ML20135H782
| ML20135H782 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 09/12/1985 |
| From: | Cooley R, Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20135H773 | List: |
| References | |
| 50-458-OL-85-03, 50-458-OL-85-3, NUDOCS 8509240224 | |
| Download: ML20135H782 (70) | |
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OPERATOR LICENSE EXAMINATION REPORT No. 50-458/0L 85-03 ;
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l Licensee: Gulf States Utilities l P. O. Box 2951 l Beaumont, Texas 77704 Docket: 50-458 i
Operator License examinations at River Bend Station (RBS)
Chief Examiner: k-R-6 John L. Pellet, Examiner Date Signed Approved by: , a/, 9 f5 D._.A..fooley,/SectionQKlef 06te Srigned Summary Operator license examinations for five (5) Reactor Operator candidates and twelve (12)-Senior Reactor Operator candidates were administered at the RBS facility during the week of July 29, 1985. Two (2) of the five (5) Reactor Operator candidates and nine (9) of the twelve (12) Senior Reactor Operator candidates passed these examinations and have been issued the appropriate licenses.
t 8509240224 850919 PDR ADOCK 05000458 G PDR 2
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O O RBS OPERATOR LICENSE EXAMINATION REPORT Report Details
- 1. Persons Examined SR0 Candidates:
PASS FAIL TOTAL 9 3 IZ--
R0 Candidates:
_P_ ASS _ _F_AI L. TOTAL p 5 4
- 2. Examiners J. Pellet, NRC (Chief Examiner)
R. Cooley, NRC D. Graves, NRC J. Whittemore, NRC A. Mendiola, NRC
- 3. Examination Report Performance results for individual candidates are not included in this report because these reports are placed in NRC's Public Document Room.
A. Examination Review Comment Resolution t
- In general, editorial comments or changes made during the exam,
- the exam review, or subsequent grading reviews are not addressed 1
by this' resolution section. This section reflects resolution of substantive comments made in the RBS Memorandum NT-094, dated August 2, 1985. The modifications discussed below are included in the master exam key which is provided elsewhere in this report, as are all other changes mentioned above but not discussed herein.
- 1. 1.02c The question states that both jet pumps have '
failed. Part 'c' assumes only 1 jet pump failed and is misleading.
L Resp.: ACCEPT. For future reference.
l 2. 1.10b The reference used in the key does not inform the
- reader whether the reactor is closer to or farther
! from transition boiling with no use of feedwater ,
- heaters. It only states that critical heat' flux increases. Therefore, this question should be omitted.
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0 1 Resp.: REJECT. If critical heat flux increases then the distance to transition boiling is increased. See page 9-85 of reference for discussion of subcooling effects on critical power.
- 3. 2.01a An additional correct answer is high water level in the fifth or sixth stage heaters.
P.es p. : ACCEPT. Key modified.
- 4. 2.04 Another answer per TS page B2-7 relates to a simulated fuel thermal time constant.
Resp.: ACCEPT. Key modified.
- 5. 2.05 The last sentence of the key should delete the words " cooling line". The water does not flow into the cooling line.
Resp.: REJECT. See page 6-11 of reference.
- 6. 3.02a The answer key should accept permissives for either fast or slow start.
Resp.: REJECT. Supplied material clearly indicates only 5 permissives are required for " fast or slow starts."
- 7. 3.03 The rod blocks in the key have been changed.
Correct answer is in TS Table 3.3.6-2.
Resp.: ACCEPT. Key modified to supplied setpoints.
Material not from TS retained as-is.
- 8. 3.04 ADS Inhibit switches have been added to the plant and are another correct answer.
Resp.: ACCEPT. Key modified.
- 9. 3.05b/c Values in question are false per current TS.
Resp.: ACCEPT. Key modified.
- 10. 3.07 Setpoints in question have been changed.
Resp.: ACCEPT. For future reference.
- 11. 3.08 The answer for total core flow is wrong. Per reference page 3-16, there is reverse flow in the idle loop so flow is accurate.
Resp.: REJECT. Per reference page 7-31, with 1 RCP on LFMG total flow is NOT correct.
- 12. 3.10a Question is misleading and should be deleted.
There are not 5 ways to start the DG in the maintenance mode.
Resp.: REJECT. Question specifies 5 total answers, not 5 for each mode.
- 13. 3.10b Manual stop is not an auto-shutdown. This should not be required for full credit.
o g Resp.: ACCEPT. Key modified.
- 14. 4.05b The question does not specify procedure. There e are conflicting temperatures for suppression pool requirements. TS 3/4 4-5 indicates a pool temp.
of 105F for scram.
Resp.: REJECT. A0P values are more restrictive or conservative than TS. Acting per TS would violate l procedure.-
- 15. 4.06g TS definitions for Refuel state that the mode switch shall be in either Refuel or Shutdown.
Either or both should be acceptable.
Resp. . ACCEPT. Both answers are required for full credit.
- 16. 5.03c The key appears to require that the candidate write the period equation & indicate the delayed term can be neglected. Candidates should receive full credit without doing so.
Resp.. ACCEPT. Point breakdown shown is for partial credit use.
- 17. 5.06c/d " Number of rods" should be an acceptable answer because overall rod worth depends on the total rods available.
Resp.: ACCEPT. Key modified.
- 18. 5.10a Core and core plate differential pressures given i in the FSAR are different from the key.
- Resp.
- ACCEPT. Key modified.
- 19. 6.03 Setpoint for RPCCW low hdr. press. is now 56 psig.
Resp.: ACCEPT. Key modified.
- 20. 6.04c MSIV opens differently for different test methods.
The valve could also be opened by placing the switch to OPEN if a partial closure test had been done.
, Resp.: ACCEPT. Key modified.
j 21. 6.04d Answer key is wrong. Single closure will not give a half scram.
Resp.: ACCEPT. Key modified.
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1 Resp.: ACCEPT. Key modified.
- 23. 6.07a This also can indicate failure of BOTH seals.
I Resp.: REJECT. This is specific to failure of the upper j seal only. If both failed then both pressures
- will decrease.
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- 24. 6.08 Containment flooding can only be done with the B loop.
Resp.: ACCEPT. Key modified. <
- 25. 6.09h The answer may be expressed as " UPSCALE NEUTRON FLUX".
Resp.: ACCEPT. Key modified.
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- 26. 6.091 The correct answer is 0.66W + 48%.
Resp.: ACCEPT. Key modified.
! 27. 6.12 Additional correct answers are: 1)glandseal,2) pump discharge pressure, 3) suction valves open, &
- 4) reactor water level increasing. ,
Resp.: ACCEPT. Key modified to include 1 & 2. 3 & 4 NOT .
l added. l
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- 28. 7.02b Acceptable notch positions should be 00-24.
i Resp.: ACCEPT. Key modified.
- 29. 7.02c Caution is required any time a rod is moved in high Xe conditions.
Resp. . REJECT. Question is drawn directly from facility i material. It asks where additional caution is l required.
! 30. 7.03b Turbine load limit in key is drawn from 3 curves
- which should be acceptable answers. They are
- 1)
, VCurve,2)Loadchangecurves,3) Capability
- curves.
Resp.: ACCEPT. Key modified.
i 31. 7.04b Question requires knowledge in performance of radiation surveys. The' candidates should not be graded on knowing if beta is included with whole
- body dose. Therefore full credit should be allowed for calculations using open window dose.
Resp.: REJECT. Material is pertinent per 10CFR55.21(a) &
j (1) and NUREG-1021, E5-402, p. 3. Full credit was
! given for using open window readings only if r candidate specifically stated assumption of dose to the eye.
l 32. 7.04c The question does not illicit answer given in key.
i An acceptable answer should be use of the EIP's.
Resp.: REJECT. See reference.
l 33. 7.04d Again, candidates should not be penalized for
- using the open window dose rate in the calculation.
Resp.: REJECT. See response to coment 31.
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- 34. 7.06 Since the question does not specify procedure, additional correct answers from other sources should be added to key.
- b. All actions appropriate to restart af ter a scram.
Resp.: ACCEPT. Key modified.
- 35. 7.10 There are more than 6 conditions which require emergency depressurization. Add answers from E0P-0002. Correct DW temperature is 330F.
Resp.: ACCEPT. Key modified.
- 36. 8.04b Question modified to insert " safety-related".
Resp. ACCEPT. Test modified.
- 37. 8.05a Acceptable answers should include Pdw=1.68 psig, rv water level = -145.4", or bus undervoltage.
Resp.: ACCEPT. Included in current key.
- 38. 8.05b There are additional verifications required per TS which should be added as correct answers.
Resp. ACCEPT. Key modified.
- 39. 8.06 Answer should be changed per TS B3/4 3-7 and should not require "which could impact..." for full credit.
Resp.: ACCEPT. Key modified.
- 40. 8.08 Answer should be changed per TS B3/4.9.2.
Resp.: REJECT. Coninent does not answer question.
- 41. 8.09 Parts d, e, & g should be corrected as discussed.
Resp.: ACCEPT. Parts d & g changed. Part e specifies 3 parameters required to be classified as SAE & is correct as-is..
- 42. 8.10a Emergency Director should be full credit answer.
Resp.: ACCEPT. Key modified.
- 43. 8.10c Change key per EIP-2-006 R1, p. 4, 5 and ADM-0010, p. 3.
Resp.: ACCEPT. Key modified.
- 44. 8.12 Accept Exclusion Area or Site as substitute for Owner Controlled Area.
Resp.: ACCEPT. Key modified.
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- 8. August 2, 1985 Exit Meeting Summary j At the conclusion of the site visit, the examiners and the senior resident inspector met with representatives of the plant staff to ;
! discuss the results of the examinations. The following personnel were present for the exit interview:
NRC UTILITY
- D. Chamberlain 'T. Plunkett D. Gipson i D. Graves D. Andrews C. Bogolin J. Whittemore J. Deddens W. Odell J. Pellet- C. Roberts E. Brohl !
Mr. Pellet started the discussion by noting that the examiners as a group had encountered a positive, helpful attitude in everyone i concerned. The following general topics were discussed:
- 1. Preliminary oral results are as follows: Two of the five R0's and three of the twelve SR0's are NOT clear passes.
. All others are clear passes. These results are for the oral 1 portion of the exam only and are subject to regional review. !
! 2. An area of weakness does not imply unacceptable performance i by any candidate. A weak area is simply one where knowledge i or skill is less completely developed than in other areas.
- 3. NRC will attempt to return formal results within 30 days of I leaving the site.
- 4. The following areas of weakness were observed in more than one candidate.
, a. Several candidates had difficulty speaking loudly and clearly. This reduces the appearance of ability to direct and control the plant when occupying a
- supervisory position.
- b. Several candidates had difficulty operating or discussing the operation of RHR in the steam condensing mode,
- c. In-plant component knowledge was weak (1. e., the
, ability to identify and discuss randomly selected major
. plant components).
d.
I Electrical system knowledge was very weak. Many !
j candidates had difficulty tracing power supplies and l feeder breakers. Proper operation of electrical
- components was also weak-(e.g., when and how does a
! fast or slow transfer take place or how to diagnose an instrumentbusfailure). This was compounded by-problems discovered in documentation addressed below.
- 5. The following areas of plant operation, documentation, or design were considered to have some deleterious effect on
! candidate performance.
- a. Procedures, design, and documentation relating to-
, electrical failures are extremely weak. They do not '
i lead to timely diagnosis of electrical bus failures.
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- b. Use of plant reference material was generally weak. In some cases, multiple revisions of documentation in the control room confused candidates. In other cases candidates were unable to properly identify a breaker for tagout from the prints available in the control room.
- c. The RWCU procedure does not correctly return the system to service.
- d. The generator synchronization procedure contains at least one out-of-sequence step.
- 6. The candidates expressed an interest in the general procedure followed for exams by the NRC. Resolution of this curiosity before taking the exam may reduce candidate tension.
C. GENERIC COMMENTS The following generic areas of weakness were identified during grading of the written examinations. This material is presented for facility use in evaluating training program effectiveness.
- 1. Knowledge of electrical system design and operation was very weak. For exampic, all candidates had difficulty accurately describing a 120 VAC Uninterruptible Power Supply System.
