______-
% Totals Final Grade a tyrh7 euuHx t.0G kktks- W uun.
All work done on this examination is my own. I have neither given i nor received aid.
& n-w oexs- S,oga & , ---_____-____ -____-_____-------___
Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examinati on means an automatic denial of your application and could result in more severe penalties.
2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anycne outside the examination room to avoid even the appearance or possibility of cheating.
3. Use black ink or dark pencil only to facilitate legible reproductions.
4 Print your name in the blank provided on the cover sheet of the ex ami nat i on.
5. Fill in the date on the cover sheet of the examination (i f necessary).
6. Use only the paper provided for answers. . 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
9. Consecuti vel y number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly on gne side of the paper, and write "Last Page" on the last answer sheet.
9. Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. 14 parts of the examination are not clear as to intent, ask questions of the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the ex ami nati on. This must be done after the examination has been completed.
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I l l l 18. When you complete your examination, you shall: a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Ex am aids - figures, tables, etc.
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(3) Answer pages including figures which are part of the answer.
b. Turn in your copy of the examination and all pages used to answer the ex ami nati on questions.
c. Turn in all scrap paper and the bal ance of the paper that you did not use for answering the questions.
d. Leave the examinati on area, as defined by the examiner. .lf after l eavi ng , you are found in this area while the examination is still in progress, your license may be denied or revoked.
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QUESTION 5.01 (3.00) a. In GP-2 Section III, "Heatup to Rated Tem, frature and Pressure," the operator is instructed to maintain a specified heatup rate.
If the heatup rate should exceed this given limit, the operator is instructed to take one of two actions. List the two (2) actions and briefly explain HOW each reduces the heatup rate. (1.5) b. During a given heatup at PBAPS, the following information was obtained: TIME RPV PRESSURE 0130 52 psig 0200 102 psig 0230 180 psi g From 0130 to 0230, what was the heatup rate? Show all work. (1.5) GUESTION 5.02 (1.50) g mod WOU0 to(5/17 If the current SRM reading is 150 cps during startup p r . c. ;st
,, gg e' i ' i c m l i + -n what will the reading be in one minute if a dood constant period of 70 seconds is maintained? (1.5)
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5 __IHEg6Y_QE_gUCLE@@_EQWE6_EL@NI_05E5@lIQNz_[LUIDS _@ND t PAGE 3 IHE5dgDYU@d1CS I OUESTION 5.03 (1.50)
In the event of a major LOCA at PBAPS. the f ollowing even t.s j can occur. For each, choose the words that CORRECTLY complete the sentence describing the effects on RPV water level indications.
a. Rapid decreases in RPV pressure will cause the YARWAYS to Indicate (HIGHER THAN ACTUAL / LOWER THAN ACTUAL / THE SAME AS ACTUAL) water level. (0.5) b. Rap 2d decreases 2n RPV pressure will cause the GEMACS to indicate (HIGHER THAN ACTUAL / LOWER THAN ACTUAL / THE SAME AS ACTUAL) water level. (0.5) c. LPCI injection flow through recirc will cause the YARWAYS to read (HIGHER THAN ACTUAL / LOWER THAN ACTUAL / THE SAME AS ACTUAL) water level. (0.5) CUESTION 5.04 (3.00) STATE whether each of the following changes would INCREASE, DECREASE, or HAVE NO EFFECT on the heat transfer rate in the RBCCW heat exchangers.
Justify your answer. Assume all other parameters are held constant and list any assumptions that you make, a. Heat Exchanger Tube failure (1.0) b. A decrease in Service Water system flow (1.0) c. An increase in Service Water temperature (1.0) ,
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_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ E _.ldEOBy_DE_ygC6E@B_EOWEB_E6@gI_g[gg@llggt_E6 gip @z_@yg PAGE 4 ISE6Dggyy@[lC@ QUESTION 5.05 (2.00) Step 10 of Section II of GP-2, " Normal P1 it Startup,' states:
"Early rods may be withdrawn in the notch override mode. ...
Rods in RWM Groups 3 and 4 shall be wi t hdr e-in using the Bank Position Withdraw Sequence until the reactor is cri ti cal . " Banked rod sequence means that rods are pulled 00-02-04 as a group, then to 04-06-08, and so on.
a. HOW are individual rod worths limited by these pr ocedur al res-tractions? (1.0) b. WHY is it necessary to limit individual rod worths at this point in the startup? (1.0) QUESTION 5.06 (3.00) a. If ONE recirc pump trips whi l e at power, DT-ll2 requires the operator to reduce power by driving in deep rods. Assume that a scram has been avoided and a restart of the pump is planned.
WHY does the procedure require the operator to CONTINUE to insert rods prior to pump restart? (1.5) b. What is the definition of a deep rod? (0.5) c. Explain the problem that could occur if shallow rods were inserted instead of deep rods. (1.0)
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QUESTION 5.07 (3.00) a. in e t <,;c. uses e m c ess-Prydrogerr-+com-- r eac t er roul ant) dysome ns i t i mm havn on the m-y . u vii u ti ij. r a t i nn o+ t,t2e G nar+ ne w -+n;; anLT 4 // 9-o ,ggc/ Jeg %dae4j 2pc14%. (l.0) b. What chemical parameter serves as an indication of the condition of the +uel c l ad di rig ? (i e. , this parameter would increase in the event of minor fuel cladding defects.) (0.5) j c. PBAPS TS 4.6.B contains the following required survei l l ance intervals for r eac t or coolant c hemi str y sampling for chlorides and conductivity.
REACTOR CDOLANT
- PLANT CONDITIONS - - CHEMISTRY SAMPLING INTERVAL -
Steaming rates above Every four (4) days 100,000 lb/hr Startups and steamir.g Every day rates below 100,000 lb/hr EXPLAIN WHY TECH SPECS require more frequent reactor coolant chemistry sampling during startups and when steaming rates are below 100,000 lb/hr. (1.5)
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51__1dE931_Q{_yyCh[@@_[QWE@_[(@$1_Q[[@@llQNz,[(gip @t_@ND PAGE 6 ISE50QQyN@DlC@ QUESTION 5.08 (3.00) A caution statement in GP-2 warns the ope- stor of hi gh rod notch worths experienced during startups at other facilities.
Two separate cases of high rod notch worths are mentioned: 1. " Withdrawal of the first rods in a group, especially groups 3 and 4, will usually exhibit high rod notcn worths."
2. "Also during startups wi th xenon peak conditions and no voids in the core, withdrawal of all rods and especially edge rods may exhibit high rod notch worths."
ANSWER THE FOLLOWING OUESTIONS CONCERNING THESE TWO CASES.
a. WHY do the FIRST rods pul l ed in a group display higher rod notch worths than the LAST rods pulled in that group? (1.0) b. WHEN are " peak xenon" conditions (as mentioned in "2" above) expected? (1.0) c. WHAT causes the edge rods discussed in "2" to have higher rod notch worths? (1.0)' DUESTION 5.09 (2.00) Answer the following questions concerning post-LOCA conditions at PBAPS.
a. If the core was not adequat el y c ool ed , increased hydrogen concentrations will exist in containment. Explain an expected hydrogen producing reaction in containment during post-LOCA conditions. (1.0) b. Even if the core is covered, is it possi bl e to obtain a significant steam-z i rconi um reacti on rate? EXPLAIN. (1.0)
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1. Two inboard MSIVs inadvertently close, causing an increase in flow in the two Main Steam Lines that remain open.
2. A malfunction wi thi n the Feedwater Control System causes reactor water level to reach 47".
b. A turbine trip occurring at full power imposes a more severe reactor pressure transient if it occurs at the end of a fuel cycle (EOC) as compared to a turbine trip occurring at the beginning of cycle (BOC).
Give two (2) reasons for this difference. (1.0)
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i et__PL@NI_@Y@IEM@_QEgl@Nt_CgNI@gLt_@yp_lN@I@UdENI@IlgN PAGE B ; i l l QUESTION 6.01 (2.00)
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a. With the mode switch in STARTUP, and IRM "C" reading 12 on j Range 5, what TRIPS, IF ANY, would occur if IRM " C was down 4 ranged to Range a? EXPLAIN. (l 0) b. If the IRMs are indicating 20 on Range 6 and an operator down ranged to Range 5, WHAT TRIPS, IF ANY, would occur? EXPLAIN. (1.0) l
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OUESTION 6.02 (2.50) a. Briefly explain how RCIC system component response di f f er s between a high reactor water level trip and a RCIC turbine trip. Indicate the REASON for this difference. (1.5) b. Assume that RCIC intiated on a valid low RPV water level signal. An SRV then inadvertently opens and reactor pressure drops to 900 psig.
Briefly explain what happens to the RCIC system flow rate as a result of the reactor pressure decrease. Assume NO (1.0) operator action.
QUESTION 6.03 (1.50) LK OMS to(5/t7 WMD
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a. What TWO (2) automatic actions will occur sotm myaLmu1if one condensate pump trips while the reactor is at full power? (1.0) b. TRUE or FALSE: Current administrative limits at PBAPS allow the plant to, operate at approximately 80% power with one of the three feedwater heater strings valved out of' service. (0.5)
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f 6.__PL@NI_@y@lEMS_QE@lgN1_CgyIBgL3 _@yp_IN@IBUMENI@IlgN PAGE 9 QUESTION e.04 (2.00) An inadvertent closure et the JNBOARD MSl') on Main Steam Line
"A" occurs while the reactor is operating at 70% power.
For the following equipment and systems listed below, state whether or not their steam supply would be interrupted.
a. High Pressure Coolant Injection (HPCI) turbine (0.5) b. Reactor Feed Pump turbines (0.5) c. Main l ur bi ne Steam Seals (0.5) d. Steam Jet Air Ejectors (0.5) GUESTION 6.05 (2.00) a. When starting an Emergency Diesel Generator from the (1.0) control room, briefly state the difference in system e qui pmen t response between a manual start using the manual start switch and a manual start using the GUICK START pushbutton.
l b. Procedure S.8.4.A, " Manual Start of Di esel s , " states l that "at least one minute must elapse between a diesel shutdown or trip and a diesel restart."
l Explain the response of the emergency diesel generator l if an operator attempted to restart it prior to the one minute time lapse. (1.0)
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QUESTION 6.06 (2.00) j a. The reactor pressure is 350 psig and the core spray system l automatically initiated because of low water level'in the I reactor vessel. When water level was recovered, the operator placed all four core spray pump handswitches to STOP.
If the water level subsequently falls below the core spray initiation setpoint, will the Core Spray pumps aut omat i c al l y restart? Briefly explain why or-why not. Assume no operator action. (1.0) I b. 1. TRUE or FALSE? It is possible for an operator to manually provide suction to the Core Spray system tram the CST from the main control room. (0.5) 2. TRUE or FALSE? It is possible to manually open the outboard motor operated core spray injection valve from the control room if the inboard valve is closed AND reactor pressure is above 450 psig. (0.5)
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e __EL@NI_5YSIEUS_DESIGNz_CgNIBgL 2 _gND_INSIBudENIGIlON PAGE 11 QUESTICN 6.07 (2.50) For each of the lettered conditions l i s t e .d below. indicate which of the following will AUTOMATICALLY occur. IF more than one action automatically occurs, state only the most severe action.
(ie., A half scram is more severe than a rod block.)
Assume no operator actions.
ASSUME THAT THESE EVENTS OCCUR ON PBAPS UNIT 2.
No Action Rod Block Half Scram, Scram a. During startup and on range 3 ot the IRMs, IRMs 'A' and "H" lose voltage. (0.5) b. At 20% power, all turbine stop valves close. (0.5) c. At 10% power, APRM "C" loses voltage. (0.5) d. While operating at 60% reci rcul ati on flow, reactor (0.5) power is 85%. e. At 60% power, one APRM flow converter fails downscale. (0.5) i QUESTION 6.08 (2.00) { During a PBAPS Unit 2 startup, the Reactor Water Cleanup (RWCU) System is in use in the blowdown mode to help control reactor water level. A reactor operator makes an error which causes ) blowdown flow to increase. I l i After some time, an increase in recirc pump seal temperature is noted. HOW did the increase in blowdown flow cause the in- l crease in retirc pump seal temperature? (2.0) i
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a. During operation at rated conditions it 15 d i sc over'ed that the scram discharge volume high water level bypass switch has been in the BYPASSED posi ti on since the mode switch was placed in ] STARTUP.
Could this error have prevented a VALID high instrument vol ume scram from occurring? Explain. (1.0) b. List ALL automatic attacus which occur on increasing level in the Scram Discharge Instrument Volume. Include setpoints. (1.0) DUESTION 6.10 (3.00) a. Main Steam Line radiation monitor "B" fails upscale.
(The instrument indicates past the Hi-Hi setpoi nt. ) STATE ALL AUTOMATIC ACTIONS and ALARMS that will occur. (1.5) b. While attempting to repair Main Steam Line radiation monitor
"A" which has failed upscale, an ILC technician inadvertently pulls the power supply for the "D" Main Steam Line radiation monitor.
STATE ALL AUTOMATIC ACTIONS and ALARMS that will occur. (1.5) l
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GUESTION 6.11 (1.50) j i From the following statements concerning the Standby Li quid l Control (SLC) systems at PBAPS, CHOOSE THE TRUE STATEMENT. (1.5) I l a. No temperature controls are required for the boron solution in the Unit 3 SLC storage tank.
b. When the SLC pump start handswitch is taken to either j
"A" or "B", EDTH squib valves fire, j
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c. If the SLC system is being tested locally in the full flow test mode, it is not possible for the control room operator to override local control of the SLC system.
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PBAPS 5.9.4.2.D, "High Radiation in ESW EI41uent," states that the ESW discharge is continuously monitort by a gamma sensitive scintillation detector mounted directly on the pipe.
If high radiation is detected on the ESW discharge, the proceoure gives two (2) possible system leak locations. List these two system leak locations. (2.0)
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I l i i QUESTION 7.01 (2.50) Answer the f oll owi ng questions concerning an automatic Emergency ) Di esel Generator tEDG) start on a LOCA signal.
I' a. Procedure S.B.4.J, "DG Load Restrictions Under Emergency (LOCA/ Dead Bus) Condi ti ons, " requires the operator to load the diesel to >1.4 MW within 30 minutes. 1 State the adverse consequence that could occur if the D/G is ! run without sufficient l oad AND ex pl ai n how this condition develops. (1.5) b. An ILC technician makes an error which inserts an inadvertent LOCA signal into the EDG initiation logic. This fact is ; i mmedi atel y recognized by a control room RO who decides to J trip the EDG f rom the control room using the manual trip push- . button. Will he be successful in tripping the EDG7 j Explain why or why not. (1.0) { t
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In accordance with OT-112, "Recirc Pump Trip," if the tripped retirc pump cannot be returncd to service, the OT directs the operator to establish conditions for single d loop operation. The procedure states:
" Maintain reactor thermal power below 35% and actual core flow below 45% of rated (46.] Mlb/hr) until power and flow ;
increases can be taken under the direction of a Reactor Engineer.
and Actual Core Flow = (Active Loop indicated Flow) - 0.95
' (Inactive Loop Indicated Flow)."
l. LIST TWO (2) reasons for requiring these limits when in single loop operation. (2.0) 2. AN RO reports the following plant parameters to you.