- 2. Understanding of how types of radiation affect dose rate was weak, for example, that beta is not counted as whole body exposure unless it is to the lens of the eye.
- 3. The effects of high Xenon worth during startup at E0L conditions was not clearly understood.
D. EXAMINATION MASTER COPY (SR0/R0 QUESTIONS AND ANSWERS)
The R0 and SR0 examination master and answer keys follow, including the formula and instruction sheets supplied during the examination.
I U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _BlVEB_QEUQ_1____________
REACTOR TYPE: _QWB-QEk_________________
DATE ADMINISTERED: _Q54QZ422________________
EXAMINER: _MEUQ1QL8t_ot ___________
APPLICANT: _________________________
185IBUGIl005_IQ_oEELIQou11 Uso separate paper for the answers. Write answers on one side only.
Stcple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at locst 80%. Examination papers will be picked up six (6) hours after the examination starts.
t 0F CATEGORY % OF APPLICANT'S CATEGORY
__V8LUE_ _1016L ___500BE___ _VaLUE__ ______________C6IEQQBl_____________
_25t00__ _251QQ ___________ ________ l. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25100__ _25tDQ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_25t00__ 25tQQ ___________ ________ 3. INSTRUMENTS AND CONTROLS
_25t00__ _25100 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100tDQ__ IQQtDQ ___________ ________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither given nor received aid.
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EPPL5CAbiI5~55G5AYURE
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Iz__EBINCIELE1_QE_NVQLEoB_EQWEB_ELoNI_QEEBoI100& PAGE 2 IBEBdQQ1Ued1 Cit _BEoI_IBoNSEEB_oNQ_ELVIQ_ELQV QUESTION 1.01 (3.00)
- e. At the beginning of a fuel cycle, control rod density is approximately 10 to 12% at equilibrium full power. Approximately one third into the cycle, the control rod density is about 15 to 16% at equilibrium full power. Why is there a difference? (1.50)
- b. What effect does this increase in control rod density have on the void coefficient of reactivity? Explain your answer. (1.50)
QUESTION 1.02 (2.00)
You are checking to verify whether or not a set of jet pumps (#5&6) have failed. If the set has failed, indicate whether the following i INCREASE, DECREASE, or REMAIN THE SAME. (0.50 each)
- c. Core plate DP indication.
- b. Recirculation drive flow in the loop containing the defective jet pump.
- c. The indicated DP on the jet pump sharing the riser with the 4 defective jet pump.
- d. Reactor power.
QUESTION 1.03 (2.00)
The reactor scrams at 0700 after 3000 hours0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> at 100% power. In order to meet commitments to the grid, the Operations Manager desires that the reactor be immediately restarted.
- a. What are your considerations regarding poisons in the core if the reactor is brought critical during the afternoon '
of the same day? (1.25)
- b. How will control rod density change over the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the startup? (0.75)
QUESTION 1.04 (2.00)
A cooling water pump is operating at 1000 rpm. Its capacity is 500 gol/ min at a pump head of 25 psi which requires a power of 100 KW.
Determine the pump capacity, discharge head, and power requirements if the pump speed is increased to 3600 rpm. (2.00)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
li__EBIUQlELE1_QE_UVQLE8B_EQWEB_EL8UI_QEEB81106t PAGE 3 IUEBdQQ168d101t_BE8I_IB861EEB_800_ELu1D_ELQW QUESTION 1.05 (2.50)
The reactor is brought critical at 40% on range 2 with a stable period of 50 seconds. Heating power is determined to be 40% on range 8 of the IRM's.
- c. What is doubling time if period remains constant? (1.00)
- b. How long will it take for power to reach the point of adding heat if period remains constant? (1.50)
QUESTION 1.06 (1.00)
The reactivity worth of a single control rod will ____________,
(For each statement below indicate INCREASE or DECREASE) o, if the void content around the rod decreases. (0.25)
- b. if the moderator temperature increases. (0.25)
- c. if an adj acent control rod is withdrawn. . (0.25)
- d. if Xe-135 concentration around the rod decreases. (0.25)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
1 lt__EBINCIELEH_QE_NVQLEo8_EQWEB_EL6NI_QEEBollQNu PAGE 4 IBE30001Ned1GSt_BEoI_IB8NSEEB_oNQ ELul0_ELQW QUESTION 1.07 (2.00)
Match the Failure Mechanism from column 'l' and the Limitir.g Condition from column '2' with the associated Power Distribution Limits (a-c) below.(Your answer should have letter-number-number) (2.00)
- c. Linear Heat Generation Rate (LHGR)
- b. Average Planer Linear Heat Generation Rate (APLHGR)
- c. Minimum Critical Power Ratio (MCPR) 1-FAILURE MECHANISM 2-LIMITING CONDITION
- 1. Fuel clad cracking due 1. It Plastic Strain to lack of cooling caused by DNB.
- 2. Fuel clad cracking due to 2. Prevent Transition Boiling high stress from pellet expansion
- 3. Gross clad failure due to 3. Limit clad temp to 2200 F decay heat and stored heat following a LOCA QUESTION 1.08 (2.00)
- e. What is the minimum reactivity that must be added to a critical reactor for it to be prompt critical? (1.00)
- b. HOW and WHY does the delayed neutron fraction change over core life? (1.00)
QUESTION 1.09 (2.00)
- c. What effect would there be on plant efficiency if the temperature of the condensate leaving the main condenser hotwell was lowered more than norn.al to 30 F below the saturation temperature? (0.75)
- b. Why does this effect occur? (1.25)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
l lt__EB16CIELES_QE_UUCLE68_EQWEB_EleUI_QEEBoI1QUt PAGE 5 IBEBdQQ18851 cst _BEoI_IB6USEEB_oNQ_ELulD_ELQW QUESTION 1.10 (2.00)
State whether each of following statements are true or falst;
- o. As condenser vacuum is increased (pressure decreased), more energy can be extracted from the steam. (0.50)
- b. The use of feedwater heaters places the reactor farther from transition boiling than not using feed water heaters. (0.50)
- c. Air ej ectors would not be needed if the condenser Wes absolutely air tight. (0.50)
- d. The main condenser uses the conduction mode of heat transfer to reject heat to the circulating water system. (0.50)
QUESTION 1.11 (2.50)
Explain what would happen to core flow DISTRIBUTION on a power increase with NO CHANGE IN MEASURED CORE FLOW if the core fuel bundles were NOT ORIFICED. t (2.50)
QUESTION 1.12 (2.00)
Briefly explain how the SRM's could be used during accident (low level) conditions to provide a crude method for water level indication
^
within the core regions. (2.00)
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-2t__EleNI_QE11GN'_INCLUDINE_18EEIX_8NQ_EdEBGENGl_111IEd1 PAGE 6
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QUESTION 2.01 (2.50)
- c. List three causes for-a GENERATOR LOAD-SET RUN8ACK. (1.00)
- b. The ESTIMATED CAPACITY' CURVES are used to protect against what? (0.50) 11
, c. List TWO reasons why Hydrogen is better than air for the generator ,
h' cooling medium. (1.00) {
i QUESTION 2.02 (2.00) ,
?
- When operating the Residual Heat Removal (RHR)' system in SHUTOOWN
- COOLING MODE:
! a. WHY is it necessary to insure that the RHR piping is
! completely filled prior to-opening the SHUTOOWN COOLING
! ISOLATION VALVES? (1.00)
- b. WHY must you establish greater than 750 GPM flow within l- 8 seconds of a RHR pump start? -
(1.00)
QUESTION- 2.03 (2.50) f- List five (5) SIGNALS / CONDITIONS by which a SRV may be opened. (2.5) l 1 QUESTION 2.04 ..
(2.00)
. State the purpose of the time delay in-the UPSCL THERMAL PWR TRIP. (2.00) i L
! QUESTION 2.05- (2.00)
) When control rods are being moved, control rod-drive water enters the oxhaust header. Where does the water in the exhaust header go and how j does it get there? (2,00) i i
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- . l 2t__ELaNI_DESIGU_16CLUDINQ_SaEEIl_aND_EUEBGEUCX_SYSIEd1 PAGE 7 QUESTION 2.06 (3.00)
- b. What is the reason for this value? (0.50)
- c. LIST four RhCUpumpstrips and their setpoints (if any) (2.00)
QUESTION 2.07 (3.00)
Sketch one Standby 120V A-C Uninterruptible Power Supply System. Include in your answer all breakers or disconnects, power sources, switches, and other active components. Label all subsystems and devices shown. (3.0)
QUESTION 2.08 (2.50)
- c. The RCIC flow transmitter providing the input to the RCIC flow controller has failed to minimum (0 gpm output). RCIC is in a norasi standby lineup. How will the RCIC turbine respond if an initiation signal is received? Continue your answer until the turbine is in a stable condition. (1.0)
- b. How is radioactive steam prevented from leaking past the RCIC turbine gland seals into the atmosphere? (1.0)
- c. What is the only RCIC turbine trip that must be reset locally?
SETPOINT NOT REQUIRED. (0,5)
QUESTION 2.09 02.00)
- o. What systems provide the normal and alternate heat sinks for the RPCCW System? (0.5)
- b. Indicate what actions occur in the RPCCW System at each of the following pressures. Include pumps starting and/or stopping and valve-realignments. Valve numbers not required. (1.5)
- 1. 100 psig
.2. 56 psig
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CATEGORY 02 CONTINUED ON NEXT PAGE
2___EL881_QE11GN_INQLUDINQ_18EEI1_aN2_EMEBQENQ1_111IEd1 PAGE 8 QUESTION 2.10 (3.50)
Answer the following with regard to the STEAM BYPASS VALVE (SBV) and TURBINE CONTROL VALVE (TCV) operating systems:
- a. Are the f ast-acting solenoid v alves for each operating system normally ENERGIZED or OEENERGIZED? (0.5)
- b. WHAT FUNCTION is performed on the TCV and SBV by their respec-tive fast-acting solenoid valves, and WHAT PROTECTION does the function provide? (2.0)
- 1. Loss of electrical signal to the servo valve coils (0.5)
- 2. Loss of hydraulic fluid (FJS) to the control jet pipe (0.5)
(***** END OF CATEGORY 02 *****)
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2t__1N11BVdEN11_6NQ_GQNIBQL1 PAGE QUESTION 3.01 (3.00)
HOW will the FEEDWATER CONTROL SYSTEM (FWCS) respond to each of the following instrument failures? Your answer should include how the instrument will fail (high or low), the FWCS response and the effect
'on actual water level. ASSUME THAT YOU ARE OPERATING AT 100 % POWER.
- n. A steam flow detector's sensory diaphragm ruptures. (1.00)
- b. The level. channel selected for input to the FWCS loses instrument power. (2.00)
QUESTION 3.02 (3.00)
- s. List the five REACTOR RECIRCULATION PUMP start permissives which must be satisified prior. to starting a pump-in slow or fast speed. (1.50)
- b. When s' RECIRCULATION PUMP is. started, the pump' remains at fast-speed or transfers to slow speed.
- 1. Other than reactor level, what directly measured parameter determines the. final pump speed? (0.50) 2..What condition does this interlock / requirement prevent? (1.00)
. QUESTION. 3.03 ('3.00)
LIST FIVE WITHDRAWAL rod blocks-NOT associated with the SRM,IRM AND APRM SYSTEMS and-REFUELING platforms. Include their SETPOINTS and when each is BYPASSED.- (3.00)
QUESTION 3.04 (1.50)
Assuming valid initiation signals exist,.name TWO conditions which will close the, ADS. valves once blowdown has commenced? (1.50) 1 i
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LC*****' CATEGORY _03 CONTINUED ON NEXTLPAGE
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2t__INSIBQMENIl_6NQ_QQNIBQLg PAGE 10 i
't QUESTION 3.05' (2.00)
Answer the following statements'about the APRM system either TRUE or FALSE:
{
- b. 'The UPSCALE THERMAL' POWER TRIP UNIT's maximum trip setpoint is 120%.- (0.50)
- c. The UPSCALE'HIGH TRIP' UNIT trips at.120% when the mode [
. switch is in RUN. (0.50)
- d. The. UPSCALE HIGH ALARM TRIP UNIT's maximum trip setpoint j is 108% when the' mode switch is in RUN. (0.50)
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QUESTION 3.06 (3.25) t
- a. LIST the five (5) the reactor water level ranges and their range j of indication. (1.50) ,
- b. Which ranges are calibrated for hot operating conditions, i.e. 1035 4
psig saturated. conditions in the-vessel ? (0.75)
, c. .There is a condensing chamber which taps off'the vessel l head vent. Which range (s) use this~ chamber?
(1.00) 1 4
QUESTION 3.07 (2.75)
For the following REACTOR PRESSURE AND LEVEL setpoints,fgive
-the AUTOMATIC. ACTIONS which occur. (2.75) i a.-.50 psig
- b. 135~psig E c. 1130.psig.
I d. 10 "
- e. - '144 "
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- 2A__INSIBMUEN11_8NQ_QQUlBQL3 - PAGE 11 QUESTION. 3.08 (1.50)
Your reactor is shutdown with A RECIRC pump running, and B RECIRC pump secured.
- a. Are your. indications for TOTAL CORE-FLOW and running LOOP FLOW correct ? (0.50) b.'Whyoor why not ? (1.00)
- QUESTION 3.09 (2.00)
BRIEFLY EXPLAIN-what condition will gener. ate EACH of the following.
indications on the Operator Control Module.-
- 1. Data Fau'lt (0.50)
- 2. Scram Valves (0.50)
- 3. . Channel Disagree (0.50)
- 4. Insert Required (0.50)
QUESTION 3.10 (3.00).
- a. LIST ALL the_ ways (5 total) the STANOBY DIESEL-GENERATORS will' startand accelerate-to. normal speed.when they are in the:
- 1. MAINTENANCE MODE (0.50)
'2. OPER^TIONAL-MODE (1.00)
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- b. When a Standby diesel generator'is running in the Emergency Operational. mode, what signals will cause it to Auto-shutdown and its-output breaker to trip ?. (1.50).
(*****
END OF CATEGORYLO3.'*****)
, St__BBQQEQQBEl_:_dQBd8kt_60NQBd8Lt_EdEBGEUQ1_6NQ PAGE 12 88Q10LQQ1QaL_QQUIBQL t
4 QUESTION 4.01 (3.50)
State the FOP PROCEDURE entry conditions and setpoints for PRIMARY
- CONTAINMENT. CONTROL. (3.50) f QUESTION 4.02 (2.50)
In the general notes of E0P 0001 procedure, the. standard operating practice of.taking manual control of. an'ECCS system is discussed.
- Under what conditions can manual control of an ECCS system'be allowed?
- (2.50) 1 4 i P
~
f QUESTION 4.03 (2.50) -
i According to GOP-0001, " PLANT.STARTUP.TO LPAP (35%)", 4 I a. How do you maintain RV level betweenR35" and 45" on the
- - wide range level indication with the MSIVs shut and the core heating up?- (1.00)
- b. What APRM/IRM indications must the operator verif y j ust prior to placing the mode selector switch in RUN? (1~.5L) i 4
QUESTION 4.04 .(2.75)
- a. LIST THE TWO. conditions-which require you toimanually scram the reactor.during a low CRD' scram air header pressure casualty? (0.75)
- b. What are the operator immediate actions for a: loss of feedwater heating? (1.00)
, c. One RFP trips and'the Reactor Recirculation FCV begin to' runback.
l What ere the operators immediate actions with; regard-to the Reactor
- Recirculation runback only? ASSUME THAT NO SCRAM OCCURS. (1.00)-
I ,
i j-t:
i' i
l -(***** . CATEGORY 04 CONTINUED'ON NEXT PAGE *****)-
. . . - ,. . - - ,- - , . - , . . _ . - . a ..- - . ~ . . - . , . . - , . . . . . - - , - . . - , , . _
it- EBQGEQUBE1_: NQBd8Lt 8BNQBd6Lt EdEBQENG1 8NQ PAGE 13 BaDIQLQQ198L'_CQUIBQL l
QUESTION 4.05 (2.00)
, n. At what point do you initiate suppression pool cooling if a SRV is stuck open at power? (0.50)
- b.- At what point would a manual scram-be required if the SRV cannot
- be closed ? (1.50)
QUESTION' 4.06 (2.00) j Complete the following chart of.0perational' conditions. (2.00)
L MODE SWITCH AVE. REACTOR CONDITION POSITION COOLANT TEMP.
3 4
- 1. . Power Operation Run Any Temperature
- 2. Startup (a.) (b.)
- 3. Hot Shutdown (c.) (d.) <
1 4. Cold Shutdown (e.) (f.)
t- 5. Refueling (g.) Ch.)
i r
QUESTION '4.07 (2.50) j
- a. State the quarterly whole body- administrative exposure -limits ,
assigned to the workers at. River Bend Station. (1.00)
} 'b. List'the-radiation levels that require the following special
-precautions:
i, l. Posted as a radiation area.- (0.50)
- 2. Posted as a high radiat' ion area, RWP required for entry,
~
and must.be~1ocked if no access watch sat. (0.50)
- 3. Posted as.a HOT SPOT. (0.50) t 1
i 4
i i
- C***** CATEGORY 04 CONTINUED:0N NEXT-PAGE *****)1
_ _ . . . - . . - . . . - . _ - . _ . . . _ , - - . _ , - . . . , . . _ . , . - _ . - , . ~ . . . , _ . - . . - _ . . . . - .
i -.
f 4 t__EBQQEQuBES_:_NQBdakt_6BUQBd8Lt_EDEBQENQ1_6NQ PAGE 14 BaQIQLQQ1 gal _QQUIBQL QUESTION 4.08 (2.00) 7 You are the reactor operator during a reactor startup~. You have j ust
~
completed pulling a rod to its full out position. .You perform a rod coupling check.
- a. LIST THREE indications would you see to tell you that the rod is uncoupled. (1.50)
- b. What would be your immediate actions? (0.50) t QUESTION 4.09 (2.00) i' You are the REACTOR OPERATOR during a reactor start up in accordance with GOP-0001. The reactor has just been made critical. LIST FOUR OF
, THE FIVE items which need to be recorded in the Control Room Log and in the copy.of GOP-0001. (2.00)
-QUESTION 4.10 (2.00)
- a. With the facility at power,-what actions must be taken from-the control room before evacuating.it because of a flash fire? (1.0) 1 i b. Why is.the' operator required to deenergize circuits at-panels IC71-PNLP001 and IC71-PNLP002, located on the 115'ft. level of the
. control building, after leaving the control room? (1.0)
' QUESTION 4511 L C1~25)
AnTirradiated component measuring 6"x'6"x'la" is1 surveyed with a portable instrument. At a distance of 2 feet, the open. window reading is 2.65 Rem /hr and the closed window. reading is 1.75 Rem /hr.. SHOW ALL WORKl
- n. What is the beta dose rate? (0.'50)
- b. What would the WHOLE BODY dose rate be if yo'u moved back- to a distance 10 ft, from the component ? (0.75) t
(***** END OF CATEGORY 04 *****)-
(************* END OF EXAMINATION ***************)
Iz__EBINDIELE1_QE_UUQLE68_EQWEB_ELaNI_QEEB811QUt PAGE 15 IEEBdQQ188010Ht_ Heal _IBaNSEEB_aNQ_EL91D_ELQW ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 1.01 (3.00)
- a. As the reactor operates during the early part of the cycle, the burnable poison depletes more rapidly than the fuel, therefore, control rods must be inse~rted to hold the power constant. (1.50)
- b. As the control rod density increases, the power producing regions of the core become more undermoderated; as total power production has remained constant but power producing volume has become smaller.
Because of this effect, the void coefficient becomes more negative.
(1.50)
REFERENCE RB THEORY HANDOUT ANSWER 1.02 (2.00)
- c. decrease (0.50)
- b. increase (0.50)
- c. decrease (0.50)
- d. decrease (0.50)
REFERENCE RB SIMULATOR MALFUNCTON LIST ANSWER 1.03 (2.00)
- c. Xenon will peak at its highest reactivity approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the scram. The worth of the rods will be relatively lower than the what they would be worth on a xenon free startup. Thus the reactor will go critical at a lower rod density than before. (1.25)
OR Peripheral rods may exhibit higher than normal notch worth,
- b. The rods will need to be inserted after startup to control power.
Thus rod density will increase. (0.75)
REFERENCE RB THEORY HANDOUTS
Iz__EBIUCIELES_QE_NUCLEaB_EQWEB_EL8HI_QEEBoI1QNt PAGE 16 IBEBdQQIdadlCSt_BE8I_IB8HSEEB_aNQ_ELylQ_ELQW ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 1.04 (2.00)
Using pump laws, since speed increased by a factor of two, then capacity increased by a factor of two, pump head by a factor of four, and power required by the pump a factor of eight. The final values will be:
Capacity = 1000 gal / min (0.66)
Pump Head =100 psi (0.66)
Power required = 800 KW (0.66)
REFERENCE RB THEORY HANDOUTS ANSWER 1.05 (2.50)
- c. Doubling time equals 50/1.44 = 34.7 seconds. (1.00)
- b. 40% range 2.is equal to 0.04% on range 8 P(0) = 0.04 P(t) = 40 Period = 50 seconds P(t) = P(0) e ^(t/ period) 40 = 0.04 e ^(t/50 sec)
Time = 345.4 seconds or 5 min. 45.4 see (1.50)
REFERENCE RB THEORY HANDOUT ANSWER 1.06 (1.00)
- a. Increase (0.25)
- b. Increase (0.25)
- c. Increase- (0.25)
- d. Increase (0.25)
REFERENCE RB THE0RY' HANDOUT
11-__EBINQ1ELES_QE_NUQLE88_EQWEB_ELANI_QEEBoI1QUt PAGE 17 IBEBdQQ1NedlGSt_BE61_IB86SEEB_6NQ_ELUIQ ELQW ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 1.07 (2.00)
FAILURE MECHANISM LIMITING CONDITION
- a. LHGR 2 1
- b. APLHGR 3 3
- c. MCPR 1 2 (6 ANSWERS AT 0.33 EACH)
REFERENCE RB THEORY HANDOUTS ANSWER 1.08. (2.00)
- a. p>/= beta (0.005< beta (or bar-eff)<0.007) (1.00)
- b. The fraction changes over core life as more PU-239 is used for.
power. Since beta (U-235) > beta CPU-239), the overall-beta decreases (to 0.0054) (1.00)
REFERENCE RB THEORY HANDOUT ANSWER 1.09 (2.00)
- a. Plant efficiency would decrease. (0.75)
- b. All condensate /feedwater heat that is rejected to the Cire_ water system must be added by the reactor. (1.25)
REFERENCE RB THEORY HANDOUTS ANSWER 1.10 (2.00)
- c. True (0.50)
- b. False (0.50)
- c. False (0.50)
- d. True. (0.50)
i- . .
11__EBINQ1ELES_DE_NUCLEeB_EQWEB.lL8HI_QEEB8I1QNt PAGE 18 IBEBdQQ1Ned1 cst _BEoI_IB6NSEEB_6NQ_ELu1D_ELQW 4
ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A. 1
]
1
- REFERENCE GE Thermodynamics, Heat Transfer and Fluid Flow k
. ANSWER 1.11 -(2.50)
As power' increases the amount of boiling'(two-phase flow) increases.
The? boiling will be the greatest in the core center due to it being the region of highest power. Two-phase-flow restricts 4
cooling water' flow due to the boiling action. This will cause
'the higher powered bundles to receive less. cooling water since
-their-higher resistance to flow sill divert flow to~1ower power fuel bundles, starving-the higher power bundles. (2.50)
REFERENCE
! RB THEORY HANDOUT ANSWER 1.12 (2.00)
This could be accomplished by withdrawing the already inserted detectors and observing the count rate. When the detector reaches the water level the count rate should significantly change due to more moderation of fast neutrons. Knowing the detector withdrawal- ,
speed and the time it takes.for count rate to change, the approximate water level can be determin d. (2.00) 4 REFERENCE RB MITIGATING' CORE-DAMAGE-4
(
4
.r= - y ae- , - % y g ,.,w . . , . . , , - , ,,- ,,, ,w . = -.-..rg---i-- ,..*,eg- ym . ---e..y -,.