Is the plant operating within the above l i mi ts for single loop operation? Prove your answer. (1.0) Rx Thermal power = 1100 MWth Recirc FI-3-2-3-92A = 50 Mlb/hr Retirc FI-3-2-3-92B = 100 Mlb/hr
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21__EBggEQUBE@_;_NgBM9L 1_9@Ng8M9L1_EDE8gENgy_QNQ PAGE 16 50919L991986_G9NIBg6 GUESTION 7.03 (3.00) For each of the lettered conditions given below, state the emer gency trip procedures, if any, that are requi r ed to be entered.
ASSUME THAT ALL EVENTS OCCUR WHILE THE PLANT IS AT FULL POWER.
a. Drywell temperature = 150 deg F (0.5) b. Dr ywel l pressure = 2.8 psig (0.5) c. RFV water level = ~39 inches (0.5) d. Torus Temperature = 92 deg F (0.5) e. All MSL rad monitors read 3.5 times their normal full power background reading (0.5) f. Power in 6 "/. after a main turbine trip. (0.5) QUESTION 7.04 (3.00) Step DHI-5 of section DHI of T-111 instructs the operator to maximize CRD flow into the RPV with procedure 5.4.2.L,
" Maximum CRD Flow to the Reactor Vessel Under Emergency Conditions."
a. WHY does S.4.2.L direct the operator NOT to reset the scram? (1.5) b. Once CRD flow to the RPV is maximized, your RO reports that several Control Red Drive indications show a green background with "00" p osi t i on. Is this considered normal? Explain. (1.5) i
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2t__PBggggUggg_;_yg8Dg62_ggNgBdg61_EbEBGENgy_gNg PAGE 17 589196991986_G9NIBg6 QUESTION 7.05 (2.00) PBAPS CN-100 lists the symptoms of a fail J jet pump. For each of the parameters listed below, state whether an INCREASE, DECREASE, or NO CHANGE would be indicated if a jet jump failed.
a. Differential Pressure on the jet pump sharing a riser with
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the f ailed jet pump (0.5) b. Drive flow in the def ec ti ve Jet pump loop (0.5) c. Core Plate differential pressure indication (0.5) d. Reactor power (0.5) QUESTION 7.06 (3.00) l , For each of the conditions listed bel ow, state whether or not l an i mmedi ate manual reactor scram is required in accordance with l either the PBAPS Operational Tr ansi ent (OT) procedures or the l Offnormal (ON) procedures. I I a. A valid Offgas High Radiation al ar m is received in the I RUN mode. (0.5) b. Both Retirc pumps trip while a c ontr ol l ed shutdown is being conducted. The current reactor power is 35%. (0.5)
I c. All Main Steam Line Radiation monitors alarm on high 1 radiation while the plant is at full pcwer. (0.5) d. Several control rods begin drifting into the core at random. (0.5) e. An SRV f ail s open at power; RPV pressurta has dropped to 900 psig. (0.5) l
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f. SLC inadvertantl y initiates at rated plant conditions. (0,5)
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1 QUESTION 7.07 (2.50) A caution in the PBAPS trip procedures states, -
. "DO NOT USE THE MECHANICAL VACUUM PUMP IF EVIDENCE OF GROSS FUEL FAILURE EXIETS."
a. WHY is the use of the mechanical vacuum pump prohibited under these circumstances? (1.0) b. LIST THREE (3) specific control room indications that you could use to determine if gross fuel failure has occurred. (1.5) QUESTION 7.08 (1.50) STATE each of the following limits as given in PBAPS GP-2,
" Normal P1 ant Startup."
a. Maximum Heatup Rate (0.5) b. Maximum Transient Period (0.5)
. Maximum differential temperature between the reactor coolant in the operating and idle recirc loops prior to start of the idle recirc loop. (0.5)
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QUESTION 7.09 (2.00) Step 74 of PBAPS GP-2 Section I, " Pre-Startup Preparations," states: " verify that MO-29 A&B are open and mousetrapped." l l l l Explain wnat these valves are, list what plant systems are invrived, and state WHY this step is required BEFORE the ap-
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QUESTION 7.10 (2.50)
I a. Answer the f oll owing questions in accord # _e with PBAPS FH-6C. " Fuel Movement and Core Alterations Procedure During a Fuel Handling Outage."
1. How often do the fuel pool and reactor tag boards need to be updated during fuel assembly movement' (0.5)
2. Who (by title) i s responsible for updating the fuel pool and reactor tag boards? (0.5) 3. TRUE or FALSE? If the Refuel SLO determines that a change to a Core l Component Transfer Authorization Sheet (CCTAS) is I required, he can make a temporary change to the CCTAS I if he documents the change on the CCTAS, and gets the approval of a PORC member. (0.5) 6. State the immediate actions you would take as the Refuel SLO if a spent fuel assembly was dropped dur i rig fuel r el oc at i on. 11.0) i l l
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e __B9 MIN 1@l6(IlVE_PRggEDU6Eg3_ggNDillgNSg_@Ng_LIMll@llgNS PAGE 20 QUESTION 8.01 (1.50) Answer the following questions in accordance with the Procedure for Corrective and Preventive Maintenance Using CHAMPS (A-26A).
a. TRUE or FALSE: It is NOT mandatory to use INFORMATION TAGS on equipment that is located in a radi ation area. (0.5) b. TRUE or FALSE: If a component is marked with a DEFICIENCY STICKER then a MRF has been initiated, but the component has not been blocked. (0.5) c. TRUE or FALSE : A VALVE REPAIR tag is removed by the individual who actually repairs the valve. (0.5) QUESTION 8.02 (3.00) a. State the PBAPS administrative quarterly whole body dose limits, BOTH WITH AND WITHOUT an NRC Form-4, for radiation workers at PBAPS. (0.5) b. What addi t i on al access restriction is set for rsdiation areas with wh ol e body dose l evel s in excess of 1000 mrem /hr, as compared to areas with lower levels of radiation exposure? (0.5) c. For each of the following postings, list ALL radiation pro-tection entry requirements AND the maximum whole body dose l i mi t . 1. CAUTION RADIATION AREA (1.0) 2. CAUTION t HIGH RADIATION AREA (1.0)
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QUESTION 8.03 (3.00)
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Answer the f oll owi ng TRUE or FALSE questions in accordance with the PBAPS Emergency Response Plan. If it .s FALSE,' explain WHY.
l a. The Emergency Director is authorized to deestalate the emergency classification of an event without the concurrence of the Site Emergency Coordinator. (1.0) b. If the Emergency Director feels he knows a better way to control the plant emergency he is allowed to deviate from the Emergency Response Pl an. (1.0) c. The Emergency Director is authorized to declare when the plant can enter the recovery phase after the Emergency Plan has b eert activated. (1.0)
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QUESTION 8.04 (3.00) Classif y the f ollowing events in accordance with Appendix I of EP-101, " Classification of Emergencies," which is attached.
l For each, list the applicable event c l assi f i c a t i on , the classi-fication title, and explain how the classification criteria are satisfied.
Assume that the unit is in the RUN mode for each event, a. At 0800 on 10/05/87, you discover that the Standb Gas gcggypy Treatment system was made inoperable due to a mai tenance error committed on 10/03/87. (1.5) b. The following information i s reported to you: c t6ldU Containment pressure reads 2.4 psig on PR-2/3508 RWCU MD-15 fails to close on 0" RPV water l evel . j CORE SPRAY PUMP AUTO START annunciator is in. (1.5) l
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4 et__9Pd1NJgI69]]VE_PSQCEQUBE@3_CQNp]]]QN@g_@yp_LJy]]@IJgNg PAGE C2 , QUESTION 8.05 (3.00) Using the attached copy of PBAPS A-31. " Procedure for Notifi-cation of the NRC," determine the time limits for"notifytng the NRC for each of the following events.
Include in your answer: the time limit (either I hour. 4 hours, or 30 day LERs) AND the A-31 step number that describes the event (A.6.4, for example).
i Consider each case (a-c) separately.
a. An MSIV fails to close during the full closure surveillance test while the unit is in cold shutdown. (1.0) b. As the Shift Supervisor, you have just declared an ALERT in vi ew at the current plant conditions. (1.0) c. During an ATWOS, it becomes necessary to defeat the MSIV isolation logic in order to reopen the MSIVs and reestab-lish the Main Condenser as a heat sink. (1.0) OUESTION 8.06 (2.00) ggg g,g g g,g omi%L frrm TT howdewk.
PBAPS Unit 2 is in a refueling ou ge, and a core reload is in UL (Ol6117 progess. Each SRM has four (4) uel assemblies surrounding it, gggy g4s All control rods are f ull y ins.rted and electrically disarmed. aw eKh* ZO Ewdu ' The reactor operator n fi s you, as the Shift Supervisor, that SRM "A" i s read ~ ig cps.
CAN core reloadi operations continue? IF NOT, WHY NOT IF SD, under WHAT conditions?
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USE T ATTACHED PBAPS TECHNICAL SPECIFICATIONS IN AN ERING THIS QUESTION. FULLY REFERENCE ALL TE H SPECS THAT YOU USE IN DEVELOPING YOUR ANSWER.
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@__ 89 MINI @IB911VE_PSgCEQUBE@z_CQNQlIlgN@z_@Np_LIMlI@IlgN@ PAGE 23 QUESTION 8.07 ( 1. 50 ) - PBAPS Unit 2 is at 100*/. power; APRM "E" 's inoperable.
. An ILC technician, conducting an APRM cal i br.s. .i on check, discovers that the setpoint for the APRM "C" upscale +1ow-biased trip is incorrect in-the nonconservative direction. He tells you that it will take three (3) hours to return APRM "C" to within specifications.
What actions are required in accordance with the PBAPS Technical Speci f i cati ons? (1.5)
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USE THE ATTACHED PBAPS TECHNICAL SPECIFICATIONS IN ANSWERING THIS QUESTION. FULLY REFERENCE ALL TECH SPECS THAT YOU USE IN DEVELOPING YOUR ANSWER.
- i QUESTION 8.08 (3.00)
a. While Unit 3 is operating at full power, the HPCI Room Fire detector H116 is declared inoperable. What actions are required per Tech Specs? - ( 1. 5 ) b. What actions would be required in accordance with Tech Specs if H116 was found to be inoperable while Unit 3 was in cold , shutdown? (I.5)
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USE THE ATTACHED PBAPS TECHNICAL SPECIFICATIONS IN ANSWERING THIS QUESTION. FULLY REFERENCE ALL TECH-SPECS THAT YOU USE IN DEVELOPING YOUR ANSWER.
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! -QUESTION 8.09 (2.50)
Unit 3 ts in a refueling outage; you are'the on-coming. Unit 3.
Shift Supervi sor. You learn during shift turnover'that painting j of the Torus internals is in progress.
" During your shift SBGT train "A" is declared inoperable.
The Refuel Floor SLO calls you to talk about the possibility of commencing the care offloading. Can this be done? If not, why not? If so, under what conditions? (2.5)
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USE THE ATTACHED PBAPS TECHNICAL' SPECIFICATIONS IN ANSWERING THIS QUESTION. FULLY REFERENCE ALLL TECH SPECS THAT YOU USE IN DEVELOPING YOUR. ANSWER. j
************************************************************** j GUESTION 8.10 (2.50)
While Unit 2 i s at power, Main Steam Line Radiation monitor
"D" becomes inoperable. ' List all actions required per the PBAPS Technical Specifications. (2.5) **************************************************************
USE THE AT1 ACHED PBAPS TECHNICAL SPECIFICATIONS IN J ANSWERING THIS QUESTION. FULLY REFERENCE ALL f TECH SPECS THAT YOU USE IN DEVELOPING YOUR ANSWER. l
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15 81 5.__IHEQBY_g[_ NUCLE @6_EgWE8_E(@NI_gEEB@IlgNz_ELUlDSt_@ND PAGE 25 JHEEDOD]NSDJ CS ANSWERS -- PEACH BOTTOM 2&3 -87/10/05-NOLONAUSKI, L.
ANSWER 5.01 (3.00) i a. (Because BWRs operate under saturated conditions: ) 1. Insert control rods power : , pressure : , temperature : (0.75) 2. Open the Bypass Valves - pressure : , temperature : (0.75) o. At 0130, RPV pressure = 67 psia; T sat = 300 deg F (0.5) At 0230, RPV Pressure = 195 psia; T sat = 380 deg F (0.5) Heatup rate = 380 - 300 = 80 deg F[pg 7g (0.5) Hw+vy un. l REFERENCE ! PBAPS GP-2, " Normal Plant Startup," p. 13 ; LOT-1530 GP-2, Obj 5 LOT 1160 Use of Steam Tables, p. 5, Obj. 2 KA Thermo: 293003 Steam, K1.23 - Use of sat steam tables (2.8/3.1) KA 241000 Rx/Turb Press Control A1.01 AB to predict change in rx pressure (3.9/3.8) KA295005 Control Rods K1.04 Predict change in power per change in rod position (3.5/3.5) 293903K123 ...(KA'S) ANSWER 5.02 (1.50) P = Po exp(t/T) (1.5) In (P/150) = 60/70 In P = 0.86 + In 150 = 5.87 P = 354 cps ,
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REFERENCE LOT-1430 Rx period, OBJ 5 ) 293003 Rx Kinetics K1.85 Define Rx period (3.7/3.7) ; K 1. 08 Given power eqn, solve probl ems (2.7/2.8) ) 293003K105 293003K108 ...(KA'S)
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9 __IbEQBy_gE_NygLE98_EgWE8_EL@NI_gfE8@IlgN1_ELylggt_999 PAGE 26 i ISE6DggyN@ Dig @ j l ANSWERS -- PEACH BOTTOM 2&3 -87/10/05-KOLONAUSK1, L. j ANSWER 5.03 (1.50) a. higher than (0.5) b. the same as (0.5) c."4 "' W O A Y t 0.4 0t M g gg (0.5) REFERENCE LOT-1690, Mitigation of Core Damage, pp. 6-26 Obj. 7 LOT - 00TO, AY'V luthwrwd7dicm KA 295031 Low Rx Water Level EA1.13 AB to .aonitor RX water level control (4.3/4.3) 295031A113 ...(KA'S) ANSWER 5.04 (3.00) a. Decrease (0.25), high pressure fluid will mix with low pressure fluid and lower the temperature difference between the cooling medium and the cooled medium OR decrease in mass flow rate (0.75).
b. Decrease (0.25), a decrease in velocity will decrease the mass flow rate of the cooling fluid and decrease the heat transfer rate (0.75).
c. Decrease (0.25), as the t o np>, r at ur e of the cooling medium approaches the temperature of the coc,l sd medium, the heat transfer rate is - reduced (0.75).
REFERENCE LOT-1240 Power Plant Components KA 291006 K1.04 (2.8/2.8) KA 291006 K1.08 (2.9/3.0) KA 293007 K1.06 (2.7/2.8) 291006K108 ...(KA'S)
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a. The banked series of pulls allows the l oc al and average core flux to increase together. (0.5) This limits individual control rod worths by the relation CRW oC / l ocal /4 avg (0.5).
b. Individual control rod worths are minimized so that the increase l in fuel enthalpy (0.5) and the potential for fuel damage is minimized should a rod drop accident occur (0.5).
REFERENCE PBAPS GP-2 " Normal Plant Startup", p. 12 LOT-1530, GP-2 Lesson Plan, p. 10 LDT-1490, Control Rod Worth, pp. 4-7 Obj. 4, 7 LOT-1620, Positive Reactivity Transients, pp. 2-3 Obj. 2, 3 ; KA 292005 Control Rods j Kl.10 Explain purpose of rod sequencing (2.8/3.3) 295005K110 ...(KA'S) l l
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ANSWERS -- PEACH BOTTOM 2L3 -87/10/05-KDLONAUSK1, L.
l i ANSWER 5.06 (3.00) I l o VC tolTilT g3(agant> a. An adequate scram margin (f or the flow biased APRM scram) '.sie must be established in order to prevent a scram upon restart ot"* p{ngdMfbWbb1 the recirc pump. A flux spike (of approximately 10%) can be expected from the first opening jog of the retirc discharge valve . M because of the i n tr oduc ti on of cooler water to the core ( b d i ,. l
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b. A rod positioned between 08 and 26. - (0.5) A'Lk W p, c. The reverse power effect could cause a slight increase in reactor power by causing the met void fraction an localized areas to I decrease. (A flow-biased APRM scram could occur.) (1.0) EFERENCE T-ll2, Recirc Pump Trip Bases, pp. 1-3 L-T-1860, Power to Flow Map, pp. 5-6 l Obj 4 - Purpose of P/F Map , J KA 20 001 Recirc ) A1.0" AB to predict change in ex power w/recirc (3.9/3.9) I KA 2 5005 Control rods Kl. 1 Define deep / shallow rods (2.4/2.5) K1 12 Describe effects of d eep / sh al l ow rod motion on axial / radial flux
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l 2.0 l ANSWER 5.07 &E _ 0 m p ti 4 J 7 1
l l a. Jt rducet L':c H4e=^! d OE u ui L u i L a ,- L;.u i m ic: : oc! a:d- 40.5 l H-- 7 :., c ;_ ; c r. g ; : g the t rc C2 g udaced I- cr the -cdiciy 21 a f ater i ns iHneV e mc t : onWwn n+v y.,4;o., de&f a y ngje,) yggk , ' b. lodine isotopes indicate the fuel cladding condition.
(The 1-131/1-133 ratio increases in the event of fuel el emen t i leakage.) (0.5) j c. Boiling at higher steaming rates causes deaeration of the reactor coolant (0.5). This maintains a lower oxygen content in the ) coolant (0.5). Therefore chloride concentrations need not be checked so regularly because of the reduced possibility of stress corrossion cracking without sufficient l evel s of both chlorides and 02 present (0.5).
REFERENCE PBAPS TS 4.6.B, Bases LOT-1030, Chemistry Control, pp. 2-5 OBJ. 2 LOT-1020, Reactor Chemi s tr y , pp. 2-7 OBJ. 1,2 KA 256000 Rx Condensate System SG5 KN of LCOs (2.9/3.6) SG6 KN of TS Bases (2.7/3.4) KA 295038 High Offsite Release Rate l l
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~i ANSWER 5.08 (3.00) a. When the last rods in the previous group are finally positioned, areas of high flux surround the first rods in the next group. (0.5) Because individual control rod worth is proportional to the thermal +1ux surrounding the rod, the first rods in a group l have more worth than the last rods. (0.5) b. Peak xenon conditions occur approximately ten (10) hours after o scram from full power. OR- Rule of thumb: the xenon peak will occur the number of hours after the scram that equals the square root of the previous power l evel i ra percent of rated. (1.0) i The xenon concentration is higher in the core's center because
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it is the area of the previous highest neutron flux. The edge i rods will therefore have more worth due to the relative suppres-sion of thermal flux in the center of the core. (1.0) REFERENCE GP-2 p.ll LOT-1500, Xenon pp. 6-7 OBJ. 3,6 LOT-1530, GP-2 p. 10 KA 292005 Control Rods Kl.09 Explain changes in CRW (2.5/2.6) KA 292006 Fi ssion Product Poisons
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l l ANSWER 5.09 (2.00) a. Fi ssi on products released by the fuel Will increase the gamma radiation levels in the containment (0.5). H2 levels would increase through the radiolysis of the water in containment (0.5) Alternate correct answers as listed on p. 29 of LOT-1690 will be accepted, b. NO (0.25). Temperatures in excess of 1600 deg F are required to prcduce a significant Zr.-steam" reaction rate; these are obtainable only if the core is partially or completely un-covered (0.75). 1 REFERENCE LOT 1690, Mitigation of Core Damage, pp. 27-31 KA 295031 Low Rx Water Level i EA2.04 AB to interpret adequate care cooling (4.6/4.8) EA2.01 AB to interpret water level (4.6/4.6) 295031A201 295031A204 . . . ( i: A ' S ) J I i
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ANSWER 5.10 (3.00)
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a. Event 2 (0.5) Event 2 causes a TSV closure.(turbine tr2p) (0.5); event 1 l causes an MSIV isolation (0.5). ,
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Because the TSVs take less time to close (0.1 sec as compared.
to 3-5 sec), this represents the fastest steam flow shutoff and therefore the greatest pressure transient (0.5). ] l b. Any two - 0.5 each ' The control rods are fully withdrawn at EOC, It therefore takes a longer time for them to insert negative reactivity once a scram signal is received.
The del ayed neutron fraction decreases over core life; this causes a faster power change for a given reacti vi ty addi ti on.
In this case, the positive reactivity' addition caused by the void collapse (caused by the pressure increase) will-cause reactor power to increase at a faster rate at EOC vs.'BOC.
l The magnitude of the void coef f icient of reactivity i s greater l at EOC than at BOC. This increases the magnitude of the power-increase caused by the void collapse.
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PBAPS LOT-1600 Pressure Tr an si ent s , pp. 3-6 OBJ 3,4 ' KA 295005 Mn Turb Trip j AK2.07 KN of i nt err el ati on of TT and Rx pressure control (3.6/3.7) KA 295020 Inad Containment Isolation i AK3.03 KN of reactor pressure response (4.1/4.1) l 295005K207 295020K304 ...(KA'S)
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ANSWER 6.01 (2.00) hpiMI ;\0f) pp.tolf/31 120 (g M IC # glred 1COww i a. An IRM upsc al e trip (cac L;cc;4 would occur (0.5). because the j IRM would read 120 on the 0-125 scale 1 2 0 ,' 1 2 3 5- ( 0 . 5 ) . i b. NONE (0.5), f # EI * the IRMs would read 20 on Range [on the 0-40 scale (0.5).
REFERENCE l LOT-250 OBJ 3,5,7 KA 291000 Sensors, detectors kl.23 Neutron Monitoring Units (3.2/3.3) 291000K120 ...(KA'S) ANSWER 6.02 (2.50) . .
(pttc ;} syn odMurice vak)
A a. High RPV water level closes MO-131, instead of the trip throttle valve, as for a RCIC turbine trip. (0.5) l The trip throttle must be reset. (0.5) ! Closing MO-131 instead of the trip throttle valve allows (0.5) auto restart at 74Y inches. :
-47 ($ 16ll"5/ 8-[ $0.
b. The RCIC flow controller will maintain a constant pump discharge j flow in spite of the reactor pressure decrease. (1.0) j l REFERENCE ' LOT-380, RCIC pp. 4-20 Obj 2,3,4 l KA 217000 RCIC K5.06 KN of operational implications of turbine operation (2.7/2.7) A2.02 AB to predict impacts of RCIC TT (3.8/3.7) A3.04 AB to monitor system flow (3.6/3.5) g 217000A304 217000K506 2170002A20 ...(KA'S) ]
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_ _ _ _ _ _ _ _ 6.__EL991_SYSlEUS_pESIGyt_Cgy16gL _999_lNS16UbEy1911gg PAGE 34 ANSWERS -- PEACH BOTTOM 2L3 -87/10/05-KDLONAUSK1, L.
ANSWER 6.03 (1.50) (f_(0(l3[G YM'l ."U m4trib - 05& a. Recirc pumps runback to 60% speed N (0.5) A 90% maximum feed signal is set for the feed pumps (0,5) through the FWCS (09) M Mbod, b Mel b. TRUE (0.5) REFERENCE LOT 520, Condensate System, pp. 16-17 Obj 6,11 0 0T Gft) , FWCr , p), s KA 256000 Condensate system A2.05 AB to predict impact of inadequate system flow (2.9/2.9) 256000A205 ...(KA*S) ANSWER 6.04 (2.00)
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a. ND (0.5) I b. NO (0.5) c. ND LK.L0ldl7 (0.5) d. y5e NO on MMmakt sh sw(pl 3i s ofMMtki %qh (O.5) M WSUW TWJ . . REFERENCE l LOT-120, Main Steam and Pressure Relief System, pp. 4-15 Obj. 2,15 i KA 239001 Main and Reheat Steam Systems ' K1. KN of c onnec ti on between Main Steam and: K1.08 Condenser Air Removal System K1.09 Steam Seals X1.18 HPCI K1.22 Feedwater System 239001K108 239001K109 239001K118 239001K122 ...(KA'S) l
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_ _ _ _ _ _ _ _ _ 6 __PL9NI_gy@IEM@_DEgigNz_CONIBOL1_AND INSTRUMENTATION PAGE 35 ] 1I ANSWERS -- PEACH BOTTOM 263 -87/10/05-KOLONAUSKI, L. ) ANSWER 6.05 (2.00)
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I . a. A manual start from the control room with the manual push l l button requires the prelube pump to start and delays the ) D/G startup for 3 mi nutes. A QUICK START bypasses the ) prelube pump. (1.0) l l b. The diesel will turn over, but the fuel racks will not open. A fail to start trip will result. (1.0) J l I I REFERENCE S.8.4.A l LOT-670, Di esel Generator and Aux i l l i ari es, pp. 32-39 Obj. 3,4 l KA 264000 Emergency Di esel s K1.06 KN of ED/G starting system (3.2/3.2) 26400K106 ...(KA'S)
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i ANSWER 6.06 (2.00) l
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yy vy. tolTin I a. ARG they will nef auto restart (0.25;, Lim up c, a ie, .m u s S-([.d) l-puth nach cutc pump sim L reset p uuiiv w L L v. , . .t 'O.75tv i b. 1. FALSE (0.5) 2. TRUE (0.5) REFERENCE LOT 350, Core Spray System, pp. 4-10 Obj. 3,4 KA 209001 Low Pressue Core Spray System K 1. 01 KN of physical connections wi th CST (3.1/3.1) K4.08 KN of interlocks for auto system initiation (3.8/4.0) j 209001K102 209001K408 ... (KA'S) i i
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ANSWER 6.07 (2.50) a. Scram (0.5) . b. No Action (0.5) c. Half Scram (0.5)
, g7 d. Rod Block g(ug( (0.5)
e. ZI - - M I'JCr0VM Q T i v\0f (6 p! M L d b N (0 U) 61 MIN. d t.XW - REFERENCE
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LOT-250, IRM p.10 Obj. 5 LUT-270. APRM pp. 7-8 Obj. 2 LOT-300. RPS, pp. 11-16 Obj. 8
KA 215003 IRMs, K4.02 KN of interlocks for reactor scram (4.0/4.0) j KA 212000 RPS, K1.01 KN of cause-effect with NI (3.7/3.9) K1.10 KN of cause-ef f ec t with Mn Turb (3.2/3.4) . K6.02 KN of effect of loss of NI (3.7/3.9) l 21500K101 215000K110 215003K402 ...(KA'S) ) i; ANSWER 6.08 (2.00) Blowdown flow robs return flow from the RHX of RWCU (0.5).
This increases the heat load on the NRHX (0.5), which causes an increase in RBCCW temperature (0.5).
Because RBCCW cools the recirc pump seals, an increase in the seal temperature also occurs (0.5). ! REFERENCE I LOT-110, RWCU LOT-460, RBCCW l KA 204000 RWCU l A1.02 AB to predict changes in RBCCW with RWCU changes (2.9/2.9) 204000A102 ...(KA'S)
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6 __PL9N1_@y@lEM@_DEgigNg_ CON 16QLt_9Np_lN@l6UM@NI@llgN PAGE 37 ANSWERS -- PEACH BOTTOM 2L3 -87/10/05-KOLONAUSKI, L. ; l ANSWER 6.09 (2.00) a. NO (0.5), this switch is only active in the SHUTDOWN or REFUEL position (0.5).
b. SDV- ALARM at 5 gallons (0.33) ROD BLOCK at 25 gallons (0.33) ,
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SCRAM at 50 gallons (0.33) REFERENCE LOT-300 RPS. OBJ 7,8 T-LOT-0300-14 LOT-70 CRD Hydraulics, pp. 15-24 OBJ 8 KA 201001 CRD CRD Hydraulic System K4.11 KN of the effect of malfunction to protect against filling I the SDV during a non-scram condition (3.6/3.6) A3.11 AB to monitor auto actions on SDV level (3.5/3.5) 201001A311 201001K411 ...(KA'S) ANSWER 6.10 (3.00) a. MSL RAD UPSCALE alarm (1.5) Half Scram, RPS Channel B Half Group I I sol ati on (Mechanical Vacuum Pump Trip, Isolation - CHECK) V . b. MSL RAD UPSCALE alarm gp -ty(p g/ (/MW (1.5) ) ibil. N Scram, ^^~ N' " r IIICk'dI Tidl. W Gr oup I I sol ati on A M'dg%g grLrad.3kNFfB, (Mechanical Vacuum Pump Trip, Isolation - CHECK) REFERENCE LOT-720, Process Rad Monitoring, p. 5 OBJ 3 KA 272000 Rad Mon System ] K1.08 KN with RPS (3.6/3.9) i K1.09 KN with PCIS (3.6/3.8) K4.03 KN of Rad Monitor Power Supply (3.6/3.9) 272000K108 272000K109 272000K403 ...(KA'S) :
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6:__PL@N1_gyglEd@_QEgl@N,,_CgNlggL _gNQ,1N@lBUdENI@llgy PAGE 38 ANSWERS -- PEACH BDTTOM 2L3 -87/10/05-KOLONAUSKI, L.
ANSWER 6.11 (1.50) b. TRUE (0,5) a. FALSE (The Unit 2 SLC tank requires no heating.) ( 0. 5 )' c. FALSE (The CR operator can override l oc al control of SLC ( 0. 5 )- through manual ^CR manipulations.)