4 . - , , v.-.#.
21__EL6NI_QESIGN_INGLuRINQ_SoEEIl_680._EdEBGENGl_SYSIEd1 PAGE 19 ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 2.01 (2.50)
- n. Power / load unbalance, 1800 RPM not selected, loss of stator cooling, manual runback by lowering the load limit pot., or high level in the 5th or 6th stage heaters. (3/4 0 0.333 ea)
- b. To prevent generator overheating (at various hydrogen concentrations and generator loads.) (0.50)
- c. Higher thermal conductivity (2 at 0.500 ea)
Lower density = > less windage losses Less contamination which could lead to insulation breakdown.
Less Oxygen present which lowers the oxidation rate.
REFERENCE RB TRAINING TEXT CHAP-36,38 ANSWER 2.02 (2.00)
- n. To prevent a significant decrease in reactor level which can occur when the valves are opened. (1.00)
- b. If not, the RHR pump minimum flow valve to the suppression pool will open, dumping reactor water to the suppression pool. (1.00)
REFERENCE SOP - 0031 ANSWER 2.03 (2.50)
- 1. Manual (0.50)
- 2. ADS signal (0.50)
- 3. Relief set pressure (0.50)
- 4. Spring set pressure (0.50)
- 5. Low-Low set pressure (0.50)
REFERENCE RB TRAINING TEXT CHAP 34
2t__ELeUI_QESIGU_INGLMQINQ_SeEE11_eUQ_EUEBGEUGl_11SIEUS PAGE 20 ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
l l
l ANSWER 2.04 (2.00)
This delay is used tn delay the flux signal for a short period of time to prevent any spurous signal or transient from causing an unnecessary scram. (2.00) l OR To. simulate fuel thermal transfer transient characteristics.
REFERENCE RB TRAINING TEXT CHAP 19 RB TS, p. B2-7 ANSWER 2.05 (2.00)
The exhaust water from the HCU's which are conducting rod motion enters the exhaust header and creates a reverse flow condition in the header.
This flow exits the header into the HCO's not conducting rod motion via the reverse flow capabilities of the EP121 valves (one of the directional control valves). This valve opens at 3 psid and the. exhaust water flows into.the CRDM cooling line and into the reactor vessel.
(2.00)
REFERENCE RB TRAINING TEXT CHAP 6 ANSWER 2.06 (3.00)
- u. 70 spm (0.50)
- b. To provide sufficient cooling for pump internals. (0.50)
- c. -- Manual.
-- Electrical fault
-- RPCCW Seal cooling outlet temp high -- 165F
-- RWCU suction valves not full open (F001/004)
-- Low pump flow (15 second TD) -- 70 gpm
-- Any system isolation (any 4 for 0.50 ea)
REFERENCE RB TRAINING TEXT CHAP 9
It_~EL8Ul_QE11QN_INQEyQ1NQ_16EE11_6NQ_EMEBQENQ1_111IEt.S PAGE 21 ANSWERS -- RIVER BEND 1 ,
-85/07/29-MENDIOLA, A.
ANSWER 2.07 (3.00)
Alt. Source
- Station Batt* Primary Source *
(480vac 1ph) (125vdc) (480vec 3ph) i I I BKR BKR BKR
(~~~~~g-~~~~~~) ALT {~~~l~~~~~~~~~~~~~~~~~~~~~~~~~l~~~~~~~)
{ .480/120V )SRC ', BKR BKR }
{ REG XFORM* )XFORM* { l I )
{ l ) { DIODE (>l)* RECTIFIER * )
{ _________________
( i I )
{ i } l ( ________________-_-_______I }
{- D ) I { SYNC CHK----------INVERTER * )
{ l } I { l I )
{ MAN BYP------- --------D---------STATIC---------- )
{ SWITCH * ) I { SWITCH * )
{ l ) I { l )UNINT.
~~~~~l~~~~~~~ -------------D--- ---------
)PS*
I ( )
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
(LOAD)
(11
- answers e 0.2 ea., 9 other answers 0 0.05 ea., 0.35 for wiring / layout)
REFERENCE RBS AC ELECTRICAL DISTRIBUTION SYSTEM, p. 44-16, Figure 6 ANSWER 2.08 (2.50)
- b. The RCIC gland seal system provides air to the turbine seals (1.0) at slightly higher than turbine exhaust pressure.
- c. Overspeed (0.5)
REFERENCE River Bend Simulator. Text, Chapter 11, RCIC, pg 11-9,11,12,13,14
2___ELaNI_QE11GN_INGLVQ1NQ_16EEI1_aNQ_EMEBEEUGl_1111EUS PAGE 22 ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 2.09 (2.00)
- n. Normal Service Water is the normal heat sink (0.25). Standby Service Water is the alternate heat sink (0.25).
- b. 1. At 100 ps ig, the standby RPCCW pump (s) start (0.5)
- 2. At 56 psig, the RPCCW Division I and II normal supply and return valves shut (0.5), and the Standby Service Water supply and return valves open (0.5) to supply Standby Service Water directly into the RPCCW system.
REFERENCE River Bend Simulator Text, Chapter 48, RPCCW, pg 48-4,5 ANSWER 2.10 (3.50)
- n. Both normally deenergized (0,5)
- b. SBV: Actuation of the fast-acting solenoid causes a fast opening of the SBV (0.5) to reduce the pressure transient from a closure of the main stop or control valves (0.5).
TCV: Actuation of the fast-acting solenoid causes a fast closing of the TCV (0.5) to minimize the turbine overspeed on a loss of generator load (0.5).
- c. 1. shut (0.5)
- 2. shut (0.5) 4 REFERENCE River Bend Simulator Text, Chapter 36, EHC/ Steam Bypass and Pressure Regulating Systems,'pg 36-16,33 l
I l
l w.
2___IN11BudEN11_6NQ_CQNIBQL1 PAGE 23 ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A. .
4 l
l ANSWER 3.01 (3.00)
, n. Detector fails low, system assumes a power decrease and reduces feed. flow accordingly. A steam flow / feed flow mismatch will cause
-level to fall and create a level error signal. This signal will eventually offset the flow error and level will stabilize at i a lower. level than the original level. (1.00) i
- b. Detector fails low. Relays in the master level controller open the input to the operator selected setpoint and inserts a l'
preprogrammed setpoint. This setpoint is a higher setpoint and is ,
called for immediately. Actual level will begin to increase if
, the operator does not change level indication to the operating detector and manually reset the setdown circuit. (2.00)
I 4
REFERENCE '
RB TRAINING TEXT CHAP 33 j
! ANSWER 3.02 (3.00)
- s. 1. INCOMPLETE START SEQUENCE RELAY NOT ACTUATED ,
- 2. CB-5. FULLY RACKED IN
- 3. FCV IN MANUAL MODE AT MINIMUM FLOW POSITION
- 4. SUCTION AND DISCHARGE VALVES > 90% OPEN
- 5. VESSEL THERMAL SHOCK INTERLOCK PERMISSIVES SATISIFIED i- (0.30 each) i b. 1.-Feedwater flow (0.50)
- 2. The feedflow requirement is~to prevent cavitation of the-RR FCVC1.00)
REFERENCE l RB TRAINING' TEXT CHAP 8 i
i i h l
. . - . _ _ _ _ ......__...n- .-..,_..___,______..___._.._,___.,,.a....,. _,_.u_,_, -
2t__IN118MdEN11 8NQ_QQN189L1 PAGE 24
-ANSWERS -- RIVER' BEND 1 -85/07/29-MENDIOLA, A.
-ANSWER 3.03 (3.00)-
BLOCK SETPOINT BYPASS
- a. ROD PATTERN Sequence violation >27.5 +/-3 %
CONTROL
- b. SDV HI LVL. 32.5" Key switch _in " bypass"and in Shutdo'n w or Refuel mode
- c. Rod Position Data fault Never Information
- d. Gang Misaligned Gang rods at . Individual mode different positions _
- o. Reactor ~ mode In SHUT 00WN Not in SHUTDOWN switch (0.20 each)
. REFERENCE RB TRAINING TEXT CHAP 21 ANSWER 3.04 -(1.50)
- 1. ADS logic is' manually ~ reset.by the operator.
- 2. Shutdown of all-low pressure ECCS pumps.
3.= ADSLinhibit-switches actuated.
Cany.2/3 0 0.75 ea.)
REFERENCE RB TRAINING TEXT-CHAP;34
. ANSWER i3.05 (2.003'
- o. False
- b. False.
ac . False
- d. True (0.50Eeach)
REFERENCE RB TRAINING TEXT CHAP 19 I
L l_
2t__IU118UMENI1_aND_GQNIBQL1 PAGE 25 ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 3.06 (3.25)
- c. -- NARROW 0 - 60 "
-- UPSET 0- 180 "
-- WIDE -160 - 60 "
-- SHUTDOWN 0 - 400 "
-- FUEL ZONE -150 - 50 " (0.15 each part)
- b. NARROW, WIDE, AND UPSET (0.25 each)
- c. SHUTDOWN AND UPSET (0.50 each)
REFERENCE RB TRAINING TEXT CHAP 3 ANSWER 3.07 (2.75)
- o. -- RCIC ISOLATION (0.25)
- b. -- RHR ISOLATION (SHUTDOWN COOLING MODE) (0.25)
- d. -- SCRAM SHIFT RR FLOW TO SLOW SPEED ADS CONTRIBUTING SIGNAL CLOSE RHR SHUTDOWN ISOLATION VALVES FWLCS SETPOINT SETDOWN (5 AT 0.20 each)
- e. -- INITIATE RHR PUMPS AND LPCS ADS CONTRIBUTING SIGNAL START DIV 1 AND 2 DIESEL GENERATORS CLOSE MSIV's (4 at 0.25 each)
REFERENCE RB TRAINING TEXT CHAP 3 ANSWER 3.08 (1.50)
- c. Total core flow is NOT accurate (0.25)
Running loop flow is accurate . (0.25)
-b. The CORE FLOW SUMMING NETWORK will subtract the non-running loop flow from the running loop flow even though reverse flow does not exist in the non-operating loop.(0.75). Running loop flow would not be affected (0.25).
REFERENCE RB TRAINING TEXT CHAP 3
1 21__IN11890ENI1_oND_GQUIBQL1 PAGE 26
. ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 3.09 (2.00)
- 1. RCIS has detected a data fault in rod position indication. (0.50)
- 2. Indicates that all scram valves are not in the same state. (0.50)
- 3. Indicates that the RGDS finds disagreement between the signals from the two RACS. (0.50)
- 4. Indicates the selected rod must by fully inserted before any other -od can be moved. (0.50)
REFERENCE RB TRAINING TEXT CHAP 21 ANSWER 3.10 (3.00)
- c. 1. MAINTENANCE -- diesel will not start in this mode (0.50)
- 2. OPERATIONAL -- Local manual start
-- LOCA signal (2# or LEVEL 1)
-- Sustained bus undervoltage
-- Emergency start pushbutton (local and remote)
(4 at 0.25 each)
- b. -- Engine overspeed
-- Generator differential trip (2 at 0.75 each)
REFERENCE RB TRAINING TEXT CHAP 45
4 dz__BBQQEQMBES_:_UQBd8Lt_6QUQBU6Lt_EdEBQEUQ1_6UQ PAGE 27 88010LQQlCeL_CQUIBQL ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
, ANSWER 4.01 (3.50)
- 1. Supression pool temperature > 95 F
- 2. Supression pool level outside 19.5' - 20.0'
- 3. Containment temperature > 90 F
- 5. Drywell temperature > 145 F
( 0.45 for each condition and 0.25 for each setpoint)
REFERENCE E0P 0001/0002 ANSWER 4.02 (2.50)
In order to place an ECCS system in manual ~ mode, one must verify by at least two independent i n'd i c at io ns that:
-- Misoperation in automatic mode is confirmed, OR
-- Adequate core _ cooling is assured. (2.50)
REFERENCE E0P 0001 ANSWER 4.03 (2.50)
- c. RWCU blowdown (1.00)
- b. Verify the following:
- 1. All APRM ONSC indicator lights are extinguished, all APRM's read > 3% power and the APRM ONSC annunciator-is clear.
(0.75)
REFERENCE GOP 0001 l~
. _ . _= , - .
4 1__EBQQEQUBES_:_NQBUALt_6BNQBU6Lt_Edg8EEUQ1_68Q PAGE 28 88Q10LQQ1 col _QQN18QL ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 4.04 (2.75)
- a. When the second individual control rod scrams or it is determined that instrument air cannot be restored. (0.75)
- b. -- Reduce RR flow as required to reduce reactor power by 20%.
-- If a scram occurs, refer to AOP - 0001, Rx scram.
-- Notify the shift supervisor. .
(0.33 ea)
- c. -- Transfer the FCV's to LOOP MANUAL and reduce flow to establish < 50% power. Balance loop flows to within 10% at
< 70% core flow. (1.00)
REFERENCE AOP - 0008,0007,0024 1
ANSWER 4.05 (2.00)
- a. When suppression pool temperature exceeds 90 F (0.50)
- b. If the SRV cannot be closed withing two minutes from the first attempt to re-close or if the suppression pool temperature reaches 100 F. (0.75 ea)
REFERENCE AOP - 0035 ANSWER 4.06 (2.00)
, n. Startup/ Hot Standby
- b. Any Temperature
- c. Shutdown d.'>200 degrees F
- e. Shutdown
- f. < or EQ 200 degrees F
- g. Refuel % Shutdown (BOTH REQ'D)
- h. < or EQ 140 degrees F (0.25 each)
! REFERENCE RB ~TS i
b i
1 ., - -, . - . - , . , - - . _ - - . , - , -
at__EBQQEQUBEl_: NQBdakt_eBNQBdabt EdEBQEUQ1_6NQ PAGE 29 B6010LQQ1GoL_QQNIBQL ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 4.07 (2.50)
- n. 1000 mrem /qtr 2000 mrem /qtr with NRC - FORM 4 (and lifetime <5(n-18)) (1.00)
- b. 1. >2.5 mrem /hr
- 2. 100 - 1000 mr/hr
- 3. Spot that is 5X > BKGD (0.50 ea)
REFERENCE RB GET II HANDOUT ANSWER 4.08 (2.00)
- c. -- Rod overtravel alarm, loss of the position 48 indicators, and the loss of the FULL OUT lights. (0.50 each)
- b. -- Notify the Shift Supervisor (0.50)
REFERENCE AOP - 0030 ANSWER 4.09 (2.00)
-- Time of criticality Rod number and position at criticality
-- Reactor temperature
-- Period
-- Operator name (4 required at 0.5 each)
REFERENCE GOP - 0001
= .
i 4t__EBQQEQuBES_:_NQBdakt_6BNQBd6Lt_EUEBQENQ1_6NQ PAGE 30 B8DIQLQQ1QaL_QQUIBQL ANSWERS -- RIVER BEND 1 -85/07/29-MENDIOLA, A.
ANSWER 4.10 (2.00)
- a. 1. Manually scram.
- 2. Verify rods in.
- 3. Mode sw. to shutdown.
- 5. Trip turbine. (5 answers e 0.2 ea.)
- b. To ensure that the reactor is shutdown and the MSIV's are shut. (1.0) 4 REFERENCE AOP-0031,.p. 4 ANSWER 4.11 (1.25)
- a. Beta dose rate = open - closed window = 0.9 Rem /hr (0.5)
- b. 1.75 x (2 ft.)^2 = Dose x (10 ft.)^2 Dose = 1.75 x 4 / 100 = 0.07 Rem /hr.
(relation w/ dist 0.5, math 0.25)
REFERENCE GET II,'p.-5, 6, 36
, , -- - -- +- -< n-e- -
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _BlVEB_QEUQ_1____________
REACTOR TYPE: _BWB-QEh_________________
DATE ADMINISTERED: _Q54QZl2R________________
EXAMINER: _EELLElt_Jz______________
APPLICANT: _________________________
INSIBUGI1QU5_IQ_6EELIQaNIl Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade o f- at loest'80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY
__YeLUE_ _IQIeL ___5GQBE___ _V8LUE__ ______________GoIEQQBY
_25tQQ__ _25tDQ ___________ ________ S. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_25tDQ__ _25tQQ ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION-
_251Q0__ _25tDQ ___________ ________ 7. . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_25tDQ__ _25tQQ ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 1DQtDQ__ 10Q100 ___________ ________ TOTALS FINAL GRADE _________________t All work done on this examination is my own. I have neither given nor received aid.
APPLICANT'S SIGNATURE e
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l St__IBE981_QE_NUGLE6B_EQWEB_EL6NI_9EEB6119Nt_ELu1QS&_6NQ PAGE 2
- IBEBdQQ1N6M191 i
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. QUESTION 5.01 (1.50) l i Why is the River Bend reactor always operated with an undermoderated core?
i, i QUESTION 5.02 (2.50) 1 How will each of the errors below affect the ability of a 1/M plot to predict criticality? Which error (s). result in an UNSAFE prediction?-
- a. The core is loaded toward the detector.
j b. The detector is located too far from the core.
- c. The detector is located too close to the source.
! d. The detector ~is located too far from the source.
t i
j QUESTION 5.03 (3.00)
I a. What is meant by the term." prompt cr'itical"? (0.75)
- b. What is the minimum reactivity that must be added to a critical j reactor for it to be prompt. critical? (0.75)
- c. How much (by what factor) would1 power increase in one second in a prompt critical reactor? (1.5) 1 i
! QUESTION 5.04 (2.'00) l 4
l Reactor power is increased by withdrawing rods .until void fraction goes t up 1.5%, fuel temperature goes up 40 degrees F, and moderator temperature goes-up 1 degree F. What is the worth of the control rod withdrawal?
Ignore Xenon effects.
r QUESTION. 5.05 (1.50) l t
j How and why does moderator temperature coefficient change with' core age?
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Et__IBEQBl_QE_NVGLE68_EQWEB_EL881_QEEBoI1QNt_ELVIQ1t_6NQ PAGE 3 IBEBdQQ1NadIQ1 i
l QUESTION 5.06 (2.00) 5 Match the blanks in the paragraph.on rod worth with the appropriate choice from-the list below. Each choice may be used more than once or not at all.
The distance over which a control rod is effective is roughly equal to, the __a__ in the reactor. The overall worth of the rods __b__ equal to the sum of the worth of each rod. The overall rod-worth.is a function of rod __c__ and __d__. A(n) __e__ rod-worth curve shows the change in rod-worth per unit length. A(n) __f__ rod-worth curve represents the slope of a(n) __g__ rod-worth curve. A(n) __h__ curve shows the-total reactivity. introduced by placing a rod at a given position.
1 1. integral 5. is 9. location
! 2. differential 6. is not 10.- reactivity excess 4
- 3. amount of fuel 7. diffusion length
- 4. number of rods B. spacing i QUESTION 5.07 (3.00) ,
- s. How long does it take to reach xenon AND samarium peaks-after a.
! reactor trip from 100% power? Assume at power lorig enough to
reach equilibrium values for both Xe and Sm. .
-(1.0)
- b. What is meant when a startup is said to be "xeno'n-precluded"?
. (0.75)
- c. Under what plant conditions is a' xenon-precluded startup most likely to occur? . .
.; (0.75)
- d. What is the longest period a Xe-precluded condition can exist? (0.5)
I.
QUESTION 5.08 (2.00') ,
[ What is the difference in CAVITATION and GAS BINDING.in a centrifugal pump?
i i
4 QUESTION 5.09 (2.00) .;
MATCHING. Match the appropriate phrase to each of the heat transfer regimes. MORE THAN ONE ANSWER MAY BE CORRECT AND-REQUIRED.
, '___ a. Single Phase Forced .1. Void fraction is significant.
Convection.. 2. . Coolant remains single phase. i
___ b. Subcooled Boiling. 3. -Occurs at bottom of. coolant. channel.
- ___.c. Bubble Flow. 4. Very small void fraction.
.___ d. Slug Flow.- 5. Quality is low.
- 6. Bulk coolant temperature increases to saturation.
- 7. Bulk boiling occurring. i i (***** CATEGORY 05' CONTINUED ON NEXT PAGE *****)
Ez__IBEQBl_QE_NMGLE88_E9 WEB _EL8N1_QEEB81198t_EL91Q$t_8NQ PAGE 4 IBEBdQQ1NedICS
,- QUESTION 5.10 (3.00)
- a. How does the pressure drop.across the core inlet orifice compare with the-pressure drop across the core at 100% power and flow? (1.0)
(Actual values or factor is required for-full credit.)
- b. Why are core' inlet orifices installed? ( 1. 0 ) .
- c. How do the core inlet orifices achieve their function? (1.0)
QUESTION 5.11 (2.00)
.a. What is " CRITICAL POWER"? (0.8)
.b. For EACH of the following cone.itions, how is cr.itical power effected?
Assume all other condit. ions ar e :onstant. Answer for each part individually. No~ explanation is required.
(1.2)
- 1. Decrease in reactor presourc.-
- 11. Increase in' inlet subcooling, iii. Increase in core flow, iv. Location of axial power moves.up in the core.
\
s QUESTION 5.12 ( .50)
- a. TRUE or-FALSE? APLHGR is a function of fuel burnup. (0.25)
- b. FILL IN THE BLANKS. The APLHGR. limit is based on keeping __(i)__
below __(11)__ degrees F. (0.25) a -
f
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(***** END OF CATEGORY 05 *****)~
u .s Ez__ELeNI_111IEd1_QE11 Gut _QQUIBQLt_8NQ_1811BudENI611QN PAGE 5 t_ -
i QUESTION 6.01~ (1.50)
What two-(2) mechanisms or methods are used in the Offgas system to maintain or reduce-hydrogen concentration of-the offges below 4%?
e QUESTION 6.02 (3.00)
Sketch one Standby 120V A-C Uninterruptible Power Supply System. Include j in-your answer all; breakers or disconnects, power _ sources, switches, and 4
other active components. Label all subsystems and devices shown.
QUESTION 6.03 (2.00)
'What signals will automatically start or in it iate the Standby Service Water System (SWP)? Include in your answer _the setpoint and logic for each initiating signal. (2 signals required for full cre~dit.)
f QUESTION 6.04- (3.00)
The reactor is operating-at-75% power while an MSIV surveillance is performed. Consider each---of the following cases separately.
~
switch is placed to CLOSE. How does the MSIV-respond (open fast, close fast, or no change)? _
. . (0.5)
- switch is placed to-OPEN. How does the MSIV respond (open fast,
- close fast, or no change)? (0.5)
, c. Once the MSIV is shut during testing, how is it reopened? (0,5)-
- d. What-RPS signal (s) is(are)~ generated ,'directly or indirectly by the MSIV testing above? (0.5)
-e. How and why would plant operation have baen;affected if the reactor had been at 90% power when the testing res-otar.ted? (1.0)~
4 QUESTION 6.05 '(3.00)/
-c. . When (under whatcplant~ conditions) is' Primary' Containment Integrity '
required to be ma'intained? (1.0)
~
- b. What;ste four of the six--conditions which must be met'for Primary Containment Integrity to exist?- ( 2. 0 3-1 1
L
- )
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t ht__EL8NI_SISIEdS_DESIGNi_CQNIBQLt_6ND_INSIBudENIeIlQN PAGE 6 i.
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QUESTION 6.06 (2.00)
'a. What happens to the reactor wnter-level setpoint if reactor water I
level decreases below Level 3 (+10") with the Feedwater Control s
, ~ System in automLtic? (1.0)
- b. :What operator action (s) is(are) required-to raset this' action? (0.5) c.' What problem'does.the action in (a.) help prevent? (0.5) i I
- QUESTION' 6.07 (2.00)
- Each condition below may indicate failure of one or both' Reactor Recirculation Pump' Seals.- State whether each. case indicates' failure of the upper. seal,.-lower seal, bota seals, or-neither seal.
a.- -Lower seal cavity pressure decreases.-
.b. Upper. seal cavity pressure-decreases.