REFERENCE LOT-310 OBJ 10,11 KA 21100 SLC K4.02 KN of SLC design features and system testing (3.0/3.2) K4.03 KN of keeping sodium pentaborate in solution (3.8/3.9) K4.08 KN of system interlocks (4.2/4.2) 21100K402 211000K403 211000K408 ...(KA'S)
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Ai-JS WER 6.12 (2.00) RBCCW HX (1.0) or RHR Pump Seal Water Coolers (1,0) REFERENCE PBAPS 5.9.4.2.D, p. 1
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KA 272000 Rad Monitoring System K1.04 KN of interrelation with component cooling water (2.9/2.9) 234000G015 234000K102 272000K104 ...(KA'S) 1
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gl0lt7ll1 l vg. tom /F7 5* V#M , ANSWER 7.01 (2.50) . Ad(g d I g y'; 'v3 - ms e m ftY NDTVW6M9 g ( w is t y e w h er % k W l a. Unless a diesel is run with ufficient load, temperatures will Wrrin* i l not get high enough to bur the carbon deposits and gases- l exhausted by the engine .5). An EXPLOSION (0.5) could result if these gases are allo ed to collect which could result in damage to EDG blower (0.5).
i b. He won't be able to trip the EDG from the control room (0.5) I until ten minutes have past since the LOCA signal initiation (0.5).
REFERENCE l PBAPS S.8.4 EDGs KA 264000 EDGs A2.03 AB to predict impact of light loading (3.4/3.4) K4.02 KN of LOCA trips (4.0/4.2) 264000A203 264000K402 ...(KA'S) l l !
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l' Z:__PBggEQUBES_;_NgBM@L _@@NgBM@L z _EMg8@ENgy_@NQ 1 PAGE 49 69919L99199L_G9N169L ANSWERS -- PEACH BOTTOM 2L3 -87/10/05-KOLONAUSKI, L.
ANSWER 7.02 (3.00)
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1. These limits place the plant in the most stable area of the Power / Flow map. .
(1.0)
gghilt LT- Luisf ri ( h @ TOC wht r ) The TSgmonitoring of APRMS, LPRMS, and Core Plate dP is not required when operating in this area. (1.0) 1 2. 1100 MWth is less than 35% of rated thermal power which ] is 1152 MWth. (0.33) ]
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Actual Core Flow = 100 M1b/hr - 0.95 (50 M1b/hr) )
= 52.5 M1b/hr (0.33)
l So core flow is too high; the limits are exceeded. (0.33) REFERENCE OT-112 Bases, pp 1-6 PBAPS Unit 3 TS 3.6.F, Bases KA 201001 Recirc K1.01 KN of recirc effect on core flow .(3.6/3.7) K1.02 KN of recirc effect on power (4.1/4.1) 202001K101 202001K102 ...(KA'S) ANSWER 7.03 (3.00) a. T-102 (0.5) b. T-102 and T-101 bY (0.5) c. 47+wiww.fl00 C.lf t6ld p gggt(burrfidh # ql$lN '(0.5) . l d. none (0.5) T 101, W *
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REFERENCE LOT-1560 Introduction to Transient Response, p. 6 Obj. 9 XA APE 295024 High Drywell Pressure . SG 11 Recognize entry conditions (4.3/4.5) ! 295010G011 ...(KA'S) I
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Z___EB9CEQUBES_;_NgBd@Lt_@@NQBd@Lt_EUEB@ENCY_@NQ PAGE 41 60919Lg@lC@6_CgNIBgL ANSWERS -- PEACH BOTTOM 2tc3 -87/10/05-KOLONAUSKI, L.
. ANSWER 7.04 (3.00) g l#[FfI7 a. If th: -,m '~ red 'N ED . e. n t ana ur oi. . cl . c a
;culd bc s e m i c- u ; If the scram is not reset, th C D '.' H1
_. en t =11 7 riti mna -- t 'ca r mur u. This would maximize CRD fl ow to the RPV.by minimizing the losses a heme rgh the C nU- Mh H,g M gg d phry , (1.5) No n '0I'Il 7 . b. M , a ereen "D3" inotcation i s,oV consi dered normal 4. hen CRD injection flow to the RPV is maximized (0.5). LEl0lF/f7 This is because the CRDs have been inserted.to the over-travel position (0.5) ( because of the high flows through ' and the resulting high pressure of the CRD cooling water .
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M "00# (nd(CAOM wil\ NOT'(LchM omd % MP _OM]' FtEFERENCE win YMI. p g/lff)7 (0S7 i
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T-111 T-111 Bases PBAPS system procedure S.4.2.L
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KA 295031 Emerg Evol- Low Rx Water Level , EA1. 4 AB to monitor CRD for low water level (3.6/3.7)' 295Rd f,-,110 ...(KA'S) ANSWER 7.05 (2.00) l l a. DECREASE in dP on jet pump sharing riser (0.5) ;
. .j b. INCREASE in defective loop drive flow (0.5)
c. DECREASE in Core Plate dP (0.5) d. DECREASE in reactor power (0.5)
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REFERENCE LDT-1553 ohs, OBJ 1 DN-100 Failure of a Jet Pump p. 7 j KA 202002 Recirc Flow Control System K3.xx KN of the effect of a malfunction on: K3.01 Core Flow (3.5/3.5) K3.02 Reactor Power (4.0/4.0) SG 15 AB to recognize abnormal indications which are entry l
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ANSWERS -- PEACH BOTTOM 2&3 -87/10/05-KOLONAUSKI, L.
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a conditions to abnormal procedures (3.8/3.9) q 202002G015 202002K301 202002K302 ...(KA*S) (
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ANSWER 7.06 (3.00) I i
l EACH RESPONSE IS WORTH 0.5 POINTS a. NO b. YES I c. NO d. YES l e. NO ~ f. tJO REFEREf CE
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PBAPS OT and DN procedures KA 294001 Plant Wide Generics ! A1.02 AB to execute procedural steps (4.2/4.2) 294001A102 ...(KA'S) ANSWER 7.07 (2.50) a. The fission products released from the fuel will eventually end up in the main condenser. The mechanical vacuum l pump will remove the radioactive noncondensible gases along with the other non-condensibles, and they will all be released through the' plant l stack ':c" "2c nr fi' tere I c- ;Si; tic - Q yrgyp M A dujc . ( 1. 0) Main Steam Radiation High alarm O (Alk.of-ho(clWP hN . . 2 40[ @ 7 l b. (0.5).
Offgas Treatment Monitoring High al arm . @ tol9(M (0.5) .j ora c t er Crrl nt E2rpig uis' ^ct;<;ty @Tbb (n SCK. M -l
Alternate correct answers wi11 be accepted,y g. g ) REFERENCE
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JJ4e Hi vtod ct(uw g g g gg LOT-1560, Trip Procedures p. 8 D # p p M M hD j XA 295338 Hi gh Diis.i te Re3 ease Rat e h M MI -
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EK2.10 Interrel ation wi th Cond Air Removal (3.2/3.4) YM* ! EK2.04 Interrelation with stack gas raoni toring system ( 3. 9 / 4 ~. 2 ) (0.10 29503BK204 295038K210 ...(KA'S) $ 10/19/b1' l r. y ll ~ l \; i I J
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60919L9G1c6L_CgNIBQL l ANSWERS -- PEACH BOTTOM 2L3 -87/10/05-kDLONAUSKI, L.
l ANSWER 7.08 (1.50) a. 80 degrees F/hr (TS LCO: 100 deg F/hr) (0.5) b. 30 seconds (0.5) c. 50 degrees F (0.5) REFERENCE LOT-1530, GP-2 OBJ 2 PBAPS GP-2 KA 294001 Plant Wide Gener i c s A1.02 AB to execute procedural steps (4.2/4.2) 294001A102 ...(KA'S) ANSWER 7.07 (2.00) i i l MO-29 ALB are the feedwater stop valves (0.5) They are mouse-trapped to prevent inadvertent closure and subsequent loss of high pressure feed (0.75). (RFPT) HPCI, RCIC all feed prior to MO-29 ALB; closure of these valves would inop these systems (0.75).
REFERENCE /WM d20A (&f LM0ld&7 PBAPS GP-2 LOT-1530, GP-2 KA 295SB1 Mai n Feedi at er K1.02 KN of cause/effect with HPCI (3.6/3.8) K1.14 KN of cause/effect with RCIC (3.1/3.1) 259001K102 295001K114 ...(KA'S)
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Z __P89CEQU6Eg_;_NQ@[@61_g@NQBD@Lg_EDEB@ENCy_gNp PAGE 44 1 59919L991 COL _CgN1ggL i ANSWERS -- PEACH BOTTOM 2&3 -87/10/05-KOLONAUSKI, L.
! i . l ANSWER 7.10 (2.50) 1
a. 1. These status boards must be updated after EACH FUEL ASSEMBLY RELOCATION. (0,5) 2. The REFUEL FLOOR SLO is responsible for updating the g4t0l
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Acc0wvdttbilib Itchn(gp OMNadual O l 3. FALSE, (NO temporary procedures changes are al l owed for CCTAS.) jve (0.5) l Wedf.
b. Suspend fuel handling operations (0.5) Evacuate the refuel floor (0.5) l REFERENCE FH-6C " Fuel Movement and Core Alterations During a Fuel Movement Datage" KA 234000 Fuel Handling Equipment SG 015 Recognize conditions that entry conditions for ADPs (3.8/4.1) KA 294001 Plant Wide generics K1.02 Knowledge of procedures (4.2/4.2) 234000G015 234000K102 ...(KA'S)
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i ANSWER 8.01 (1.50) a. TRUE (0.5) b. TRUE (0.5) c. TRUE (0.5) REFERENCE A-26A, pp. 15-16 LOT-157E Admini strati ve Procedures Obj. 39 KA 294001 Plant Wide Genertcs Kl.02 KN of tagging procedures (3.9/4.5) 294001K102 ...(KA'S) ANSWER B.02 (3.00) a. 1000 mrem /qtr w/o NRC Form 4 (0.25) 2500 mrem /qtr w/ NRC Form 4 (0.25) 6. This radiation area must be locked. (0.5) c. Radiation Area All personal dosimetry devices (0.25) RWP (0.25)
> 5 mrem /hr OR total dose >100 mrem in five consecutive days. (0.5)
Hign Radiation Area All personal dosimetry devices. (0.5) RWP Constant indicating dose meter, OR Integrated alarming dosimeter, OR
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HP escort
>100 mrem /hr (BUT < 1000 mrem /hr) (0.5)
REFERENCE LOT-1730 Radi ati on Exposure Limits OBJ 1,2 KA 274001 Plant Wide Generics Kl.03 KN of facilty rad con requirements (3.3/3.8) 294001G005 ...(KA'S)
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ANSWERS -- PEACH BOTTOM 2L3 -87/10/05-KOLONAUSK1, L. I k i h ((jh M < i 1W7 #< oiltw dL) ) ANSWER 8.03 (3.00) gp g gpclD I i
(bp Gs0 T S a. TRUE (N (1.0)
b. TRUE / (1.0) c. FALSE (0.5), the concurr ence of the SEC, the ED, and the ESO is required before the reccsery phase can be entered. (0.5) i REFERENCE lbr) %ynys 046 o r i PBAPS EMERGENCY PLAN LOT-1520, E PLAN OBJ 1,2 294001 Plant Wide Generic AJ.;c AB to taue iku E M en 12.9/4.7> 294001A11e ...(KA'S) ANSWER 8.04 (3.00) a. UNUSUAL EVENT (0,5); Loss of Secondary Containment Integrity (0.5) SBGT is required for Second Containment integrity; it has been inoperable for over 12 hours. (0,5) ALERI LF 10lME7 6. CITC u lii3dsCV (0.5); Unplanned Shutdown (0.5) Level reaches -130" for auto CS pump start Containment pressure is >2 psig and <10 psig. (0.5) , l REFERENCE LOT-1520, E PLAN SRO OBJ 1,2 KA 294001 Plant Wide Generic A1.16 AB to take actions IAW E Plan (2.9/4.7) 294001A116 ...(KA'S) J
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i ANSWER 8.05 t3.00> LEl0/6l17 - Lc ; M Mhm (OOT~ (ost- , NK MyM ID i IN [g. ,
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i U . , .3 f-yg
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a. 4 mur 3 ( U . ._3 ; meus a.o..... ri g, L b. 1 hour (0.25) - step A . e .' 1. i (0.75) h aw . ' 4W WW" a$.
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;.Xi dB
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I c. I hour (0.25) - step A.6.2.1.A). 5.a. (0.75) 1 REFERENCE 1 m t1'tM1~ N Wif d ebIAr 13 U dfd (K PBAPS A-31 LOT-1570 Admin Procedures SRO OBJ 4 LS 10 /tf[t 7 KA 294001 Plant Wide Generics A1.16 AB to take actions IAW E Plan (2.9/4.7) 294001A116 ...(KA'S) t4 10/9/r7 ANSWER 8.06 (2.00) y ggy mig &m tif.0 TT hM j(Ashe h b The reload must be st ped (0,5) because the SRM in an inter- um.
mediate array is in erable (0.5) per Tech Spec 3.10.B.S.
These arrays must e monitored during loading and unl oadi ng of fuel (1.0).
Alternate ans er: Stop the re ad and remove the bundles around the inoperable detector t en resume reload in quadrants other than the one with the ailed detector. (2.0) i REFERENCE i PBAPS T 3.5.B, pp. 227-228
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I KA 2 004 SRM GOO KN of TS LCOs (3.2/3.9) 21" 04G005 ...(KA'S)
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1 9 __09M191EI5811ME_PBQQEggBES t _QQNQlligNgt_gNQ,LIMll@llgNS PAGE 48 ANSWERS -- PEACH BOTTOM 2L3 -87/10/05-KOLONAUSK1, L.
ANSWER 8.07 (1.50) f l I RPS - TS Table 3.1.1 APRM "E", "C" inop i I This drops below the minimum number of operable channels required for RPS Trip System "A". Two are required, only one is operable.
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IAW TS 3.1.A, RPS Channel "A" must be tripped, or Action "A" (.ii or 'B" of Tabie 3.1.1 taken. LDr r'"r1 kRod Blocks - TS Table 3.2.C APRM "E", "C" inop 1AW TS 3.2.C.1, A c t i c,n p .- L . '" Tab!c '
' rl i s required.
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bo Odth ' Lic.tal5/87 ' REFERENCE
- Oahh ' Arn W e d g m T.2. a t.
PBAPS TECH SPECS LOT-1840 TS LCOs, OBJ 2, SRO OBJ 2,3 KA 215005 APRMs G995 KN of TS LCOs (3.3/4.2) 215005G005 ...(KA'S) ANSWER 8.08 (3.00) i ! a. TS 3.14.C.1 - the requirements of TS Table 3.14.C.1 must be met. Table 3.14.C.1 ref ers to TS 3.14.B.1.c which states I that one heat detector can be inoperabl e f or seven days. (1.5) l l l
b. NO actions are required; HPCI is not required to be operable I in the cold shutdown mode per TS 3.5.C.1. Theref ore the (1.5) requirements of TS Table 3.14.C.1 need not be met.