- c. Upper seal cavity pressure increases'.
- d.- Lower seal cavity pressure increases.
QUESTION 6.08 ( 3 .' 0 0 ) -
)' List five (5) ofEthe six (6) modes of operation of-the Residual. Heat Removal (RHR)-System,'the purpose of each, and the number of loops available for'esch function.
1 d
QUESTION. 6.09 '(2.'00) i Fill in the, blanks in the-paragraph below concerning'the' Average Power
- Range Monitoring (APRM) Syste'm. Blanks may contain one or more words or. numbers..-
.The APRM ' system consists of _(a)_ identical channels. Each channel ~ is provided with _(b)_ inputs from _(c)_. There are _(d)_ trip units
- associated with each APRM channel.-They.-are _(e)_, _(f)_, _(g)_, _(h)_,
and'the UPSCALE THERMAL POWER TRIP,.which.provides a scram signal to the RPSLwhen _(i)_.
' QUESTION 6.10 C . . 50 3 '
f' ,
i TRUE or FALSE? The operator can manually isolate RCIC at any time by
]. initiating a division 1-isolation at panel H13-P601.
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?i . . , _ .- .. . a- 2 - _,..._.._,;_----_..,w-___ _ .. ._, , _ ,. _ u .....
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ht__EL6NI_111IEUS_DE11GNt_G9NIBQLt_6NQ_IN11BydEUlel1QU -PAGE 7 QUESTION 6.11 (2.00)
MATCHING. Match the appropriate reactor-vessel level setpoints to the automatic actions which it initiates. Each level setpoint may be used more.than once or'not at all. If more than 1 level setpoint can initiate en action, all correct answers are required for full credit.
_____ ACTION _______________________________ __ LEVEL SETPOINT___
___ a. Scram. 1. Level 1 (-144")
___ b. Isolate RHR. 2. Level 2 ( -30")
___ c. Initiate RCIC. 3. Level 3 ( +10")-
___ d. Isolate RCIC. 4. Level 4 ( +33")
___ e. Initiate LPCI. 5. Level 5 ( +37")
___ f. Trip Reactor Feedwater Pumps. 6. ---------------
___ g. Initiate HPCS. 7. Level 7 ( +41")
___ h. Start Division 1 Diesel Generator. 8. Level 8 ( +57")
___ i. Initiate Standby Gas Treatment System.
QUESTION 6.12 (1.00)
What are four control room indications available to the operator which verify normal RCIC operation after initiation?
J A
1
(***** END OF CATEGORY 06 *****) ,
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I QUESTION 7.01 (3.00) a.- Describe how impurities (air and solids) are removed from coolant in the Reactor Pressure Vessel (RPV), prior to.startup. (2.0) b.- Describe how RPV level is controlled during initial heatup. (1.0)
QUESTION 7.02 (1.50)
- a. What condition can exist during a reactor restart (startup) which could'eause unusually high control rod notch worths? (0.5) b.. What notch positions are particularly vulnerable to this effect? (0.5)
, c. If'such a condition exists, where in the-startup is additional
- caution required? (0,5)
. 1 I
( QUESTION 7.03 (2.50) 1
- a. While increasing power, what are two ways to verify rod coupling? (1.0)
- b. When increasing power to 100% by increasing recirculation flow in j- MASTER MANUAL, what three operating maps (or+11mits) govern the 4
power and rate of change of power? (1.5)
QUESTION 7.04 (3.50) l' An irradiated component measuring 6"x 6"x 6" is. surveyed with a portable l instrument. At.a distance of 2 feet, the open window ~ reading is 2.65
- . Ram /hr and the closed window reading is11.75 Rem /hr. SHOW ALL WORKl
- ' a. What is the beta dose rate? (0.5)
.b. How long could you remain at the survey distance without exceeding
, administrative and 10CFR20 quarterly limits? (1.0) c.- What are. guidelines ~that will allow exceeding these limits when
- time is not available for adequate planning? . (1.0)
- d. . What would the WHOLE BODY' dose rate be if you-moved back to a i distance 10 ft. from the^ component ? (1.0) i l '. >
(
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(*****
j , CATEGORY'0'7 CONTINUED ON NEXT PAGE *****)
Zz__EBQGgDUBgS_:_NQBeatt_agugBuatt_gegBhgggy_aNQ PAGE 9 88Q10LQE1 cal _GQUIBQL QUESTION 7.05 (2.50)
- c. Other than the interim storage area, list 3 areas where new fuel may be stored? Indicate if the fuel would be stored wet or dry. (1.5)
- b. List two places where irradiated fuel may be stored AND any restrictions on when each may be used for such storage. (1.0)
QUESTION 7.06 (3.00)
- c. Describe how the RPV pressure is maintained after a reactor scram when the Main Steam Isolation Valves (MSIV's) are:
-1. OPEN. (1.0)
- 11. CLOSED. (1.0)
- b. What two actions must be completed after a reactor screm before a reactor startup may be initiated? (1.0)
QUESTION 7.07 (2.50)
- a. E0P-0001 (RPV Control) can be entered if any one of six different conditions exist. List four of the entry conditions. Values or setpoints are NOT required. (1.0)
- b. E0P-0001 requires the operator to take initial sction to control what three parameters? ~
(0.75)
- c. Briefly describe the three methods and general flow paths which can be used to inj ect boron into the RPV. (0.75)
QUESTION 7.08 (2.50)
- n. When would " Alternate Shutdown Cooling" (AS00) be used? (0.5)
- b. Describe the ASDC heat sink and flow path (s). (1.25)
- c. What are three temperature indications that can be used to monitor cooldown rate during alternate cooldown? (0.75)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
Zz__EBQQEQUBE1_:_NQBd8L&_eRUQBdaL&_EMEBQEUQ1_6NQ PAGE 10 BaQ10LQQlCaL_QQUIBQL QUESTION 7.09 (2.00)
- n. With the facility at power, what actions must be taken from the control room before evacuating it because of a flash fire? (1.0)
- b. Why is the operator required to-deenergize circuits at panels IC71-PNLP001.and IC71-PNLP002, located on the 115 ft. level of the control building, after leaving the control room? (1.0)
QUESTION 7.10 (2.00)
- c. What are three of the six conditions that require " Emergency De-pressurization"? (1.0)
- b. What are two of the four conditions which require RPV flooding? (1.0) i
(***** END OF CATEGORY 07 *****)
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Hz__8Dd1N11188I1YE_EBQQEQUBESt_QQNQlIl0NS&_8NQ_L1011811QN3 PAGE 11 i
$ QUESTION 8.01 (1.00)
- Explain the-meaning of~the four separate items which make up a report number per the River Bend Report Identification and Tracking system
- (for example, Report Number
- NR- 02 1). (1.0) 4 QUESTION 8.02 (2.00) i j Answer the following regarding allowed working hours and overtime per the River Bend TECHNICAL SPECIFICATIONS.
- o. An individual should not work more than _____ hours straight. (0.5)
- b. There should be at least a(n)' _____ hour break between all
! work periods. . (0.5)
{ c. An individual should not work more than _____ in any seven (7) j day period.- (0.5)
- j. d. Any deviation from the above guidelines shall be authorized j by _____. (0.5) l QUESTION 8.03 (3.00) ,
J 3
- n. Explain the following terms or items as they apply to the River
! Bend Surveillance Test Program:-
j 1. Tolerance period.
! 2: . Status check report.
! 3. Exception report. (1.5) l b. What action must-be taken if a Surveillance Test Procedure step
- cannot be performed as written? (1.5) 1 i
l QUESTION '8.04 (2.'50) i o .- Describe the two conditions or. situations)where it.is'NOT required ;
to comply with the temporary' alterations procedure (A0M-0031). (1.5)
- b. What requirement of procedure ADM-0031 assuresLthat a_ safety- [
related. system or component is CORRECTLY altered and restored. i to normal? (1.0) a i
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QUESTION 8.05 (3.00)
- n. What are three of the four (4)' starting signals that may be used j when performing diesel starting surveillance testing to assure
- Emergency Diesel Generator (EDG). operability? (1.0)
- b. What are four'(4) of the six (6) other verifications _that must
! be observed or performed to verify EDG operability? Setpoints, values, or limits are not required. '(2.0) l QUESTION 8.06 (1.00)
Why is the' Turbine Generator overspeed protection system addressed in River Bend Technical Specifications when the Turbine. Generator is not safety-related? (1.0) s
. QUESTION 8.07- (2.00).
j What'are .three of the four " Leakage Detection Systems" addressed in 1 River Bend Technical Specifications? '(2.0) 1 d
j' QUESTION 8.08 (1.50) i t
What are the requirements on location and number of nuclear detectors required during core alterations? Explain the reason for the
~
i location requirements. (1.5) l i
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Ez__6QUISISIB611YE_EBQQEQUBESt_QQUQlllQNSt_6NQ_(IdlI611QNS PAGE 13 QUESTION 8.09 (2.00)
MATCHING. Match each of the Emergency Responses with the appropriate Emergency Action Level. Responses may be used more than once or not at all. Maximum partial credit will be given for conservative classification.
________________________ EMERGENCY ACTION LEVEL _____________________________
___ a. SRV stuck open at 75% power with increasing suppression pool temp.
___ b. Containment Post-Accident Monitoring system alarms high.
___ c. Total primary leakage exceeds 50 gpm.
___ d. All onsite and offsite AC and onsite DC power lost with no restoration or suppression pool cooling after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
___ e. Unable to restore or maintain reactor vessel level with low level and high drywell pressure indicated.
___ f. ECCS (HPCS) inadvertently. initiates and inj ects water to the RPV.
___ g. Reactor water level indicates below the top of active fuel.
___ h. Both Standby Service Water loops inoperable.
__________________________ EMERGENCY RESPONSE _______________________________
- l. Notification of Unusual Event 3. Site Area Emergency
- 2. . Alert 4. General Emergency QUESTION 8.10 (2.00)
- c. Who is responsible for IMMEDIATE notification of the NRC? (0.5)
- b. Who makes IMMEDIATE notification to the NRC? (0.5)
- c. Who is responsible for FOLLOW-UP notification of the NRC? (0,5)
-d. When does the FOLLOW-UP notif ication commence? (0.5)
QUESTION 8.11 (3.00)
Fill in the blanks regarding the River Bend Safety Limits: (3.0)
THERMAL POWER shall not exceed _(a)_ of RATED THERMAL POWER with the reactor vessel steam dome pressure less than _(b)_ or core flow less than _(c)_ of rated flow.
The Minimum Critical Power Ratio (MCPR) shall not be less than _(d)_
with the reactor vessel steam dome pressure greater than or equal to
_(b)_ and core flow greater than or equal to _(c)_ of rated flow.
The reactor coolant system pressure, as measured in the reactor vessel steam dome, Shall not exceed _(e)_.
The reactor vessel water level shall be above _(f)_.
QUESTION 8.12 (2.00)
List the four (4)' types of evacuations discussed in EIP-2-026,
" Evacuation".
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
Et__IBEQBI_QE_NyGLE68_EQWEB_EL6NI_QEE86I1QNt_ELylQEt_6NQ PAGE 14 IBEBUQQ188dlG1 ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 5.01 (1.50)
An undermoderated core assures-that a power increase causes a decrease in Keff (0.75). This ensures an inherently stable reactor (transients tend to self-damp) (0.75). (CONCEPT)
REFERENCE RBS GP Reactor Theory Student Handout, 17-2 ANSWER 5.02 (2.50)
- e. Overpredicts - UNSAFE
- b. Overpredicts - UNSAFE
- c. Overpredicts - UNSAFE
- d. Underpredicts - SAFE Cover /under 4 ans. e 0.4 ea.; a-c unsafe & d safe 4 ans @ 0.225 ea.)
REFERENCE RBS GP Reactor Theory Student Handout, 22-4 ANSWER 5.03 (3.00)
- c. Critical on prompt neutrons alone. (0.75)
- b. p>/= beta (0.005< beta (or bar-eff)<0.007)
(0.75)
- c. T=1-star /p + (B-p)/tp (0.25) so for prompt critical neglect delayed term & T=1-star /p (0.25) 1-star ~10EE-04 seconds (0.25)
LO.005<p<0.0075 (0.25)
T"10EE-4/0.007~ 0.014 seconds (0.25)
P/Po=eEEtime/T so power would increase ~10EE30~31 in i second (0.25)
REFERENCE RBS GP Resctor Theory Student Handout, 25-3
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j . ANSWERS -- RIVER. BEND 1 .-85/07/29-PELLET, J.
i, i ANSWER 5.04 (2.00) l For R B ', a(voids)~ -1EE-3 dK/K/% voids.
I- a(moderator)~-1EE-4 dK/d. degrees F.
a(dopir)~ -1EE-5 dK/K degrees F.
.So:
.p(voids)= 1.5%*-1EE-3 = -0.0015 dK/K p(dopir)= 40*-1EE-5 = -0.0004 "
p(moder)=~1*-1EE-4 = -0.0001 "
.p(total)= -0.0020 dK/K (3 coeff..e 0.333 ea.: 4 calc. e 0.25 es.)
i REFERENCE c ,
RBS GP Reactor Theory Student-Handout,=26-6 i
ANSWER 5.05 (1.50) '
The. resonance escape prob. decr. due to Pu buildup'and the thermal.
1 utilization increases due to rod withdrawal (WHY e 0.75). This causes
.less react.. change for a IF moderator temp. change. Thus, as the core
- cges, MTC-becomes less negat;ve (HOW e 0.75.'. . (CONCEPT)
REFERENCE
, RBS GP Reactor Theory Student Handout, 27 ~ l i
i ~ ANSWER 5.06 (2.00) '
, .cn 7, b. 6, c. & d. 2 of.4/8/9, e. 2, f. 2, g. 1, h. ~1 l .(:8 answers e 0.25 ea.)
REFERENCE i
RBS GP Reactor Theo'y r Student Handout'.31-1/5 l
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- IBEBdQDINad101 ANSWERS - RIVER BEND 1 -85/07/29-PELLET,- J.
ANSWER 5.07 (3.00)
.a. Sm: 12 days (0.5); Xe: 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (0.5)
- b. A.startup'is said.to be xenon-precluded when the excess react.
available in the core is less than-the neg. reac. from a Xe-peak (0.75)
- c. A xenon-precluded startup.is most likely to occur at EOL conditions (where excess reactivity is ~ at a minimum). (0.75) d.- This condition can exist for up to about a day'C~). (0.5)
REFERENCE RBS'GP. Reactor Theory Student Handout, . 32-4, 33-5 ANSWER 5.08 (2.00)
Ges bin'ing d occurs when a pump is filled with a gas, such as air, so that the pump'cannot pump'the. fluid: desired (1.0). Cavitation occurs when inadequate NPSH is available so voiding or boiling of-the pumped fluid occurs in the pump suction-(1.0).
REFERENCE r
RBS GP Thermodynamics, 4-4 ANSWER 5.09 (2.00) ,
- n. 2, 3, (5)
- b. 4, (5).
- c. (4), 5, 6, 7 i
- d. 1, (5), (7) (a-d at-0.5 ea.; () answers OK but not req'd.)
REFERENCE-RBS GP Heat Transfer Fundamentals, p. 17, 18 i
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c .vn.,, -n - . , , . . . . . , . , - - , -
3-n.., ,.- -,-,-.--,-,-,n,. .,--,.,,,,,,,,,,--,wn,-r-n, .,y, r - m wnne-,.-n+,,,.,-~.,,,..n,-, ,
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ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, 1.
ANSWER 5.10 (3.00)
- c. dPorifice 20~60#, dPcore ~5# OR dPorifice 4~10* dPcore (CONCEPT)
- b. Core orificing is required to assure high power / void fraction channels get enough flow to prevent fuel / clad damage.
- c. Core orifices add a large resistance to flow to each channel so that any changes in resistance due to increased voiding are small compared to the total dP across a-channel.
REFERENCE R8S GP Thermal-Hydraulics, p. 28 RBS FSAR Table 4.4-1 ANSWER 5.11 (2.00)
- c. The power at which the onset to transition boiling will occur.
- b. i. Increase.
ii. Increase.
iii. Increase.
iv. Decrease.
(s. 0.8' CONCEPT; b. 4 answers e 0.3 ea.)
REFERENCE RBS GP 8WR THERFAL-HYDRAULICS, p. 37 GE THERMODYNAMIC 3 HEAT TRANSFER AND FLUID FLOW, p. 9-89 ANSWER 5.12 '( .50)
. c. TRUE. (0.25)
- b. i. clad temperature (0.125)
- 11. 2200 (0.125)
REFERENCE RBS GP THERMAL-HYDRAULICS, p. 65 l_ J
kt__EL8HIi1XIIEdi_QE11GNt_GQNIBQLt_6NQ_IU118UDENI8I1QN PAGE 18
}
ANSWERS -- RIVER BEND.1 -85/07/29-PELLET, J.
i
. ANSWER 6.01 (1.50)
- 1. Dilution of the offgas stream.
- -
- 2 . . Hydrogen recombiners.
4 (2 answers e 0.75 es.)
4 j' REFERENCE-RBS OFFGASiSYSTEM, .p. 41-2
]
i i
j ANSWER 6.02 (3.00) i Alt. Source
- Station Batt* Primary Source * '
i (480vac 1ph) (125vde) (480vac 3ph) i i i l BKR BKR BKR
(~~~~~g ~~~~~~) ALT {~~~l~~~~~~~~~~~~~~~~~~~~~~~~~l~~~~~~~)
{ 480/120V )SRC { BKR BKR }
{ REG-XFORM* }XFORM* { l. l- }
{ l ) { DIODEC>l)* RECTIFIER *- )
} { _________________
{ i i }
j { l ) i {- __________________________i. ) ,
, { 0, } I { SYNC:CHK----------INVERTER * -) .
- { l ) I { l I- )
{ MAN BYP------- --------D---------STATIC----__---- )
{ SWITCH * ) I { SWITCH * )
{- I ) I ( 'l }UNINT.
, ~~~~~i-~~~~~~ o_____________ }ps*
l 1 { }-
i ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ ,
- (LOAD)
(11
- answers e 0.2 ea., 9 other answers e 0.05 es., 0.35 for wiring / layout)-
a
- e i REFERENCE
t.
l I
.2
[
1
__.b _
ht__PL6NI_1111Ed$_QEHIGNt_QQUIBQ(t_6NQ_lNSIBQUEN1611QN PAGE 19 ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 6.03 (2.00)
- 1. Normal service water extreme low pressure, 55 psig, 1/2 taken twice.
- 2. RPCCW extreme low pressure, 56 psig,~1/2 taken twice.
(total of 6 answers, 3/part, 8 0.,333 ea.)
REFERENCE RBS STANDBY SERVICE WATER SYSTEM, p. 47-4 RBS RPCCW SYSTEM, p. 48-5 ANSWER 6.04 (3.00)
- a. Fast close.
- b. No change.
- c. Release the " MAIN STEAM LINE A OUTBD VALVE TEST" pushbutton (if slow test done) OR place switch to open'(if partial test done).
- e. Reactor. trip (0.1) due to all MSIV's closing (0.4) on high steam flow (0.4) exceeding 140% setpoint when 1 MSIV closes (0.1).
OR High flux scram (0.5) due to pressure increase (0.5).-
(a-d 8 0.5 ea., e- 8 1.0 as marked)
REFERENCE RBS SOP-0011, MAIN STEAM SYSTEM, Rev.: DRAFT, p. 8, 9 s
3 4
d
[* ,
l Ez__ELaNI_SIEIEdi_QESIQUt_QQNIBQLt_aNQ_IN11BudENI611QN PAGE 20 I
j ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
1 I
I ANSWER 6.05 (3.00) 1 a. Primary Containment Integrity is required:
- 1. during power operations, (0.33)
} 2.. during startup or hot standby, and , (0.33) i
- 3. during hot shutdown with reactor temperatures above 200 F; (0.34) i OR
- any plant mode with reactor temperature above 200 F. (1.0) 4
- b. 1. All primary containment penetrations required to tut closed during.
I an. accident are either closed or operable automatically.
j 2. All primary containment equipment hatches are closed and sealed.
- 3. All primary containment air. locks operable.
- 4. Primary containment leak rate within allowable limits.
- 5. The suppression pool is operable.
6.' The seal mechanisms for each primary containment seal are operable.
! (any 4/6 answers e 0.5 ea.)
! l 1 REFERENCE RBS PRIMARY CONTAINMENT, p. 23-28, 29, 30 ANSWER 6.06 (2.00) 4 1
- a. A highe'r than normal setpoint (+55") is set'for 10 seconds (0.5)
! then a lower then- normal setpoint (19") is set (0.53 (until level i >10" & reset).
i b. .Setdown circuit must be manually reset (via pushbutton on H13-P680). >
i c. Helps prevent high level trip of the reactor feed pumps (by attemping
(a 0 1.0, b e 0.5, c e 0.5) l
! REFERENCE i RBS FEEDWATER CONTROL SYSTEM, p. 33-8, 9 i.
t 4
v l ANSWER 6.07- (2.00) t f
- c. Upper c. Lower
- b. Upper d. Neither REFERENCE l RBS REACTOR 1 RECIRCULATION SYSTEM, p. 7-18, 19 i
4
ki__ELeNI_SISIEdS_QEEIGNt_QQNIBQLt_8NQ_INSIBydgNI611QN PAGE 21 l
ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 6.08 (3.00)'
LPCI - restore vessel level after LOCA - 3 loops
' Suppression Pool-Cooling remove heat from'supp. pool - 2 loops Shutdown Cooling - remove. decay heat from reactor -2 loops Steam Condensing -' limit pool heat load and reactor makeup required when the vessel is isolated - 2 loopsL Fuel Pool Cooling - supplement SPC as required - 2 loops Containment Flooding - inj ect Service Water to the vessel - 1 loop (any 5/6, mode-0.2, purpose-0.3, #1 oops-0.1)
REFERENCE RBS RESIOUAL HEAT REMOVAL SYSTEM, various pages ANSWER 6.09 (2.00)
- a. 8 e. INOP i. FLUX >'(0.66Wr +.48)t,
- b. 16 or 17 f. DWNSCL
- c. LPRM- g. UPSCL H1 ALM
- d. 5 h. UPSCL HI / NEUTRON FLUX Cc-1 0 0.222 ea.)
REFERENCE RBS AVERAGE POWER RANGE MONITORING SYSTEM, p. 19-1 to 19 ANSWER 6.10 ( .50)
FALSE. (Only if RCIC already initiated - a' low RV level exists.).
REFERENCE
- RBS RCIC, various pages ANSWER 6.11 (2.00)
- a. 3, 8 d. 8 g. 2
- b. 3 e. 1 h. 1-
- c. 2 f. 8 1. 2
-(10 answers at 0.2 ea.)
. .. . -. --. -.- . . . - _ _ _ _ = - - . - - . - . . . - - .
1> LEz__PL6NI_glgIgdg_QESIGNt_QQNIBQLt_6NQ_IN11BQUENI611QN PAGE 22 u ANSWERS -- RIVER BEND 1 -85/07/29-PELLET,-J.
L, I
REFERENCE ,
l RBS NUCLEAR BOILER PROCESS INSTRUMENTATION SYSTEM, p. 3-19, 20-
{
L i
- -ANSWER 6.12 '(1.00)
RCIC-steam supply stop valve (F045) opens.
l- RCIC pump supply to lube oil. cooler (F046)-opens.
1 RCIC. minimum flow. bypass stop valve (F019) opens.
RCIC turbine accelerates into normal speed range.
] RCIC injection shutof f valve (F013) opens.
RCIC inj ection inboard check valve (F066) opens.
RCIC inj ection outboard check valve (F065) opens.
! RCIC pump minimum flow bypass stop valve (F019) closes.