. REFERENCE PABPS TECH SPECS p. 240r LOT-1840 TS LCOs, SRO OBJ 3 ) KA 286000 Fire Protection SG 5 KN of TS LCOs (3.1/3.2) KA 206000 HPCI SG 5 KN of TS LCOs (3.6/4.3) 206000G005 286000G005 ...(KA'S) I i l i
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l ANSWER 8.09 (2.50) TS 3.7 B.1 (SEGT) requires both trains of 38GT to be operable (2.5) i when Sec Cont is required by TS 3.7.B.3; Sec Cont is required for fuel handl i rig at t i va ti es.
TS 4.7.B.3.b. requires the remaining 5BGT t.r a i n ("B") to be demonstrated as operable immediately. Piet a ^ C I 4 ' - p . ;; ; .g OM ; ut L:rr *nc&i 'g CCC' 'C- t u ..vu gunvc. sr t'5 7. 2 dok al(M SUW ret M t j 4tetig gli h applinte TS 3.7.B.1 and4 T S 3. 7. B. 3 c are be met , iTS 7 7. B. 4 deco NOT (BOTH) units to cold sh u t d own . ,' Fuel handling W prohibited.
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REFERENCE g f(~7cjggfg
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" " h" PBAPS TECH SPECS LOT-1840 TS LCOs, OBJ 2, SRO OBJ 2,3 *NY I KA 261000 SBGT SG 5 KN of TS LCOs (3.0/4.1)
261000G005 ...(KA'S) l l l ANSWER 8.10 (2.50)
(M 3 L " \) " b xc);O rd;{t ) .
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RPS - TS Table 3.1.1; <2 operable channels per trap system; (1. b{ TS 3.1.A requires that RPS "B" be tripped or TS Table 3.1.1 action A be taken.
p hh Cont Isol - (PC is required by TS 3.7) TS Table 3.2.A; there are ( 1. [) less than 2 operable channels per trip system; TS 3.2.A requires tripping of the applicable PCIS channel; or action B of Table 3.2.A.
TS 3.8.G.1 - Mechanical Vacuum pump can not be isolated automa- p ') ticall y on high steam line radioactivity; TS 3.8.G.2 must be taken - isolate the vacuum pump.
Y REFERENCE h b5 L d4WO " }) p(( [g.
PEME TECH SPECE ki([2td ; 60 u.Q /rh h LOT-1840 TS LCOs, OBJ 2,3 ) M D L -tT d M Ptth vu .yy. ,. KA 272000 Rad Mon [g.10fg g SG 5 KN of TS LCOs (2.9/3.9) -
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Awa"s 3 1 PHILADELPHIA ELECTRIC COMPANY PE ACH BOTTOM ATOMIC POWER ST ATION M. D.1. BOX 205 3 D ELT A. PEN NSYLV ANI A 173144 Mr. Allen U. S. Nuclear Regulatory Commission Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406
Dear Mr. Howe:
The attachment to this letter documents the complete formal ) comment summary for the Reactor Operator and Senior Reactor Operator J l License Examinations administered on October 5, 1987. ! All comments have been limited to those questions and/or answers which were specifically addressed during the two hour post-examination review session conducted on October 5, 1987. The facility examination review team did not feel it necessary nor appropriate to comment on questions outside those discussed with the examination authors, Ms.
Kolonauski and Dr. Robinson.
In the majority of cases, the referenced supporting. documentation l can be found in the materials forwarded to'your office ~ for exam preparation. In several cases, particularly on the Reactor Operator examination, specific references are not appropriate due to the nature' of the reviewer's comments.
1 Two general comments concerning the Senior Reactor Operator J examination are included in view of their effect on the overall -! L examination process for those candidates. These comments, j l unfortunately, are based on information which was not completely. , available during the post-examination review and, as a result, were not j discussed thoroughly with Ms. Kolonauski at that. time. ; i
Sincerely, - A O Dickinson M. Smith j Manager 1
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Attachment DMS/RGA:cj e ) cc: R. W. Bulmer J. F. Franz E. G. Firth 1 File
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Attachment.1 REACTOR OPERATOR EXAMINATION COMMENTS QUESTION 1.01 (2.0)
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Using the enclosed Power-Flow diagram (T-lDT-0040-6), answer the following: a. Insertion of control rods at full power with constant pump (1.0) speeds results in an ine:: case in core flow. Briefly explain why.
ANSWER 1.01 (2.0) a. Reduced power lowers the coro and separator pressure drop (1.0) thereby reducing resistance to flow. Thus core f1ciw increases.
REFERENCE LOT 40, Recirculation Flow Control, LO 11, pgs. 18, 19 K/A 202002 Recirculation Flow Control System Kl. Knowledge of cause-effect between RFCS and K1.02 Reactor Power (4.2) K1.03 Reactor Core Flow (3.7) K4.06 Knowledge of RFCS interlocks which prevent NPSH (3.1) FACILITY COMMENT: 1.01 a. A more appropriate answer for a Category One question would be based on core delta P drop as a result of reduction of two-phase channel flow.
REFERENCE: Lesson Plan LOT-1350, pages 6 and 7 ,
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QUESTION 1.02 (3.0) For each of the following changes, will reactor thermal power increase, decrease, or remain the same? Briefly explain each cnswer.
a. Condenser vacuum goes from 29" Hg to 26" Hg (1.0) b. Isolation of extraction steam to the Feedwater Heaters at (1.0) 90% power.
c. Sudden increase in reactor pressure (prior to a reactor _(1.0) l
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scram) ANSWER 1.02 (3.0) a. Power remains the same (0.25) Increase in condenser pressure causes a decrease in plant (0.75) efficiency but does not cause a change in thermal power b. Power increases (0.25) Due to cooler water entering reactor, there is an increase in (0.75) water density (shorter slowing down and diffusion lengths) less neutron leakage (moderator temperature coefficient) c. Power increase (0.25) Due to collapse of voids, increase in water density, (0.75) less neutron leakage (Void Coefficient) ! REFERENCE LOT 1230 Rankine Cycle, LO3, pg. 6 LOT 1250 Plant Efficiency, LO2 LOT 1440 Reactivit'y Coefficients and Defects, LO 3, 4, pg. 3 K/A 259001 Reactor Feedwater System K3.12 Know of the effect that malfunction of Feed Water System hat on Reactor Power (3.8) EPE 295025 High Reactor Pressure EK 1.01 Know. of the operational implication of press, effects on reactor power (3.9) FACILITY COMMENT: Part a. should also accept that power may decrease based on increased condensate temperatures and therefore decreased inlet subcooling which, in turn, adds negative reactivity due to decreased neutron moderation.
For all three parts of answer it should also suffice to relate power changes to reactivity changes versus neutron leakage changes since leakage is not considered a major effect at commercial BWRs.
REFERENCES: Lesson Plans LOT-1610, 0910, and 0920
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QUESTION 1.03 b. TRUE or FALSE? When starting with a Xenon. free core, it takes (0.5) longer to reach equilibrium Xenon at 100% power than it does ' at 50% power.
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1 b. FALSE (0.5) i
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LOT 1490 Control Rod Worth, LO 4, pg. 4 { LOT 1510 Xenon, LO 3, 4, pg. 5, T-LOT-1510-4 K/A REACTOR THEORY 292006 Fission Product Poisons
Kl.05 State the effect on reactor operations of equilibrium j Xenon (2.9) j K1.06 State the effect on reactor operations of Maneuvering Xenon (2.7) K/A REACTOR THEORY 292005 Control Rods K1.09 Explain direction of change in the magnitude of CRW for a change in Xenon (2.5) FACILITY COMMENT: The question is worded in such a way as to imply that times to reach equilibrium values are monitored during a single startup.
With this in mind an answer of TRUE could readily be obtained because it takes longer to reach 100% power than 50% power.
Recommend deletion of part b. to question 1.03.
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QUESTION 1.04 (2.0) a. Your reactor has just been declared critical after a refueling (1.5) outage. Because of a rod drive problem, one rod is fully inserted and disarmed and the reactor is now suberitical. The countrate does not fall to the original countrate prior to startup (when all rods were inserted). Briefly explain why.
ANSWER 1.04 (2.0) a. K-effective is less than one but not as low as it was when all (0.75) rods were in.
Therefore, the suberitical multiplication factor is higher (0.75)
(1/1-K-effective) for the one rod inserted case. Thus the countrate is higher.
REFERENCES LOT-0970 Suberitical Multiplication, LO 2, 3, pgs. 5, 8 K/A REACTOR THEORY 292008 Reactor Operation Physics.
Kl.03 Desc-ibe count rate and period response which should be rved for rod withdrawal during approach to criticality (4.1) FACILITY COMMENT: An answer based on less thermal neutron absorbers being present in I the core raising countrate due to increased fission should also bc ) acceptable.
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QUESTION 1.08 (2.0) During heatup the following data was obtained (2.0) TIME REACTOR PRESSURE 1:30 52 psig 2:00 10% psig i 2:30 180 psig From 1:30 to 2:30 what was the heatup rate? Show all work.
ANSWER 1.08 (2.0) From Steam Tables obtain reactor temperature by first corrverting PSIG to PSIA then obtaining the saturation temperature Time Pressure Temperature 1:30 67 psia 3000F (0.75) 2:30 195 psia 380 F (0.75) lleat up rate is 80 F per hour (0.5) REFERENCE
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LOT 1160 Steam Tables LO 2, pg. 5 K/A THERMO: 293003 Steam K1.23 Use of saturated steam tables (2.8) FACILITY COMMENT: Answer key shows no breakdown for partial credit based on lack of , or incorrect conversion of PSIC values to PSIA. Credit should be i given for the mechanics of attaining a heatup rate independent of the conversion.
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QUESTION 1.10 (2.0) If one recirculation pump trips while at power. OT-112 requires the operator to reduce power by driving in deep rods. Assume that a scram has been avoided and a restart of the pump is planned a. Why does the procedure require the operator to continue (1.0) to insert rods prior to pump restart? b. Explain the problem that could occur if shallow rods were (1.0) inserted'instead of deep rods.
ANSWER 1.10 (2.0) a. An adequate scram margin must be established in order to (0.5) prevent a scram upon restart of the recirculation pump.
A flux spike of approximately 10% can be expected from (0.5) the first opening jog of the recirculation discharge valve, b. The reverse power effect could cause a slight increase in (1.0) reactor power by causing the net void fraction in localized areas to decrease.
REFERENCE OT-112, Recire. Pump Trip Bases, pg. 1-3 LOT 1860, Power to Flow Map, LO 4, pg. 5, 6 K/A APE 295001 Partial or Complete Loss of Forced Core Flow Circulation AKl/02 Know. of operational implications of partial loss of forced circulation and Power / Flow distribution (3.3) FACILITY COMMENT: Part a. should not require a specific value of flux spike (10%) since this value is not referenced directly in procedural immediate operator actions. Reference: S.2.3.1 A The stated answer to b. is of little importance since this occurs late in core life and only over a few notches of rod movement.
The most severe problem would be irregular flux-shaping possibly . causing localized flux spikes in excess of acceptable limits due to the fact that any rods inserted by this procedure are fully J , ' inserted not just over a few notches late in core life. K&A Catalog or lesson plan objectives do not support questioning on the reverse power effect.
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QUESTION 1.12 (1.5) q Indicate whether the change in conditions given below would cause i
the VOID COEFFICIENT to become more negative, less negative or have no effect. Consider each change independently.
a. An increase in reactor pressure (0.75) b. Removal of several control rods while power remains (0.75) constant (due to burnup) ANSWER 1.12 (1.5) a. less negative (0.25) a decrease in the void fraction gives a less negative void coefficient (0.5) b. less negative (0.25) l l l a larger " effective" core gives a less negative void
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coefficient (0,5) REFERENCE LOT 1460, Void Coefficient, LO 3, 4, pg. 3, 4, 5 i
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K/A REACTOR THEORY: 292004 Reactivity Coefficients describe the effect on the magnitude of the Void Coeff from changes in ' Kl.ll Core Void Fraction (2.5) ! K1.13 Core age (2.1) ; i FACILITY COMMENT: No explanation of why the VOID COEFFICIENT changes is requested in I the question, yet heavy point value is assigned to explanation of .i the change. Recommend deleting explanation from answer key. ;
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QUESTION 2.03 (3.0) Consider the Standby Liquid Control System (SBLC) c. Excluding alarms, list four indications that the control (1~. 6) room operator could use to verify that SBLC is injecting into the vessel. (assume that it is too early to detect any changes in power) ANSUER 2.03 c. Any four (0.4 cach) Solution tank level decreasing Injection pump discharge header pressure Injection pump run indicator lights Squib valve loss of continuity lights Maintenance valve position indicator REFERENCE I LOT 310, SBLC, LO 7, 8, pg. 5.9 { l K/A 211000 SBLC 1 K4.02 Knowledge of design for keeping sodium pentaborate in solution (3.8) , K4.08 Y_nowledge of System Initiation (4.2) A4 Ability to predict changes in parameters l A1.01 Tank level (3.6) A1.02 Explosive valve indication (3.8) A1.03 Pump Discharge press (3.6) A1.04 Valve operations (3.6) A1.10 Lights (3.7) FACIIITY COMMENT: RWCU isolation should also be accepted as an indication based on actions occurring when SBLC initiated.
REFERENCE: LOT-0310, page 9 l
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i QUESTION 2.04 (3.0) Consider the Condensate System a. What three (3) valves automatically close on low pressure in (1.0) the condensate pump discharge header? b. TRUE or FALSE 7 The plant can operate within administrative (0,5) limits at 75% power with one of the three heater strings valved out of service.
ANSWER 2.04 (3.0) a. recirculation valve (0.34) l condensate reject control valves (0.33) CRD pump suction valves (0.33) b. TRUE (0,5) REFERENCE LOT 520, Condensate System, LO 6, 11, pg. 16, 17 K/A 256000 Reactor Condensate System A 2.05 Ability to predict impact of inadequate system flow (2.9) FACILITY COMMENT: a. The terms " condensate recire, short path, or condensate short path" are other acceptable terms for " recirculation valve."
"CRD pressure control valve" may be used to explain "CRD pump suction valve." Recommend accepting these alternate terms for identical valves.
l b. Recommend deletion of part b. Requires memorization of "S" I procedure valves - not a testable item. Procedure would have to be consulted prior to determining TRUE or FALSE.