- RCIC system flow increases to 600 gpm.
l RCIC pump discharge pressure >/= rx pressure, p, RCIC gland seal compressor running.
(any 4 answers 8 0.25 ea.)
f REFERENCE e
)
' i A
I I_
4 L i f
l.
i I ,
L l
4 l
f i
Zt__EBQQEQVBES_:_NQBdekt_8ENQBd86t_EMEBGENQ1_66Q PAGE 23 B8Q10LQQ1QaL_CQNIBQL ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 7.01 (3.00)
- c. The coolant is deaerated by lining up the RPV vent to the condenser that has a vacuum, causing boiling at low temperature (170 F) and subsequent deseration (1.0). Solid impurities are removed by routing through -the RWCU System (1.0).
- b. Level is maintained by rejecting coolant via RWCU blowdown (1.0).
REFERENCE GOP-0001, p. 11, 15 ANSWER 7.02 (1.50)
- a. During a restart from.high power (i.e., Xe inventory high).
- b. In the range of notch positions ~0-24.
- c. Near criticality (to avoid short periods).
(3 answers e 0.5 es.)
REFERENCE GOP-0001, p. 7 ANSWER 7.03 (2.50)
- n. 1. Observe Nuclear Instrument response.
- 2. Absence of " Rod Overtravel" annun, when attempting to w/d past 48.
- b. 1. Fuel Conditioning Limits.
2.. Power / Flow Map.
- 3. "V" Curve. )
- 4. Capability Curve. )-Turbine Load Limits.
- 5. Load Change Curve. }
(any 3/5 0 0.5 ea.)
REFERENCE GOP-0002, p. 12, 18
l Zz__EBQQgQUBEg_:_NQBeatt_aeNQBd6Lt_EuEBGENQ1_6NQ PAGE 24 B8Q10LQQIQeL_QQNIBQL ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 7.04 (3.50)
- c. Beta dose rate = open - closed window = 0.9 Rem /hr (0.5)
- b. Stay time: 10CFR20: 1.25 R(imts)/1.75 R/hr = .71 hr = 43 minutes (0.5)
Admin.: 1.00 RC1mts)/1.75 R/hr = .57 hr = 34 minutes (0.5)
- c. 1. 75 Rem to save a life.
2.'25 Rem to protect equipment. .(0.5 es. x 2)
- d. 1.75 x (2 ft.)*2 = Dose x (10 ft.)^2 (relation w/ dist 0.7, math 0.3)
Dose = 1.75 x 4 / 100 = 0.07 Rem /hr.
REFERENCE GET II, p. 5, 6, 36 ANSWER 7.05 (2.50)
- a. 1. New fuel vault ---------- Dry
- 2. Spent fuel storage pool - Either
- 3. Upper contmt. fuel pool - Either
- b. 1. Spent fuel pool --------- No specific restrictions
- 2. Upper contmt. fuel pool - During refueling or outage (5 answers e 0.5 ea.)
REFERENCE FHP-0003, p. 3 ANSWER 7.06 (3.00)
- c. 1. OPEN: Turbine Bypass ~(in auto or manual, or direct steam to SJAE, Turbine seals, Red Weste Reboiler, Off Gas Pre-heater, or MSL drains). (1.0)
- b. 1. Cause of trip found.
- 2. Maint./ testing to. correct cause completed.
- 3. Other post-scram recovery steps per ref.
(2 answers req'd e 0.5 en.)
REFERENCE GOP-0007, p. 5, 6, 9
Zi__EBQQEQuBE1_:_NQBdeL&_eENQBueLt_EMEBGENQ1_8NQ PAGE 25 BeQIQLQQlgeL_QQUIBQL ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 7.07 (2.50)
- e. 1. Low vessel level. 4. RPV isolation.
- 2. High vessel pressure. 5. Scram. required and power > 5%.
- 3. High drywell pressure. 6. Scram req'd and power indeterminate.
(any 4/6 8 0.25 ea.)
- c. Inj ect boron: 1. Via SLC.
- 3. From hotwell via Feed and/or Condensate pumps.
(3 answers e 0.25 ea.)
REFERENCE E0P-0001, p. 5, 7 A0P-0021, p. 2 ANSWER 7.08 (2.50)
- a. When normal RHR flow path is not available (0,5).
- b. Water from vessel through manually open SRV's to suppression pool (0.5). Return to vessel via RHR or LPCS pump (0.5). Heat sink is supp. pool to RHR hx to Srve Wtr or SWP (0.25).
- 2. Recirc loop temp. 4. RPV metal temp. (any 3/4 0 0.25 ea.)
REFERENCE A0P-0021, p. 3 ANSWER 7.09 (2.00)
- c. 1. Manually scram.
- 2. Verify rods on. bottom.
- 3. Mode sw. to shutdown.
- 5. Trip turbine. (5 answers e 0.2 ea.)
- b. To ensure that the reactor is shutdown and the MSIV's are shut. (1.0)
REFERENCE A0P-0031, p. 4
Zt__EBQQEQVBES_ _NQBd8Lt_aEUQBdakt_EUEBQENQ1_6NQ PAGE 26 BaQIQLQQ196L_QQNIBQL ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 7.10 (2.00)
- c. 1. Flooding required with less than 7 SRV"s open.
- 2. Suppression pool temp. and RVP press. cannot be restored to below Heat Capacity Temperature Limit.
- 3. Drywell temp. cannot be maintained (below 330 F).
- 4. Suppression pool level cannot be maintained above Heat Capa-city Level Limit or below pool load limit.
- 5. Secondary containment temperature, radiation levels, or water levels exceed max. safe op, limits.
- 6. RPV pressure is too high for inj ect ing adequate flow.
- 7. Containment temperature cannot be maintained below limit (185 F).
- 8. Containment-annulus differential pressure reaches 5.0 psid.
(any 3/8 0 0.333 ea.)
- b. 1. Temp. near cold reference leg instrument vertical runs, ex-ceeds the RVP saturation limit.
- 2. RPV water level cannot be determined.
- 3. Containment to annulus or Jrywell to containment different-ial pressures cannot be maintained below limits.
- 4. Drywell temp, cannot be maintained < limit (330 F).
(any 2/4 answers e 0.5 ea.)
REFERENCE ECP-0001, p. 20 E0P-0002, p. 13, 17
Hz__6DMINISIBoIIVE_EBQQEQUBESt_CQUQlIl0N$t_6NQ_LidlI6IlQN1 PAGE 27 ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 8.01 (1.00)
- b. NR_____ Non-routine (type of report) 02_____ Period or month
-84____ Year
-1_____ Sequential report for calendar year. (4 answers e 0.25 ea.)
REFERENCE ADM-0010, p. 4,-5 ANSWER 8.02 (2.00)
- c. 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />
- b. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
- c. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- d. Accept any of the following: Assistant Plant Manager, Plant Manager, or "a higher level of management."
(4 at 0.5 each)
REFERENCE River Bend Technical Specification 6.2.2.f., second draft, 3/11/85 ANSWER 8.03 (3.00)
- e. 1. Time period between scheduled completion and over due date. (0.5)
- 2. A report that provides warning of surveillance that may ex-ceed the tolerance period. (0.5)
- 3. A report that provides a list of tests for.which exceptions have been recorded on the last performance. (0.5)
- b. Stop the test (0.5) and either issue a temporary change (0.5), or revise the procedure (0.5) per administrative requirements.
REFERENCE ADM-0015, p. 4, 5, 9
A&__8QUINISIBoIIVE_EBQQEQUBESt_QQUQlIl0NSt_6NQ_LidlI611QN1 PAGE 28 ANSWERS -- RIVER BEND 1 -85/07/23-PELLET, J.
ANSWER 8.04 (2.50)
- o. 1. When alteration and restoration are steps of an approved procedure.
- 2. When alteration is of a momentary nature for testing, etc.
(2 answers e 0.75 ea.)
- b. Independent verification is required for alt. and restoration (1.0)
REF$RENCE ADM-0031, p. 2, 5 ANSWER 8.05 (3.00)
- c. 1. Simulated loss of off-site power.
- 2. Simulated loss of off-site power and ESF actuation test signal.
- 3. ESF actuation test signal.
- 4. -Manual. (3 answers 0 0.333 ea.)
- b. 1. Fuel day tk. level. 4. EDG load test.
- 2. Fuel storage tk. level. 5. Starting air pressure.
- 3. Fuel transfer pump. 6. Electrical alignment.
(any 4/7 0 0.5 ea.) 7. SV TS for > 30 days.
REFERENCE RBS TS, p. 3/4 8-3 -
4 ANSWER 8.06 (1.00)
Fttlure to trip on overspeed could result in the turbine becoming a missile hczard (which could impact safety.related equipment, systems or barriers).
REFERENCE RBS TS, p. 3/4 3-7 .
l l
l l
l
r Sz__8QU181118eIIVE_EBQGEQUBESt_CQUDIl10NSt_6UQ_L101IeIIQUE PAGE 29 ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
ANSWER 8.07 (2.00)
- 1. Drywell particulate monitoring system.
- 2. Drywell and pedestal floor drain sump flow monitoring system.
- 3. Drywell atmosphere gaseous monitoring system.
- 4. Drywell coolers condensate flow rate monitoring. (3/4 ans. e 0.666 ea.)
REFERENCE RBS TS, p. 3/4 4-8 ANSWER 8.08 (1.50)
Must have at least two SRM's (0.5). One must be located in the qua-drant where alterations are being performed and one in either ad-jccent quadrant (0.5). A detector in the opposite quadrant would not see the effects of change as rapidly as a detector in an adj acent quadrant (0.5).
REFERENCE RBS TS, p. 3/4 9-3 ANSWER 8.09 (2.00)
- a. 1-NOUE d. 3-SAE g. 2-Alert
- b. 4-GE e. 3-SAE h. 2-Alert
- c. 2-Alert f. 1-NOUE (8 answers 8'O.25 ea.; 1 too severe -0.05; 2 more severe -0.1)
REFERENCE EIP-2-001, p. 9, 11, 18, 20, 27, 36, 39 ANSWER 8.10 (2.00)
- c. Shift Supervisor (Emergency Director - ED) is responsible.
- b. Shift Communicator (or ED) makes actual notification.
- c. ED/ Shift Comm./TSC Comm./ EOF Comm./ Staff as avail.
- d. One half hour.
s .. ,
a___6DdlNISIB6I1YE_EBQQEQMBEnt_QQUQlIl0 Ult _6NQ_L1511611QU$ PAGE 30 ANSWERS -- RIVER BEND 1 -85/07/29-PELLET, J.
REFERENCE EIP-2-006, p. 3, 4 ANSWER 8.11 (3.00)'
O. 25%
- b. 785 psig
- c. 10%
- d. 1.06
- c. 1325 psig
- f. the top of the active irradiated fuel (6 at 0.5 each)
REFERENCE River Bend Technical Specifications, Safety Limits ANSWER 8.12 (2.00)
- e. Limited. Evac. b. Building Evac, c. Protected Area Evac,
- d. Owner Controlled (Site / Exclusion) Area Evac.
(4 ar.awers e 0.5 ea. - any order)
REFERENCE EIP-2-026, p. 2
i M C RULES A m EUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
l 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
l 2. Restroom trips are to be limited and only one candidate at a time may leave.
i 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 facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided fer answers.
i 7. Print your name in the upper right-hand corner of the first page of each j section of the answer sheet.
! 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write on only olie 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.
j 11. Separate answer sheets from pad and place finished answer sheets face down on j your desk or table.
- 12. Use abbreviations only if they are comonly used in facility literature.
l i 13. The point value for each question is indicated in parentheses after the j question and can be used as a guide for the depth of answer required. ,
i 14. Show all calcualtions, methods, or assumptions used to obtain an answer to j 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 BLAE.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
i 17. You must sign the statement on the cover sheet that indicates that the work is
! your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
- 18. When you complete your examination, you shall:
l a. Assemble your examination as follows:
!- Exam questions on top.
l Exam aids - figures, tables, etc.
I Answer pages including figures which are a part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions. ,
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
- e. Do not dicuss the examination with other licensee staff personnel until the formal examination review is complete.