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I QUESTION 2.06 (3.0) Consider the Diesel Generators b. On a Maximum Credible Accident start, list four (4) Diesel Generator trips that are nas byTassed. (2.0) ANSWER 2.06 (3.0) b. Any four (0.5 each) (CAF) Engine Overspeed Trip (Electrical or Mechanical) Generator Phase Differential Overcurrent Neutral Overcurrent ; Manual Cardox Injection Reverse power trip REFERENCE LOT 670, Diesel Generator and Auxiliaries, LO 3, 4, pgs. 32, 39 K/A 264000 Emergency Diesels I K4.02 Knowledge of emergency generator interlocks for emergency generator trips (LOCA) (4.0) Kl.06 Knowledge of cause-effect for starting system (3.2) FACILITY COMMENT: Reverse power trip is not applicable.
REFERENCE: Lesson Plan LOT 0670, page 39.
l QUESTION 3.03 Consider the Reactor Recirculation Flow Control System c. Briefly explain why the recirculation pump must be (1.0) started with a minimum of a 10% speed control signal.
ANSWER 3.03 l c. To insure MG does not receive a scoop tube lockout when field. (1.0) breaker closes REFERENCE LOT 30, LO 5, pgs. 20, 21 LOT 40, LO 7, 9, pgs. 8, 10 l K/A 202001 Recirculation System A4.01 Ability to manually operate recirculation pumps l (3.7) l K4.10 Knowledge of pump start permissives (3.3) FACILITY COMMENT: The M-G Scoop tube lockout on loss of speed signal is independent of the field breaker position. Therefore the statement ... "when j field breaker closes" should not be a required part of the answer.
REFERENCE: Lesson Plan LOT-0040, diagram T-LOT-0040-3 l l
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Question 3.04 (3.0) Consider the Wide Range Yarways , c. At a reactor pressure of 200 psia, does water level read (1.0) higher or lower than the actual water level? Briefly explain 1 your answer.
ANSWER 3.04 (3.0) c. higher (0.25) The water temperature in the variable leg is less than the (0.75) actual water temperature. Therefore the density of the water l the variable leg is greater than the actual water density. ; REFERENCE LOT 50, Reactor Vessel Instrumentation, LO 6, pg. 7, 8, 17, 18 K/A 216000 Nuclear Reactor Instrumentation K5.01 Knowledge of concepts of vessel level instrumentation (3.1)
A2.01 Ability to predict impact of detector equalizing { valve leaks (2.9) j A2.08 Ability to predict impact of elevated containment (3.0) temp.
FACILITY COMMENT: 3.04.c. This may also be answered " higher" due to being at lower temperatures and pressures than that required for calibration conditions.
REFERENCE: LOT 0050, pg. 26 l
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i QUESTION 3.05 (3.0) Consider the Automatic Depressurization System (ADS) . I b. For UNIT TWO, what four means are available for reclosing the (2.0) l ADS Valves once initiation has occurred? ; ANSWER 3.05 (3.0) b. Allow reactor pressure to decay below 50 psig; (0.5) Depress the "A" and "B" Timer Reset pushbuttons (to deenger- (0.5) gize logic relays and break seal-in) Shutdown the RHR or Core Spray Pumps (0.5) Place Keylock switches "A" and "B" in isolate (0.5) I REFERENCE LOT 330 LO 2, 3, pg. 8
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K/A 218000 ADS K4.03 Knowledge of ADS logic control (3.8) A2.06 Ability to predict impact of ADS initiation signal present (4.2) FACILITY COMMENT: Answer key should say " Place keylock switches A and B in inhibit" not isolate.
REFERENCE: T-224, pg. 2 i l i I l I i
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_ QUESTION 3.06 (2.5) For each of the lettered conditions given below, indicate which of the following will AUTOMATICALLY occur: (If more than one. action occurs, state the most severe action, i.e., half-scram is more severe than a rod block). Assume no operator action.
i Scram 11 Half-scram lii Rod block iv No action e. At 60% power, the APRM flow converter fails downscale. (0.5) ANSWER 3.06 (2.5) e. iii. - Rod Block (0,5) REFERENCE LOT 250 IRM , LO 5, pg. 10 LOT 270 APRM, LO 2, pg 7, 8 LOT 300 RPS, LO 8, pg. 11, 16 K/A 215003, IRM's K4.02 Knowledge of interlocks for reactor scram (4.0) K/A 212000 RPS K1.01 Knowledge of cause-effect with nuclear inst. (3.7) K1.10 Knowledge of cause-effect with main turbine (3.2) K6.02 Knowledge of effect of loss of nuclear inst. (3.7) FACILITY COMMENT: Should b0. 11 halt-scram. .66W + 54% is still the scram setting. Without a flow signal, .66W goes to zero and the scram setting is 54% power. At 60% power this would be a half-scram.
REFERENCE: LOT-270 l
- _ - _ _ _ _ - _ _ QUESTION 3.07 (2.0) I b. Which of the combinations given below of the Control Room (1,0) Ventilation Supply RAD Monitors would cause an isolation of the control room vent? i Hi A and Failed B 11 Hi Hi A and Hi B iii Failed A and Hi Hi B ANSWER 3.07 (2.0) b. (i no isolation) (0.33)
(11 no isolation) (0.33)
lii 4xr isolation (0.34) REFERENCE LOT 720 Process Radiation Monitoring, pg. 11 l K/A 290003 Control Room HVAC Kl.01 Knowledge of cause-effect of Rad, monitors - (3.4) ,
K4.01 Knowledge of design which provide system reconfigure-tion (3.1) l A3.01 Ability to monitor automatic operation of reconfigur-ation (3.3)2 FACILITY COMMENT: i Question is worded in such a way that implies to p!ck one combination that would isolate control room ventilation Therefore, full credit should be given for answer of "lii" cr l " Failed A and Hi Hi B."
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_ _ _ _ ____ _ _ _ _ _ QUESTION 3.08 (3.0)
" Synchronous Speed Not Selected" is an automatic runback associ-ated with the EHC Logic System. List the remaining two runbacks associated with the EHC and for each:
a. Give the associated initiation signal and (2.0) i ANSWER 3 08 (3.0) l a. Load Rejection, initiated at 40% mismatch between HP (1.0) exhaust pressure and generator amps ; Loss of Stator Cooling, initiated upon a loss of stator (1.0) cooling REFERENCE LOT 590,-EHC Logic LO 5, 6, pg. 37 K/A 241000 Reactor Turbine Press Regulating System i i K4.07 Knowledge of design features providing Generator Runback (3.2) K6.16 Knowledge of effect of loss of stator water cooling on system (2.9) A2.09 Ability to predict impact of Loss of Generator Load (3.4) A.2.10 Ability to predict impact of Loss of Stator Water ! Cooling (3.1)
FACILITY COMMENT Other acceptable initiations for Loss of Stator Cooling besides
" Loss of Stator Cooling" are "High Temperature or Low Pressure".
REFERENCE: LOT-630 l l l t l
QUESTION 4.02 (3.0) b. An unlocked HIGH RADIATION AREA is encountered: ;
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i. What is the minimum and maximum radiation level you would (0.5) expect to find in this area? ii. List the three (3) entry requirements which must be met (1.5) to enter this area.
ANSWER 4.02 (3.0) b. i. between 100 mr/hr and 1000 mr/hr (0.5) 11. All dosimetry devices (0.5) i
RWP (0.5) )
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iii. (Any one of the following is acceptable) ( 0.' 5 ) Constantly indicating dose rate inst.
Integrating alarming dosimeter HP Escort REFERENCE LOT 1730 Rad. Exposure Limits, LO 1, 2, 5, pgs. 4,6,8 K/A PW Gen. 294001 ) Kl.04 Knowledge of facility radeon (3.3) K1.05 Knowledge of facility req. for controlling access (3.2) FACILITY COMMENT: The question is in two parts but the answer key is in three parts.
Parts 11 and 111 of the answer key should be combined to answer part 11 of the question. Also add the " wear of proper anti-Cs, if required" for part 11.
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QUESTION 6.06 (3.0) A fire is reported in the Diesel Generator Building (ON-114) a. Under what circumstances would a controlled shutdown be (1.0) required in accordance with ON-1147 b. Under what circumstances would ON-114 require the (1.0) operator to initiate a scram? ANSWER 4.04 (3.0) l a. If the fire brigade cannot extinguish the fire and off-site (1.0) assistance is required l
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b. If the fire jeopardizes normal plant shutdown or ECCS capabil- (1,0) l ity
i REFERENCE 1 l LOT 1550 ON Procedures, LO 2, pg. 19, 20 K/A System 286000 Fire Protection System Sys. Gen. 14 Ability to perform without reference I.O. (3.8) actions K/A System 202002 Recirculation Flow Control System Sys. Gen. 10 Ability to explain system precautions (3.3) FACILITY COMMENT: The question requires knowing initiating events for followup j actions. Objectives only require restating major operator actions l and be able to discuss the bases when given. Recommend deleting ; questions "a" and "b" and reducing value of overall question to 1 point. i REFERENCE: LOT-1550 l i
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QUESTION 4.08 (2.5) i I According to SE-1, Plant Shutdown from the Emergency Shutdown Panel, it is required that the operator place the drywell instru-ment air in service prior to leaving the control room. )
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a. Explain the manipulations required by the operator in (1.5) { order to place the instrument air in service.
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b. Briefly explain WHY this step is required. (1.0) j ANSWER 4.08 (2.5) a. Plt.ce both drywell inst. N2 valve bypass switches in BYPASS (0.5) position;
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Place drywell inst. N2 valves A 0-2(3) 969 A&B in the CLOSE (0.5) position; Place drywell inst. N2 valves A 0-2(3) 969 A&B in the (0.5) 1 AUT0/0 PEN position )i b. Drywell instrument air is essential for operator control of (1.0) j l the E H and L relief valves at the ESP REFERENCE SE-1 Plant Shutdown from ESP-Basis, pg. 2 K/A APE 295016 Control Room Abandonment AK 2.01 Knowledge of interrelationship with RSP (4.4) Sys. Gen. 10 Ability to perform 10A (3.8) FACILITY COMMENT: I Answer should not require valve designators to be listed, i.e.
"AO-2(3)969A & B in part a" and "E, H, and L in part b"
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. . QUESTION 4.11 (2.5)
In accordance with the Normal Startup Procedure, GP-2 a. Under what condition should the " Emergency Rod In" switch (0,5) be used? b. Explain the reason why the use of the " Emergency Rod In" (1.0) switch should be minimized.
c. Why must the EHC be in service prior to establishing (1.0) criticality? ANSWER 4.11 We have no answer key for this question.
REFERENCE i No REF available FACILITY COMMENT: The only reference in accordance with GP-2 (as noted in question) to the " Emergency Rod In" switch use is when it is tested (GP-2 pg. 11 step 9). The only caution is that it bypasses the group l notch control mode of RSCS (GP-2 pg. 11 step 8).
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-1 Attachment 2 SENIOR REACTOR OPERATOR EXAMINATION COMMENTS GENERAL COMMENTS:
1. License candidates had several occassions to question the examiner in order to clarify the scope and/or intent of various questions. Although this information was readily provided to the candidate initiating the request, it was not shared with others. In some instances, this has necessitated additional facility comments on particular questions to clarify apparent conflicting answers. It was not clear to the exam review team that the i examiner had documented each case as to which candidate was given what information for clarification.
2. The omission of Technical Specification 3.10 from the examination package caused undue anxiety on the part of all SRO candidates in frustrating their attempts to answer question 8.06. Although delays of 15 to 20 minutes were 1 incurred (candidates' estimates) in searching for this material, the examiner granted additional time by extending the examination end-point. We agree that the extension is the only alternative feasible in this situation _, and appreciate deletion of question 8.06; however, it is difficult to measure the effect of the thought-process disruption that arose from this event.
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QUESTION 5.01 (3.00) a. In GP-2 Section III, "Heatup to Rated Temperature and Pressure," the operator is instructed to maintain a specified heatup rate.
b. During a given heatup at PBAPS, the following information was obtained: TIME RPV PRESSURE
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0130 52 psig 0200 102 psig 0230 180 psig From 0130 to 0230, what was the heatup rate? Show all work. I (1.5) { i ANSWER 5.01 b. At 0130, RPV pressure - 67 psia; T sat - 300 deg F (0,5) At 0230, RPV pressure - 195 psia; T sat - 380 deg F (0.5) l
Heatup rate - 380 - 300 - 80 deg F. (0.5) REFERENCE PBAPS GP-2, " Normal Plant Startup", p.13 LOT-1530 GP-2, Obj 5 LOT 1160 Use of Steam Tables, p. 5, Obj . 2 KA Thermo: 293003 Steam, Kl.23 - Use of sat steam tables (2.8/3.1) KA 241000 Rx/ Turb Press Control Kl.01 AB to predict change in rx pressure (3.9/3.8) KA295005 Control Rods K1.04 Predict change in power per change in rod position (3.5/3.5) ) 293003K123 ... (KA's) FACILITY COHMENT: Heatup rate should be 80 deg F./ hour i I i
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QUESTION 5,03 (1.50) In the event of a major LOCA at PBAPS, the following events can occur. For each, choose the words that CORRECTLY complete the sentence describing the effects on RPV water level indications.
c. LPCI injection flow through recire will cause the YARWAYS to read (HIGHER THAN ACTUAL / LOWER THAN ACTUAL / THE SAME AS ACTUAL) water level.
(0.5) ANSWER 5.03 (1.50) a. higher than (0.5) b, the same as (0.5) c. higher than (0.5) REFERENCE LOT-1690, Mitigation of Core Damage, pp. 6-26 Obj. 7 KA 295031 Low Rx Water Level EAl.13 AB to monitor RX water level control (4.3/4.3) 295031A113 ...(KA'S) FACILITY C0KMENT: 5.03 c.
Answer could be either HIGHER TRAN ACTUAL or THE SAME AS ACTUAL. Regardless of LPCI injection. flow through recirc the YARWAYS will read erratically and higher ) than actual due to the large RPV pressure decrease and the containment- j temperature increase. The fluctuation in indicated level due to flow through recire from LPCI injection is referenced in lesson plan LOT-0050, page 27 and ]- specifically refers to the Active Core Range which is the level indication ' l operators would use under these circumstances. Flow through recirc would not affect the YARWAY indications. The support material on page 26 of lesson plan LOT-1690, incorrectly refers to YARWAY instrumetnation in lieu of the Active Core Range instruments.
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QUESTION 5.06 (3.00) a. If ONE recire pump trips while at power, OT-112 requires the operator to reduce power by d-iving in deep rods. Assume that a scram has been avoided and a restart of the pump is planned.
WHY does the procedure require the operator to CONTINUE to insert rods prior to pump restart? (1.5) b. What is the definition of a deep rod? (0,5) c. Explain the problem that could occur if shallow rods were inserted instead of deep rods. (1.0) ANSWER 5.06 (3.00) a. An adequate scram margin (for the flow biased APRM scram) (0.5) must be established in order to prevent a scram upon restart of the recire pump. A flux spike (of approximately 10%) can be expected from the first opening jog of the recirc discharge valve (0.5) because of the introduction of cooler water to the core (0.5).
b. A rod positioned between 08 and 26. (0.5) c. The reverse power effect could cause a slight increase in reactor power by i causing the net void fraction in localized areas to decrease. (A flow-biased APRM scram could occur.) (1.0) REFERENCE OT-112, Recirc Pump Trip Bases, pp. 1-3 LOT-1860, Power to Flow Map, pp. 5-6 l Obj 4 - Purpose of P/F Map l KA 202001 Recirc K1.05 AB to predict change in rx power w/recirc (3.9/3.9) KA 295005 Control rods Kl.11 Define deep / shallow rods (2.4/2.5) Kl.12 Describe effects of deep / shallow rod motion on axial / radial flux FACILITY COMMENT: a. Change point values to (0.75) for " establishing adequate scram margin" and (0.75) for "a flux spike can be expected from the introduction of cooler water to the core". Jog of the recire discharge valve is not stressed as relevant to the overall understanding of the procedure in OT-112.
c. ' Reverse power effect is not covered in K & A's nor any LOT lesson plan or lesson plan obj ective. Responses concerning shallow vs. deep rods movements would center around upsetting the flux axial-shape throughout the core and the adverse effect of localized power peaks as a result. Suggest deletion of part c.
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QUESTION 6.01 (2.00) a. With the mode switch in STARTUP, and IRM "C" reading 12 on Range 5, what TRIPS, IF ANY, would occur if IRM "C" was down ranged to Range 4? EXPLAIN.
(1.0) b. If the IRMs are indicating 20 on Range 6 and an operator down ranged to Range 5, WHAT TRIPS, IF ANY, would occur? EXPLAIN. (1.0) ANSWER 6.01 (2.00) a. An IRM upscale trip (rod block) would o cur (0.5), because the IRM would read 120 on the 0-125 scale (> 108/125) (0.5).
b. NONE (0.5), the IRMs would read 20 on Range 4 on the 0-40 scale (0,5).
REFERENCE ! LOT-250 OBJ 3,5,7 . KA 291000 Sensors, detectors i K1.20 Neutron Monitoring Units (3.2/3.3) f 291000K120 ...(KA'S) ) l 1 FACILITY COMMENT a. 120/125 of scale is also a half-scram.
Reference: Lesson plan LOT-0250, page 9.
b. Answer Key should read "on Range 5" vice Range 4.
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.. .. .. .. . . - - . .. . . . - .. _ _ _ _ _ - - _ - QUESTION 6.02 (2.50) a. Briefly explain how RCIC system component response differs between a high reactor water level trip and a RCIC turbine trip. Indicate the REASON for this difference. (1.5) ANSWER 6.02 (2.50) a. High RPV water level closes MO-131, instead of the trip throttle valve, as as for a RCIC turbine trip. (0.5) The trip throttle must be reset. (0.5) Closing MO-131 instead of the trip throttle valve allows auto restart at -49 inches. (0.5) REFERENCE LOT-380, RCIC pp. 4-20 Obj 2,3,4 KA 217000 RCIC K5.06 KN of operational implications of turbine operation (2.7/2.7) A2.02 AB to predict impiets of RCIC TT (3.8/3.7) A3.04 AB to monitor system flow (3.6/3.5) 217000A304 217000K506 2170002A20 ...(KA'S) FACILITY COMMENT a. RCIC Autostart is -48 inches.
Reference: Lesson plan LOT-0380, page 13.
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a QUESTION 6.03 (1.50) a. What TWO (2) automatic actions will occur if one condensate pump trips while the reactor is at full power? (1.0) ANSWER 6.03 (1.50) a. Recire pumps runback to 60% speed ,
(0.5)
A 90% maximum feed signal is set for the feed pumps through the (0.5) FWCS REFERENCE
LOT 520, Condensate System, pp. 16-17 Obj 6, 11 KA 256000 Condensate system A2.05 AB to predict impact of inadequate system flow (2.9/2.9) 256000A205 ... (KA'S) EACILITY COMMENT ; a. Additional EHC runback to 95% will also occur.
Reference: Lesson plan LOT-0550, page 9.
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-QUESTION 6.04 (2.00)
An inadvertent closure of the INBOARD MSIV on Main Steam Line "A" occurs while the reactor is operating at 70% power.
For the following equipment and systems listed below, state whether or not their steam supply would be interrupted.
a. High Pressure Coolant Injection (HPCI) turbine (0.5) b. Reactor Feed Pump turbines (0.5) , c. Main Turbine Steam Seals (0,5) d. Steam Jet Air Ejectors (0,5) ANSWER 6.04 (2.00) i a. NO (0.5) b. NO (0.5) f
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c. NO (0.5) j d. YES (0.5) l REFERENCE LOT-120, Main Steam and Pressure Relief System, pp. 4-15 Obj. 2,15 )
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KA 239001 Main and Reheat Steam Systems ] Kl. KN of connection between Main Steam and: I K1.08 Condenser Air Removal System Kl.09 Steam Seals Kl.18 HPCI j K1.22 Feedwater System 239001K108 } l 239001K109 239001K118 239001K122 ...(KA'S) 1 l FACILITY COMMENT: d. Although diagrams T-LOT-0120-1 and page 4 of LOT-0120 indicate the SJAE steam supply will be interrupted initially, a back-up steam source will be established via back flow from the main turbine steam chest through the open turbine stop valves. Either a YES or NO response could be considered accurate in light of the use of " interrupted" as a break in continuity, as opposed to an isolation or completo shut-off.
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QUESTION 6.06 (2.00) a. The reactor pressure is 350 psig and the core spray system automatically initiated because of low water level in the reactor vessel. When water ! level was recovered, the operator placed all four core spray pump handswitches to STOP.
If the water level subsequently falls below the core spray initiation setpoint, will the Core Spray pumps automatically restart? Briefly explain why or why not. Assume no operator action, (1.0) ANSWER 6.06 (2.00) i a. NO, they will not auto restart (0.25); the operator must push each auto pump I start reset pushbutton first (0.75). )
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REFERENCE LOT-350, Core Spray System, pp. 4-10 Obj. 3,4 KA 209001 Low Pressure Core Spray System i Kl.01 KN of physical connections with CST (3.1/3.1) ) l K4.08 KN of interlocks for auto system initiation (3.8/4.0) j 209001K102 209001K405 ..(KA'S) l FACILITY COMMENT: 6.06 a. Answer. The core spray pumps will restart in this condition. Reference ! attached prints M-1-S-40 sheets 2 & 3. Relay 14A-K21A is the reset relay.
l It is armed by taking switch 14A-55A to stop which opens contact 14A-K21A in i l the pump start circuit thereby preventing the pumps from starting, After ) i the switch spring-returns to normal, the only way of keeping 14A-K21A l energized is through the parallel circuit containing contact 14A-K10A. This l contact is the LOCA relay as shown on sheet 3. If a LOCA is in, the " Reset" button on this circuit is armed if the pump was taken to stop (contact 14A-K21A). If the LOCA has cleared, as stated in the question (>-130"), then 14A-K10A drops out, de-energizing relay 14A-K21A, reclosing contact 14A-K21A l on the pump start circuit. Therefore, when the LOCA has cicared, the stop seal-in circuit automatically resets and the pump logic is now aligned for another automatic restart.
REFERENCE: Prints M-1-S-40, sheets 2 and 3.
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QUESTION 6.10 (3.0) a. Main Steam Line radiation monitor "B" fails upscale. (The instrument indicates past the Hi-Hi setpoint.)
STATE ALL AUTOMATIC ACTIONS and ALARMS that will occur. (1.5 ) b. While attempting to repair Main Steam Line radiation monitor l "A" which has failed upscale, an I&C technician inadvertently pulls the power supply for the "D" Main Steam Line radiation l mcnitor.
STATE ALL AUTOMATIC ACTIONS and ALARMS that will occur. (1.5) l ANSWER 6.10 (3.0) a. MSL RAD UPSCALE alarm ' (1. 5 ) Half Scram, RPS Channel'B Half Group I isolation (Mechanical Vacuum Pump Trip, Isolation - CHECK) b. MSL RAD UPSCALE alarm (1.5) Half Scram RPS Channel A-Half Group I isolation (Mechanical Vacuum Pump Trip, Isolation - CHECK) REFERENCE LOT-720, Process Rad Monitoring, p. 5 OBJ 1 l. K/A 272000 had Mon System K1.08 KN with RPS (3.6/3.9) K1.09 KN with PCIS (3.6/3.8) T K4.03 Kt: of Rad Monitor Power Supply (3.6/3.9) i 272000K108 272000K109 '272000K403 ...(KA's) FACILITY COMMENT: a. Dvlete Mechanical Vacuum Pump Trip. Isolation, since this l occurs as part of Half Group I Isolation.
b. Should be a full scram and full Group I isolation since loss of detector voltage supply causes a detector trip with , associated half-scram and half-Group I isolation from one- ] l out-cf-two twice logic. Since A and D feed separate logic l trains a full-scram and isolation will occur. All "A" andL
"D" sensors are in separate trip systems throughout PBAPS.
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_ - _ - _ _ - -- - . l OUESTION 7.01 (2.50) l Answer the following questions concerning an automatic Emergency Diesel l Generator (EDG) start'on a LOCA signal.
a. Procedure S.8.4.J, "DG Load Restrictions Under Emergency (LOCA/ Dead Bus)' Conditions, " requires the operator to load the diesel to > 1.4 MW within 30 minutes.
l State the adverse consequence that could occur if the D/G is run without sufficient load AND explain how this condition develops. (1.5)- ANSWER 7.01 (2.50) a. Unless a diesel is run with sufficient load, temperatures will not get high enough to burn the carbon. deposits and gases exhausted by the engine (0.5). An EXPLOSION (0.5) could result if these gases are allowed to collect which could result in damage to EDG blower (0.5).
REFERENCE PBAPS S.8.4. EDGs K/A 264000 EDGs A2.03 AB to predict impact of light leading (3.4/3.4) K4.02 KN of LOCA trips (4.0/4.2) 264000A203 264000K402 ...(KA's) FACILITY COMMENT: a. More emphasis should be placed on blower damage due to lack of cooling air-flow resulting in excessive expansion of blading. This is more of a concern to engineering department than explosion potential.
, REFERENCE: See attached Engineering Department package.
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_ - _ - - - _ - _ _----- _ _.---____.._-- _ ---___- , OUESTION 7.03 (3.0) For each of the lettered conditions given below, state the emergency trip procedures, if any, that are required to be entered. j ASSUME THAT ALL EVENTS OCCUR WHILE THE PLANT IS AT FULL POWER.
a. Drywell temperature = 150 deg F (0.5) b. Drywell pressure = 2.8 psig (0.5) c. RPV water level = -39 inches (0,5) l d. Torus Temperature = 92 deg F (0.5) 4 e. All MSL rad monitors read 3.5 times their normal . l full rower background reading (0.5) l T. Power is 6% after a main turbine trip. (0.5) ANSWER 7 03 (3.0) a. T-102 (0.5) b. T-102 and T-101 (0.5) c. none (0.5) j 1 i i d. none (0.5) j ' j e. T-101 (0.5)
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f. T-101 (0.5) l
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' REFERENCE LOT-1560 Introduction to Transient Response, p. 6 Obj 9
K/A APE 295024 High Drywell Pressure l SG 11 Recognize entry conditions (4.3/4.5)
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2950100011 ...(KA's) FACILITY COMMENT: i c. Entry into T.100, SCRAM, is required when vessel level drops below 0 (zero) inches.
REFERENCE: T-100 procedure NOTE: Not all candidates were advised to limit their responses 'to : T-101 and/or T-102 only, i
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- _ - _ - _ _ _ _ - . . _ _ __ QUESTION 7.04 (3.0) Step DHI-5 of section DHI of T-111 instructs the operator to maximize CRD flow into the RPV with procedure S.4.2.L,. " Maximum CRD Flow to the Reactor Vessel Under Emergency Conditions."
a. WHY does S.4.2.L direct the operator NOT to reset the '(1.5) scram? i b. Once CRD flow to the RPV is maximized, your RO reports that i several Control Rod Drive indications show a green background )
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with "00" position. Is this considered normal? - Explain. (1.5) ANSWER 7.04 (3.0) a
a. If the scram was reset,.the SDV vent and drain valves would be opened. If the scram is not reset, the SDV will eventually fill and come to reactor pressure. This would maximize CRD flow to the RPV by minimizing the losses through the SDV. (1.5) L. YES, a green "00" indication is considered normal when CRD i injection flow to the RPV is maximize (0.5). This is because f the CRDs have been inserted to the over-travel position (0.5) : because of the high flows through and the resulting high pressure of the CRD cooling water header (0.5). (1.5) l REFERENCE T-111, T-111 Bases i PBAPS system procedure S.4.2.L j K/A 295031 Emerg Evol-Low Rx Water Level i
EA1.10 AB to monitor CBD for low water level (3.6/3.7) 295031A110 ...(KA's) FACILITY COMMENT: a. SDV vent and drain valves do not auto-open at PBAPS upon a scram reset. The directive in S.4.2.L prevents any high pressure CRD flow from being diverted to pressurizing the discharged accumulators and instead maximizes flow to the i vessel.
! REFERENCE: GP-11-E, stepts 3 and 10 b. No, this is not normal. If CRD's have been inserted to the overtravel posit?on the "00" position indication will ! not activate and only the green backlighting will appear. 1 REFERENCE: Lesson Plan LOT-0080 Page 17, see switch SS1 .;
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QUESTION 7.07 (2.50) A caution in the PBAPS t'ip procedures states, .
"D0 NOT USE HE MECHANICAL VACUUM PUMP IF EVIDENCE OF. GROSS FUEL FAILURE EXISTS."
a. Why is the use of the mechanical vacuum pump prohibited under ! 'these circumstances? (1.0) l b. LIST THREE (3) specific control room indications that you.
could use to determine if gross fuel failure has occurred. (1.5) ANSWER 7.07 (2.50) a. The fission products released from the fuel will eventually i end up in the main condenser. The mechanical vacuum pump will remove the radioactive noncondensible gases along with the other non-condensibles, and they kdll all be released through the plant stack which has no filters for radiation. (1,0) b. Main Steam Radiation High alarm (0.5) Offgas Treatment Monitoring High alarm (0.5) Reactor Coolant Sample High Activity (0.5) Alternate correct answers will be accepted.
REFERENCE LOT-1560, Trip Procedures p. 8 l K/A 295038 High Offsite Release Rate EK2.10 Interrelation with Cond Air Removal (3.2/3.4) EK2.04 Interrelation with stack gas monitoring system (3.9/4.2) 295038K204 295038K210 ...(KA's) FACILITY CCMMENT: a. Responses will include references to the' lack of~any hold-up time via this exhaust route in lieu of- a lack of filters.
No reference to "no filters for radiation" exists in LOT-1560.
Also LOT-0500 paSe 3, II.C. emphasizes the.relatively "short" holdup time for evacuation of the condenser via the mechanical vacuum pump.
b. Reactor Coolant Sample High Activity is not in Main Control Room Alternate answers: SJAE Hi Rad Alarm Main Stack Rad Monitor Main Steam Line' Rad Monitors Drywell_ Rad Monitor Area Rad Monitors l l REFERENCES: Lesson Plan LOT-0500 l h
--. QUESTION 7.09 (2.0) Step 74 of PBAPS GP-2 Section I, " Pre-Startup Preparations," states:
" verify that M0-29 A&B are open and mousetrapped."
Explain what these valves are, list what plant systems are involved, and state WHY this step is required BEFORE the approach to critical. (2.0)
1 ANSWER 7.09 (2.0) MO-29 A&B are the feedwater stop valves (0.5). They are mousetrapped to prevent inadvertent closure and subsequent loss of high pressure feed-(0.75).
RFPT, HPCI, RCIC all feed prior to MO-29 A&B; closure of these valves would inop these systems (0.75).
REFERENCE PBAPS GP-2 LOT-1530, GP-2 K/A 295001 Main Feedwater K1.02 KN of cause/effect with HPCI (3.6/3.8) K1.14 KN of cause/effect with RCIC (3.1/3.1) 259001 K102 295001K114 ...(KA's)- FACILITY COMMENT: An additional plant system which would be isolated from the vessel in this case would be isolated from the vessel in this case would be RWCU.
REFERENCE: Diagram T-LOT-0540-1 in Lesson Plan LOT-0540 l f l1 k i _ - _ . - _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ -
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QUESTION 7.10 (2.50) a. Answer the following questions .in accordance wiM1 PBAPS FH-6C,
" Fuel Movement and Core Alterations Procedure During a Fuel' ;
Handling Outage."
1. How often do the fuel pool and reactor tag boards need to be updated during fuel' assembly _ movement?; (0.5) 2. Who (by title) is responsible for updating the fuel pool and reactor tag. boards? (0.5) ANSWER 7.10 (2.50) a. 1. These status boards must be updated after EACH FUEL ASSEMBLY RELOCATION. (0.5) 2. Tho REFUEL FLOOR SLO. is responsible for updating the tag boards. (0.5) 3. FALSE, (N0 temporary procedures changes are allowed for CCTAS.') (0.5) b. Suspend fuel handling operations. (0.5) Evacuate the refuel floor (0.5) REFERENCE FH-6C " Fuel Movement and Core Alterations During a Fuel Movement'0utage" K/A 234000 Fuel Handling Equipment SG 015 Becognize conditions that entry conditions for AOPs (3.8/4.1) K/A 294001 Plant wide generics ] l K1.02 Knowledge of procedures (4.2/4.2) ! 2340000015 234000K102 ..(KA's) l FACILITY COMMENT: a. 2. Page 5 of FH-6C also indicates that the Accountability Technician actually maintains the tag board.
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QUESTION 8.04 (3.00) Classify the following events in accordance with Appendix 1 of EP-101, " Classification of Emergencies," which is attached.
For each, list the applicable event classification, the classi-fication title, and explain how the classification criteria are satisfied.
Assume that the unit is in the RUN mode for each event.
a. At 0800 on 10/05/87, you discover that the Standby Gas Treatment system was made inoperable due to a maintenance , ' error committed on 10/03/87 (1.5) i l b. The following information is reported to you: l Containment pressure reads 2.4 psig on PR-213508.
RWCU MO-15 fails to close on~ 0" RPV water level, i CORE SPRAY PUMP AUTO START annunciator is in. (1.5) ANSWER 8.04 (3.00) a. UNUSUAL EVENT (0.5); Loss of Secondary Containment Integrity (0.5) ! SBGT is required for Second Containment integrity; it has been inoperable for over 12 hours. (0.5) b. SITE EMERGENCY (0.5); Unplanned Shutdown (0.5) Level reaches -130" for auto CS pump start l Containment pressure is ) 2 psig and < 10 psig. (0.5) REFERENCE q l LOT-1520. E PLAN SRO OBJ 1,2 K/A 294001 Plant Wide Generic
L A1.16 AB to take actions IAW E Plan (2.9/4.7) 294001A116 ...(KA'S) ,
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b. SITE EMERGENCY requires 210 psig Containment pressure as per EP-101-1, page 2. Could classify as ALERT only if-you assume Lo-Lo-Lo (-130") initiated Core Spray, however, initiation could have resulted f rom vessel pressure less than 450 psig with greater than 2 psig present in containment which would classify as an UNUSUAL EVENT.
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QUESTION 8.07 (1.50) PBAPS Unit 2 is at 100% power; APRM "E" is inoperable. An 16C technician, conducting an APRM calibration check, discovers that the setpoint'for the APRM "C" upscale flow-biased trip is incorrect in the nonconservative direction. He tells you that it will take three (3) hours to return APRM "C" to within specifications.
What actions are required in accordance with the PBAPS Technical Specifications? (1.5) ANSWER 8.07 (1.50) RPS - TS Table 3.1.1 APRM "E", "C" inop This drops below the minimum number of operable channels required for RPS Trip System "A". Two are required, only one is operable.
IAW TS 3.1.A, RPS Channel "A" must be tripped, or Action "A" or "B" of Table 3.1.1 taken. (0.75) Rod Blocks - TS Table 3.2.C APRM "E", "C" inop IAW TS 3.2.C.1, Action 10.b. of Tabic 3.2.C is required. (0.75) REFERENCE PBAPS TECH SPECS ! LOT-1840 TS LCOs, OBJ 2, SRO OBJ 2,3 ; l
K/A 215005 APRMs G005 KN of TS LCOs (3.3/4.2) 215005G005 ...(KA'S) j i FACILIIT COMMENT: Table 3.2.C is satisfied because 4 channels are avialabic requiring no action. Six (6) instrument channels provided by design. The Rod Block Trip system in RMCS differs form the PRS trip system in that there is only one trip system which contains all six APRM sensors; it is noncoincident. See Column 4 of Table 3.2.C.
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While Unit 2 is at power, bbin Steam Line Radiation monitor "D" becomes inoperable. List all actions required per the PBAPS Technical Specifications. (2.5) ANSWER 8.10 (2.50) RPS - TS Table 3.1.1; 2 operable channels per trip system; (1,0) 1 TS 3.1.A requires that RPS "B" be tripped or TS Table 3.1.1 action A be taken.
Cont Isol - (PC is required by TS 3.7) TS Table 3.2.A; there are (1.0) ! less than 2 operable channels per trip system; TS 3.2.A requires l tripping of the applicable PCIS channel; action B of Table 3.2.A. { l TS 3.8.G.1 - Mechanical Vacuum pump can not be isolated automa- (0.5) l tically on high steam line radioactivity; TS 3.8.G.2 must be taken - isolate the vacuum pump.
REFERENCE PBAPS TECH SPECS l LOT-1840 TS LCOs, OBJ 2,3 K/A 272000 Rad Mon SG 5 KN of TS LCOs (2.9/3.9) 272000G005 ...(KA'S) FACILITY COMMENT: It would not be necessary to isolate the Mechanical Vacuum Pitmp if the "D" MSL Rad Monitor Drawer is placed in the tripped condition. This will satisfy the logic to trip the pump. I
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PHILADELPHIA ELECTRIC COMPANY PE ACH DOTTOM ATOMIC POWER STATION R. D.1, DOX 20 L j l DELT A, PENNSYLV ANI A 17314 Mr. Allen U. S. Nuclear Regulatory Commission . Region I l 631 Park Avenue j King of Prussia, Pennsylvania 19406 l
Dear Mr. Howe:
In reference to your telephone conversation of 13 October 1987 j with R. G. Andrews of my staff, please include the attached information i with the Facility Comment Summary for the Senior Reactor Operator License Examination administered on 5 October.
Sincerely,
n , Dickinson M. Smith , l Manager Attachment DMS/RGA:cje cc: R. W. Bulmer J. B. Cotton File l l
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SENIOR REACTOR OPERATOR EXAMINATION COMMENTS QUESTION 8.02 (3.00) c. For each of the following postings, list ALL radiation protection entry requirements AND the maximum whole body dose limit.
1. CAUTION RADIATION AREA (1.0) 2. CAUTION HIGH RADIATION AREA (1.0) j l ANSWER 8.02 (3.00) c. RADIATION AREA
. >5 mrem /hr or total dose' >100 mrem in five consecutive days. (0.5).
HIGH RADIATION AREA
>100 mrem /hr (BUT <1000 mrem /hr) (0.5)
REFERENCE LOT-1730 Radiation Exposure Limits OBJ 1,2 KA 294001 Plant Wide Generics K1.03 KN of facility rad con requirements (3.3/3.8) l 294001G005 ... (KA's) l FACILITY COMMENT: 8.02 c. Student (John B. Cotton) asked if PBAPS or 10CFR20 limits <cre required and was told te,give PBAPS limits. Answer reflects 10CFR20 limits. Did exan.iner make appropriate notation on her , answer key for this instance?
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i l ATTACHMENT 4 NRC Resolutions of Comments on Written Examinations The following represents the NRC resolution to the facility comments made as a result of the current exam review policy.
Only those comments resulting in.significant changes to the master answer key, or were "not accepted" by the NRC, are listed and explained below. Comments made that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post exam review are not
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listed, i.e., typo errors, relative acceptable terms, minor set point changes.
SR0 EXAMINATION I General Comments: 1. The examiner took detailed notes on each question asked by the SRO 4 candidates during the written examination. The examiner documented each question with the candidate's name, the candidate's question, and the response given by the examiner. These notes were used in the grading of the examination in order to determine if conflicting candidate answers resulted because of clarifications provided by the examiner, j Additionally, it is not a good practice for the examiner to share each question with entire class of candidates because it would result in an excessive amount of disruptions to the examination. Only those questions that have the potential to generically affect the examination are announced. Questions that result from interpretations on an individual basis are not announced.
I 2. Because PBAPS Technical Specification 3.10 was required in answering > question 8.06, its omission from the candidate's Tech Spec handout invalidated the question. This situation was handled in the only fair way: the question was deleted from the examination and the candidates were allowed extra time to compensate for the time expended in attempting to answer 8.06. It is noteworthy that those candidates who hadn't yet reached question 8.06 at the time the missing Tech Spec was discovered and announced also benefitted by the extension.
Question 5.01 - Comment accepted. Typographical error.
Question 5.03 - Comment accepted in part. The question asks for the effect of LPCI injection flow through recirc on the Yarway indication. The facility comment states that " flow through recirc will NOT affect the Yarway indications." The answer is changed to "THE SAME AS ACTUAL."
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i l Question 5.06 - a. Comment accepted. The revised point distribution better represents the important information in the answer.
b. Comment rejected. The term " reverse power effect" will not be required for full credit, but the candidate will still be required to correctly compare the effects of shallow and deep rods on the q core flux distribution.
Question 6.01 - a. Comment accepted. 120 is the scram setpoint. I b. Comment accepted. Typographical error. !
Question 6.02 - j a. Comment accepted. Typographical error.
Question 6.03 - i I a. Comment accepted. Additional correct answer. j Question 6.04 - d. Comment accepted in part. Tne answer will be changed to "N0". The initial steam flow interruption referred to in the facility comment would be irr. perceptible.
Question 6.06 - a. Comment accepted. Original answer incorrect.
Question 6.07 e. See 3.06e.
Question 6.10 - a. Comment accepted.
i b. Comment accepted. Original answer incorrect. j Question 7.01 - r a. Comment accepted. The correct answer was not located in the reference material provided to the examiner.
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Question 7.03 - c. Comment accepted. Original answer incomplete.
Question 7.04 - a. Comment accepted. The correct answer was not located in the reference material provided to the examiner.
b. Comment accepted. Original answer was inaccurate.
Question 7.07 - a. Comment accepted. Original answer was , inaccurate.
b. Comment accepted. The listing includes possible alternate correct answers.
Question 7.09 - Comment accepted. Original answer was incompletei Question 7.10 - a. Comment accepted. Candidates could misinterpret the question and state the person who does the actual updating of the boards instead of citing the person who is responsible for the updating of the boards.
Question 8.04 - b. Comment accepted. Answer key in error, i Question 8.07 - Comment accepted. NO action is required for the inoperable APRMs in association with the Rod Block Monitor. Therefore, the only action required to answer the question is that in accordance with TS 3.1.A. The point distribution has been changed to reflect this.
Question 8.10 - Comment accepted. The isolation of the mechanical vacuum pump is accomplished by the action required by TS 3.2.A.
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r ATTACHMENT 4 4 RO EXAMINATION Question 1.01 - COMMENT ACCEPTED Question 1.02 - COMMENTS ACCEPTED Question 1.03 - COMMENT NOT ACCEPTED. Do not agree that the wording of the question could lead to either a true or false answer.
' Question 1.04 - COMMENT ACCEPTED , Question 1.08 - COMMENT ACCEPTED. Partial credit deducted for failure to l convert PSIG to PSIA. .
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Question 1.10 - COMMENTS ACCEPTED Question 1.12 - COMMENTS ACCEPTED Question 2.03 - COMMENT NOT ACCEPTED. Do not agree that isolation of RWCU could be used to verify that SBLC is INJECTING into the vessel.
Question 2.04 - COMMENT ACCEPTED FOR PART a.
, COMMENT NOT ACCEPTED FOR PART b. This type of question comes
, under General Guidance given in ES202 pg. 3, E3. ,
l "The examination should include questions to determine a-candidate's understanding of his responsibilities related to ' the administrative procedures, precautions, environmental and radiation release requirements, and pressure / temperature limits. ) Question 2.04c. - See 6.03a. 4 Question 2.05a.iv - See 6.04d Question 2.06-COMMENT ACCEPTED ; Question 2.08a. - See 6.06a.
( Question 3.03 - COMMENT ACCEPTED t Question 3.04 - COMMENT ACCEPTED. Only partial credit was given since the facility suggested alternate answer, did not explain why the water level read l higher.
I Question 3.05 - COMMENT ACCEPTED Question 3.06e. - COMMENT ACCEPTED
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ATTACHMENT 4 5 Question 3.07 - COMMENT ACCEPTED i Question 3.08 - COMMENT ACCEPTED I Question 4.02 - Comment concerning " wearing of proper anti-Cs, if required" not accepted since the question asks for requirements which MUST be met. { i Question 4.04 - COMMENT NOT ACCEPTED. Peach Bottom Off-Normal procedures do- j not differentiate between immediate operator actions and follow-up action but
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just call them operator actions. The answers to question 4.04a. and b.
require the operator to initiate either a controlled shutdown or a scram.
These types of actions, which are important to safety and may have to be performed in a relatively short time period, should be known by the operator.
Question 4.08 - COMMENT ACCEPTED Question 4.11 - COMMENT FOR PART a. ACCEPTED. LOT 1530, GP-2, NORMAL PLANT STARTUP explains that the " Emergency Rod In" switch should be used when a time card malfunction occurs; however, the comment suggested by the facility is accepted as an alternate answer.
COMMENT FOR PART b. ACCEPTED
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